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DOE OSTI2024

A spatially-resolved model of neutron-irradiated tungsten coupling stochastic cluster dynamics and finite deformation plasticity

Structural materials used in nuclear reactors face severe degradation in mechanical properties, such as hardening and embrittlement. At the microscopic scale, this occurs due to creation and accumulation of irradiation-induced defects and their interaction with system dislocations. Although techniques exist which can model evolution of irradiation defects, for instance kinetic transport theory-based models, their interaction with mechanical deformation of the bulk material has not been investigated extensively. In this work, we demonstrate a novel spatially-resolved multiscale coupling between microscopic irradiation defect evolution, modeled using Stochastic Cluster Dynamics (SCD) and macroscopic mechanical deformation modeled using a finite-deformation plasticity model. SCD is used to determine the statistically averaged defect cluster spacing, dependent on operating conditions such as irradiation dose and temperature. This acts as an initial condition that governs the critical resolved shear stress of dislocation glide in the macroscopic plasticity model. This framework is used to predict mechanical behavior in post-mortem test of irradiated Tungsten samples, which has found its importance as structural material used in nuclear reactors. The results obtained using the coupled approach are in good agreement with experimental data of uniaxial tension tests. The model is able to capture the effect of temperature and irradiation dose on the material hardening. Two methods are proposed to estimate hardness – using Tabor's Law relating uniaxial yield stress to hardness and from flat-punch simulations. The results are in reasonable agreement with hardness data from micro-indentation experiments of irradiated Tungsten samples. Finally, the model is also able to reveal microstructural details such as spatial variation in defect density and local stress.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Data Challenges in Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process

Idaho National Laboratory (INL) is maintaining and gaining knowledge into the nuclear fuel cycle by building a test bed to allow researchers the opportunity to study nuclear fuel processing operations. This includes studying solvent extraction processes that use centrifugal contactors. As part of INL’s mission, the goal of this project is to develop a system that utilizes non-traditional measurement sources such as vibration, acoustics, current, light, flow, and temperature in conjunction with data-based, machine learning techniques that will allow for signal discovery. This multisensory data can support the development of safeguards by design, provide operator process awareness, and discover process anomalies. This poster will highlight some of the data collection and analytics challenges for the multi-sensor system as well as the mitigation strategies to build a robust system. Additionally, some preliminary data from the first testing campaign will be shown to help illustrate the data needs of the system.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2015

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Critical Experiment Design Phase 1 Report for Integral Request 441

This report documents the analytical methods used in the first phase of critical experiment design (CED-1) conducted as part of integral experiment request (IER) 441. The purpose of IER-441 is to develop a capability to test the epithermal/intermediate cross sections of materials at the Sandia National Laboratories (SNL) critical experiment facility using low enriched uranium oxide fuel. This effort determined that the capability can be achieved using a varying number of test samples placed in a central test region of a hexagonally pitched lattice driven with a more moderated exterior region. In general, the results show that it is possible to develop sufficient worth to measure any nuclear data deficiencies, as well as the ability to shift the neutron spectrum for dysprosium, hafnium, silver, tantalum, indium, cobalt, tungsten, and antimony. There is marginal capability to test manganese due to its relatively low worth at higher boron concentrations. At this stage, there is interest in identifying the material or materials of greatest interest to the nuclear data community so that this project can proceed to CED-2 with a clear goal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Non-Nuclear MicroReactor Test Bed and Article

The goal of the project is to provide proof of concept using sodium heats pipes as an efficient way to transfer heat from to rods to an integrated Brayton Cycle. A single simulated heat pipe experiment was performed and the results of the experiment are discussed. The results are also compared to the ABAQUS model that was made for the simulated experiment. The second portion of the poster discusses the design and challenges related to the test bed and the instrumentation required for the project. This brushes over the general requirements and the application for the data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Yields and Cavity Radii of Underground Nuclear Tests with Specified Yields Reported in DOE/NV -- 209-REV 15 (2000)

The letter report provides a compilation of yield and cavity radius information for all underground nuclear tests with specified yields reported in DOE/NV -- 209-REV 15 (2000). Yield data were taken directly from DON/NV -- 209-REV 15 (2000). Cavity radius information was taken from a personal computer based database resident at Lawrence Livermore National Laboratory. This database is associated with the written report by Miller et al. (2002).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

An uncertainty quantification method relevant to material test reactors

Within material test reactor calculations, energy dependent flux and reaction rate uncertainties are typically not quantified when performing as-run analyses to determine the neutron field experienced by the experiment. When high fidelity Monte-Carlo codes are used in such analyses, straight forward methods to calculate output uncertainties are not available, instead expert opinion is used to postulate computational uncertainties. New methods to propagate uncertainties through these high fidelity simulations are available when sufficient computational power is available. A tool is developed here for sampling any part of an MCNP input from random distributions to determine output uncertainties based on those inputs. Another tool is developed to sample nuclear data cross-section in ACE format using multi-group nuclear data covariances. The Total Monte-Carlo Method and Gesellschaft für Anlagen-und Reaktorsicherheit method (GRS) are implemented and compared to one another as well as MCNP sensitivity and uncertainty calculations. The methods were applied to the Godiva critical sphere k-eigenvalue, the UAM pin-cell benchmark energy dependent flux and reaction rates, and the Advanced Test Reactor energy dependent flux within an experimental location. Furthermore, the two methods agree well, with GRS allowing for an order of magnitude speedup for reaction rate uncertainty calculations and several orders of magnitude for eigenvalue uncertainty calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Application of Artificial Neural Networks to Reliable Nuclear Data for Nonproliferation Modeling and Simulation

Detection and identification of special nuclear materials (SNMs) are an essential part of the US nonproliferation effort. Modern cutting-edge SNM detection methodologies rely more and more on modeling and simulation techniques. Experiments with radiological samples in realistic configurations, is the ultimate tool that establishes the minimum detection limits of SNMs in a host of different geometries. Modern modeling and simulation approaches have the potential to significantly reduce the number of experiments with radioactive sources needed to determine these detection limits and reduce the financial barrier to SNM detection. Unreliable nuclear data is one of the principal causes of uncertainty in modeling and simulating nuclear systems. In particular, nuclear cross sections introduce a significant uncertainty in the nuclear data. The goal of this research is to develop a methodology that will autonomously extract the correct nuclear resonance characteristics of experimental data in a reliable way, a task previously left to expert judgement. Accurate nuclear data will in turn allow contemporary modeling and simulation to become far more reliable, de-escalating the extent of experimental testing. Modeling and simulation techniques reduce the use and distribution of radiological sources, while at the same time increase the reliability of the currently used methods for the detection and identification of SNMs.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2022

Status of the OECD/NEA Watts Bar unit 1 benchmark and calculations of local reactor core power data of the Zero Power physics tests

The paper aims to provide an update on the status of the Organization for Economic Cooperation and Development (OECD) / the Nuclear Energy Agency (NEA) Tennessee Valley Authority (TVA) Watts Bar 1 (WB1) Multi-Physics Multi-Cycle depletion benchmark and the results associated with Exercise 1. The benchmark relies on the set of benchmark progression problems developed by the Department of Energy (DOE) Consortium for the Advanced Simulation of Light Water Reactors (CASL) for the Virtual Environment for Reactor (VERA), which are based on real plant design and operational data. The OECD/NEA TVA WB1 benchmark is designed for validation of both traditional and novel high-fidelity multi-physics codes to analyze Pressurized Water Reactors (PWR) depletion cycles. The activities are conducted under the Expert Group on Reactor Systems Multi-Physics (EGMUP) at NEA/OECD. In this work, we analyze Exercise 1 of the benchmark: stand-alone Three-Dimensional (3D) neutronics at Start-up Zero Power Physics Test (ZPPT) at Hot Zero Power Conditions (HZP) and the resulting power maps. The code used to model the exercise is the continuous energy Monte Carlo code Serpent 2.1.31 along with the nuclear data library ENDF/B-VII.1. Serpent results are compared with the high-fidelity deterministic code MPACT, which is part of VERA. The paper presents three selected results from power calculations required output. The results compared are the normalized axially integrated radial core power maps, the normalized axial averaged core power shapes, and the normalized core hottest and coldest assemblies radial power maps. The Root Mean Square Deviation (RMSD) between Serpent and MPACT is 0.51 % for the axially integrated radial core power map, 1.41 % for the axial averaged core power shape, 2.42 % for the hottest assembly axial power shape, 0.88 % for the coldest assembly axial integrated hottest assembly radial pin power map, and 0.17 % for the axially integrated coldest assembly radial pin power map. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Conceptual Design of Engineering-Scale Molten Salt Spill Tests to Support Molten Salt Reactor Licensing

Nuclear accident progression analysis codes for reactor licensing must be validated using experimental data that was generated at a large enough scale so that the results are representative of full-scale accidents. This report describes a conceptual design of engineering-scale molten salt spill tests that will provide the necessary datasets to support molten salt reactor (MSR) licensing. Design considerations, salt compositions, test variables, processes to be quantified, and suitable measurement techniques are presented and discussed. Aspects of the conceptual engineering-scale molten salt spill test design presented in this report were made using insights from recent laboratory-scale molten salt spill tests conducted at Argonne. Additional input from model developers, MSR vendors, and technical experts will benefit the design of the proposed engineering-scale molten salt spill tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Characterization of Intrinsic Fiber Optic Temperature Sensors by In-core and Furnace Testing

Increased research and development in nuclear technology has raised the demand for novel sensors and instrumentation to meet data objectives, and survive in different conditions and environments, beyond conventional light water reactor (LWR) environments. Expediting the deployment of advanced nuclear technologies by developing, demonstrating, and qualifying advanced reactor fuel forms necessitates a deeper understanding of how irradiation affects fuels and materials' performance. To achieve a more comprehensive understanding of fuels and materials performance, researchers require more specialized experiments and measurements. The demand for innovative sensors to support nuclear fuel development arises from the complexity of materials' behavior under irradiation and the challenges of deploying instrumentation in Material Test Reactors (MTRs) for irradiation tests. Additionally, measurements of material properties require integrated measurement systems to characterize thermal properties, mechanical properties, chemistry, and microstructure [1]. Among the potential measurement techniques, optical fiber-based sensors have been identified as potential sensors to measure different physical phenomena such as temperature, strain, pressure, and fluid level. Optical fiber sensors have the capability to provide multi-sensing and multiplexing instrumentation, allowing the measurement of different physical parameters within a single sensor configuration, and transmitting data collected at multiple locations through a single fiber. They offer immunity to electromagnetic interference, electrical passivity, compatibility with various sensing methodologies, and cost-effectiveness. Beyond their widespread use in telecommunications, silica fiber-based instruments are utilized in industrial applications, even at temperatures reaching 300?400°C, such as distributed temperature sensing in oil and gas recovery. The Department of Energy (DOE) is interested in using fiber optics to support fuel cycle development [2]. Fiber optics are an excellent candidate for harsh environment sensing, including sensing at very high temperatures (1900oC for sapphire optical fibers). Distributed fiber optic sensing has been deployed in other harsh environments like coal gasification plants [3]. Distributed strain sensing, which operates similarly to distributed temperature sensing, has been deployed to monitor underground mines and fibers have been imbedded in soil to monitor sink-hole development [4][5]. The application of fiber optic sensors to advanced reactor development is a promising area of research.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Perspectives on Tailoring Neutron Energy Spectra in Material Test Reactors

Material test reactors (MTRs) are used to irradiate nuclear fuels and materials to develop data for how they endure neutron bombardment in or near reactor cores. Most historic MTRs, and all that remain operational in the US today, are water-cooled types and produce a thermalized neutron flux. New irradiation facilities are needed which can produce neutron energy spectra relevant to fast and fusion reactor environments. Construction of these facilities will take several years of steadfast funding to complete, which poses a formidable schedule challenge for current fast and fusion reactor developers. Irradiation designs which modify the neutron energy spectra delivered to test specimens in thermal spectrum MTRs, an approach referred to as “spectral tailoring”, can be used to approximate several relevant phenomena in the materials needed to enable fast and fusion reactor technologies. This approach is imperfect, but still valuable in the present situation. Here, the two highest flux MTRs operational in the United States, the Advanced Test Reactor (ATR) and High Flux Isotope Reactor (HFIR), have rich histories, ongoing developments, and new potentials for spectral tailoring that will be reviewed and discussed in this paper.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Energy Harvesting RMSA Field Test

Sandia National Laboratories (SNL) is investigating photovoltaic (PV) cell configurations, integrating them with the battery-operated Remotely Monitored Sealing Array (RMSA), and testing and evaluating performance for enhanced battery life under various environmental conditions at the K-Area Material Storage (KAMS) facility at the Savannah River Site (SRS). Unattended safeguards equipment (e.g. seals) incorporates many low-power electronic circuits, which are often powered by expensive and environmentally toxic lithium batteries. These batteries must periodically be replaced, adding a radiological hazard for both safeguards inspectors and operators. An extended field test of these prototype PV energy harvesting (EH) RMSAs at an operational nuclear facility will give additional data and allow for an analysis of this technology in a variety of realistic conditions, which will be documented in a final report. RMSAs are used for this testing, but SNL envisions energy harvesting technology may be applicable to other safeguards equipment.

25 ENERGY STORAGE↗

DOE OSTI2018

Verification of the MC2-3 Gamma Library

The new gamma library was generated based on the 94-group gamma structure, including delayed gamma and beta data, using the PreGAMMA and GenGAMMA codes with ENDF/BVII. 0 data. The new gamma library was first verified using the 2D benchmark problems that were introduced in the GAMSOR manual, showing that the neutron and gamma heatings from DIF3D were in good agreement with MCNP solutions within 5% even in the non-fuel regions of both cores with uranium or plutonium dominant fuel compositions. Similarly, the DIF3D power solutions for additional simpler 1D problems with 3 or 6 regions showed good agreement with MCNP solutions. Note that all DIF3D calculations were based on the 230- group neutron and 94-group gamma data. For the EBR-II benchmark problem, the DIF3D solution with the new gamma library maintained good agreement in the power distribution with the MCNP solution, which is comparable to the DIF3D results with the current 21-group gamma library. The use of delayed gamma and beta data increased powers at the dummy assemblies by 26%, which appears to be reasonable. Further detailed verification tests may be necessary and, more importantly, validation tests are required against measurement data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Energy Harvesting RMSA Field Test

14 SOLAR ENERGY↗

DOE OSTI2019

Energy Harvesting Remotely Monitored Sealing Array (RMSA) (Field Test Annual Report)

Sandia National Laboratories (SNL) is investigating photovoltaic (PV) cell configurations, integrating them with the battery-operated Remotely Monitored Sealing Array (RMSA), and testing and evaluating performance for enhanced battery life under various lighting conditions at a facility at the Savannah River Site (SRS) or Savannah River National Laboratory (SRNL). Unattended safeguards equipment (e.g. seals) incorporates many low-power electronic circuits, which are often powered by expensive and environmentally toxic lithium batteries. These batteries must periodically be replaced, adding a radiological hazard for both safeguards inspectors and operators. An extended field test of these prototype PV energy harvesting (EH) RMSAs at an operational nuclear facility will give additional data and allow for an analysis of this technology in a variety of realistic conditions, which will be documented in a final report. RMSAs are used for this testing, but SNL envisions energy harvesting technology may be applicable to additional safeguards equipment.

14 SOLAR ENERGY↗

DOE OSTI2013

Progress Report on Reverse Bending Fatigue Testing of Zry-4 Surrogate Rod

Testing high-burnup spent nuclear fuel (SNF) presents many challenges in areas such as specimen preparation, specimen installation, mechanical loading, load control, measurements, data acquisition, and specimen disposal because these tasks are complicated by the radioactivity of the test specimens. Unique features of the test and testing specimens demand that the test plan address these issues from the very beginning.

Fuel Transportation↗

DOE OSTI2017

Analyses of Field Test Data at the Atucha-1 Spent Fuel Pools

A field test was conducted at the Atucha-1 spent nuclear fuel pools to validate a software package for gross defect detection that is used in conjunction with the inspection tool, Spent Fuel Neutron Counter (SFNC). A set of measurements was taken with the SFNC and the software predictions were compared with these data and analyzed. The data spanned a wide range of cooling times and a set of burnup levels leading to count rates from the several hundreds to around twenty per second. The current calibration in the software using linear fitting required the use of multiple calibration factors to cover the entire range of count rates recorded. The solution to this was to use power regression data fitting to normalize the predicted response and derive one calibration factor that can be applied to the entire set of data. The resulting comparisons between the predicted and measured responses were generally good and provided a quantitative method of detecting missing fuel in virtually all situations. Since the current version of the software uses the linear calibration method, it would need to be updated with the new power regression method to make it more user-friendly for real time verification and fieldable for the range of responses that will be encountered.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗