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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 418 nuclides; (2) Covariance uncertainty data for 185 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions at higher energies for isotopes of F, Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new Decay Data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [1].

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Analyzing data from a PDV microphone

Photon Doppler Velocimetry (PDV) is used at Los Alamos National Laboratory to measure extremely fast velocities (on the order of 0.01 - 5 km/s) of projectiles during material testing procedures. This kind of testing is extremely important to understand how materials behave in extreme conditions (such as in nuclear explosions). PDV data is created by summing an incident beam with a Doppler shifted reflection, forming a beat frequency. The beat frequency is directly proportional to the velocity of the reflecting surface. We explored a new way of using PDV: measuring slow velocities (less than 1 m/s). The principle method we used was to measure vibrations in a Mylar sheet due to incident sound waves. I analyzed these files using both my own coding as well as a program designed at NSTec for this purpose. Analysis included performing Fourier Transforms, extracting velocity as a function of time, and playing the velocity function through a speaker to recreate the initial sound. We were able to create a successful microphone, though playback quality was poor. The limiting factor was the low velocity of the Mylar sheet, which was about 0.2 m/s. Other methods are now being explored to extract velocity data in a more efficient manner. During this project I gained extensive experience in programming, data analysis, problem solving, and communication. I learned that one must be creative in order to solve complex problems, approaching them from various angles. I also gained a lot of practice working on a team, which consisted of a high school student, another college intern, and our mentor.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2022

Self-powered Through-wall communication for dry cask storage monitoring

Many nuclear facilities, such as spent fuel storage dry casks and nuclear reactor pressure vessels, are entirely sealed by metal layers to prevent harmful radiation. For safety and security operations, the temperature, pressure, radiation, and humidity inside the vessel needs to be closely monitored. However, no practical technology is currently available to realize the through-wall data communication and monitoring for these vessels due to the inside harsh environment of high temperature and nuclear radiation. In this paper, an innovative self-powered wireless through-wall data communication system for the nuclear environment is presented, which demonstrates a successful solution to such challenges. The presented system is composed of four modules, i.e., energy harvester with power management circuits, ultrasound wireless communication using high-temperature piezoelectric transducers, electronic circuits for sensing and data transmission, and radiation shielding for electronics. Here, constitutive functions of each module were firstly designed and followed by the system integration. Experiments were conducted subsequently to validate the designed functions and evaluate the performance of the integrated system. Results showed that the average power of over 40 mW was harvested from the thermal flow inside the nuclear spent fuel canisters which could provide enough energy to operate the sensing and data communication systems. The gamma radiation test results showed that the thermoelectric energy harvester and ultrasound transceivers can withstand radiation dosing over 100 Mrad. Furthermore, temperature shock tests demonstrated that the entire system including the shielded electronics can survive and maintain their functionalities at temperatures as high as 195°C. Under the in-lab mocked-up high temperature conditions and radiation shielding, the proposed system is foreseen to survive and operate stably for fifty years inside a nuclear spent fuel canister, and send the frequency modulated data out of the canister for 3 s in every 10 min.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

DATA QUALITY OBJECTIVES FOR SELECTING WASTE SAMPLES FOR THE BENCH STEAM REFORMER TEST

This document describes the data quality objectives to select archived samples located at the 222-S Laboratory for Fluid Bed Steam Reformer testing. The type, quantity and quality of the data required to select the samples for Fluid Bed Steam Reformer testing are discussed. In order to maximize the efficiency and minimize the time to treat Hanford tank waste in the Waste Treatment and Immobilization Plant, additional treatment processes may be required. One of the potential treatment processes is the fluid bed steam reformer (FBSR). A determination of the adequacy of the FBSR process to treat Hanford tank waste is required. The initial step in determining the adequacy of the FBSR process is to select archived waste samples from the 222-S Laboratory that will be used to test the FBSR process. Analyses of the selected samples will be required to confirm the samples meet the testing criteria.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Fallout Cloud Regimes

The U.S. Department of Defense (DOD), Department of Energy (DOE), and other organizations maintain operational nuclear explosion and atmospheric dispersion models to provide critical guidance on the expected effects of an accidental or deliberate explosion of a nuclear weapon (in this paper simply referred to as “device”). To be effective, these models must represent, as accurately as possible, the complex interactions of the blast, fire, and residual radiological hazards with the environment and population. One hundred atmospheric nuclear tests that form the basis for many models were conducted at the Nevada Test Site (NTS) (now referred to as the Nevada Nuclear Security Site, NNSS) in a dry desert environment. Other environments should be studied, but have less data available and are beyond the scope of the work presented in this paper. The debris clouds produced by the NTS tests, frequently called “mushroom clouds,” are familiar, with common structural elements such as a buoyant cap connected to a skirt of raised dust at the desert surface by a thin, dirt-filled stem. The film scanning project at LLNL has investigated historical film records of nuclear weapons tests. Here, we summarize findings showing that the mushroom cloud behavior for historic U.S. tests conducted in Nevada, has similar characteristics based on the distance of the device from the ground surface or Height of Burst (HOB), scaled by the energy release, or yield, of the device. This scaled height is referred to as the scaled-height-of-burst (SHOB). The findings discussed below show that mushroom clouds look and behave similarly when detonated at the same SHOB. The amount of residual radiation that is produced by a nuclear detonation is proportional to the yield. But, the amount of that residual radiation that actually becomes local fallout is strongly dependent on the SHOB and the type of surface over which the detonation occurs. In order to develop a more comprehensive model that predicts the fraction of the residual radiation that becomes local fallout, it is convenient to define a series of regimes based on SHOB values in which all detonations that occur within a given regime can be modeled using the same algorithms. The purpose of this paper is to provide a framework for defining different regimes, and, in a qualitative way, a basic understanding of the fundamental characteristics of each of these regimes.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2020

Direct astrophysical tests of chiral effective field theory at supranuclear densities

Recent observations of neutron stars with gravitational waves and X-ray timing provide unprecedented access to the equation of state (EoS) of cold dense matter at densities difficult to realize in terrestrial experiments. At the same time, predictions for the EoS equipped with reliable uncertainty estimates from chiral effective field theory ( x EFT) allow us to bound our theoretical ignorance. In this work, we analyze astrophysical data using a nonparametric representation of the neutron-star EoS conditioned on x EFT to directly constrain the underlying physical properties of the compact objects without introducing modeling systematics. We discuss how the data alone constrain the EoS at high densities when we condition on x EFT at low densities. Here, we also demonstrate how to exploit astrophysical data to directly test the predictions of x EFT for the EoS up to twice nuclear saturation density, in order to estimate the density at which these predictions might break down. We nd that the existence of massive pulsars, gravitational waves from GW170817, and NICER observations of PSR J0030+0451 favor x EFT predictions for the EoS up to nuclear saturation density over a more agnostic analysis by as much as a factor of 7 for the quantum Monte Carlo (QMC) calculations used in this work. While x EFT predictions using QMC are fully consistent with gravitational-wave data up to twice nuclear saturation density, NICER observations suggest that the EoS stiffens relative to these predictions at or slightly above nuclear saturation density. Additionally, for these QMC calculations, we marginalize over the uncertainty in the density at which x EFT begins to break down, constraining the radius of a 1.4M $\bigodot$ neutron star to R 1.4 = $11.40^{+1.38}_{–1.04}$ ($12.54^{+0.71}_{–0.63}$) km and the pressure at twice nuclear saturation density to p(2n sat ) = $14.2^{+18.1}_{–8.4}$ ($28.7^{+15.3}_{–15.0}$) MeV=fm 3 with massive pulsar and gravitational-wave (and NICER) data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Core-to-Specimen Energy Coupling Results of the First Modern Fueled Experiments in TREAT

For this study, a series of experiments was conducted in the Transient Reactor Test (TREAT) Facility following a decades-long operational hiatus. A series of physics tests were then performed emphasizing relevant transients for Light Water Reactor (LWR) fuels research. A series of fueled experiments were then conducted using LWR-type sub-length specimens in inert gas capsules with the primary objective to quantify the ratio of nuclear heat deposited in the core to that deposited in the specimens. This type of data can be difficult to measure but is highly important in the design and interpretation of transient tests. Five fueled capsules were irradiated and energy coupling factors (which express the ratio of nuclear heat deposited per mass in the test specimen to that deposited in the reactor core in units of J/gUO 2 MJ TREAT ) were observed ranging from 2.0 to 2.2 over a wide range of transient energies with good agreement between measurement methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Can we safely go to 200 °C? An integrated approach to assessing impacts to the engineered barrier system in a high-temperature repository

Abstract. This presentation gives on overview of the complex thermo-hydro-mechanical-chemical (THMC) processes occurring during the disposal of heat-producing high-level radioactive waste in geologic repositories. A specific focus is on the role of compacted bentonite, which is commonly used as an engineered backfill material for emplacement tunnels because of its low permeability, high swelling pressure, and radionuclide retention capacity. Laboratory and field tests integrated with THMC modeling have provided an effective way to deepen our understanding of temperature-related perturbations in the engineered barrier system; however, most of this work has been conducted for maximum temperatures around 100 ∘C. In contrast, some international disposal programs have recently started investigations to understand whether local temperatures in the bentonite of up to 200 ∘C could be tolerated with no significant changes to safety relevant properties. Raising the maximum temperature is attractive for economical and safety reasons but faces the challenge of exposing the bentonite to significant temperature increases. Strong thermal gradients may induce complex moisture transport processes while geochemical processes, such as cementation and perhaps also illitization effects may occur, all of which could strongly affect the bentonite and near-field rock properties. Here, we present initial investigations of repository behavior exposed to strongly elevated temperatures. We will start discussing our current knowledge base for temperature effects in repositories exposed to a maximum temperature of 100 ∘C, based on data and related modeling analysis from a large heater experiment conducted for over 18 years in the Grimsel Test Site in Switzerland. We then show results from coupled THMC simulations of a nuclear waste repository in a clay formation exposed to a maximum temperature of 200 ∘C. We also explore preliminary data from a bench-scale laboratory mock-up experiment, which was designed to represent the strong THMC gradients occurring in a “hot” repository, and we finally touch on a full-scale field heater test to be conducted soon in the Grimsel Test Site underground research laboratory in Switzerland (referred to as HotBENT).

Birkholzer, Jens T. (ORCID:0000000279891912)↗
DOE OSTI2014

Reversible Bending Fatigue Testing on Zry-4 Surrogate Rods

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. Research and comparison studies conducted at Oak Ridge National Laboratory (ORNL) resulted in a new concept in 2010 for a U-frame testing setup on which to perform hot-cell reversible bending fatigue testing. Subsequently, the three-dimensional finite element analysis and the engineering design of components were completed. In 2013 the ORNL team finalized the upgrade of the U-frame testing setup and the integration of the U-frame setup into a Bose dual linear motor test bench to develop a cyclic integrated reversible-bending fatigue tester (CIRFT). A final check was conducted on the CIRFT test system in August 2013, and the CIRFT was installed in the hot cell in September 2013 to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The fatigue responses of Zircaloy-4 (Zry-4) cladding and the role of pellet pellet and pellet clad interactions are critical to SNF vibration integrity, but such data are not available due to the unavailability of an effective testing system. While the deployment of the developed CIRFT test system in a hot cell will provide the opportunity to generate the data, the use of a surrogate rod has proven quite effective in identifying the underlying deformation mechanism of an SNF composite rod under an equivalent loading condition. This paper presents the experimental results of using surrogate rods under CIRFT reversible cyclic loading. Specifically, monotonic and cyclic bending tests were conducted on surrogate rods made of a Zry-4 tube and alumina pellet inserts, both with and without an epoxy bond.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Assessing the Predictive Capability of the LIFEIV Nuclear Fuel Performance Code Using Sequential Calibration

This report considers the problem of calibrating a numerical model to data from an experimental campaign (or series of experimental tests). The issue is that when an experimental campaign is proposed, only the input parameters associated with each experiment are known (i.e. outputs are not known because the experiments have yet to be conducted). Faced with such a situation, it would be beneficial from the standpoint of resource management to carefully consider the sequence in which the experiments are conducted. In this way, the resources available for experimental tests may be allocated in a way that best 'informs' the calibration of the numerical model. To address this concern, the authors propose decomposing the input design space of the experimental campaign into its principal components. Subsequently, the utility (to be explained) of each experimental test to the principal components of the input design space is used to formulate the sequence in which the experimental tests will be used for model calibration purposes. The results reported herein build on those presented and discussed in [1,2] wherein Verification & Validation and Uncertainty Quantification (VU) capabilities were applied to the nuclear fuel performance code LIFEIV. In addition to the raw results from the sequential calibration studies derived from the above, a description of the data within the context of the Predictive Maturity Index (PMI) will also be provided. The PMI [3,4] is a metric initiated and developed at Los Alamos National Laboratory to quantitatively describe the ability of a numerical model to make predictions in the absence of experimental data, where it is noted that 'predictions in the absence of experimental data' is not synonymous with extrapolation. This simply reflects the fact that resources do not exist such that each and every execution of the numerical model can be compared against experimental data. If such resources existed, the justification for numerical models would be reduced considerably. The authors note that the PMI is primarily intended to provide a high-level, quantitative description of year-to-year (or version-to-version) improvements in numerical models, where these descriptions can be used as a means of justifying funding requests to support further model development research. It is in this context that the present report should be considered: the availability of data from experimental tests should be viewed as a time-dependent variable, where experiments are added to the calibration suite as resources become available. For the present report, the experimental data is of course already available (permitting demonstration of the proposed methodology). Furthermore, the authors are not proposing this methodology as the answer to the question of how to allocate resources for experimental tests, and readers are directed to [5] and the references contained in Section 1 of [5] for additional information on the subject. However, the strength of this methodology is that it offers a means by which to select the sequence of experiments in a pre-arranged experimental campaign (a situation for which the methods discussed in [5] are less appropriate). The report is organized as follows. Section 2 describes the methodology employed to formulate the sequences of experiments for the calibrations performed for this study. Section 3 then presents the results associated with two sequences; supplementary results are provided in the Appendix. The report then concludes in Section 4 with a brief summary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. Thirteen additional tests would be performed in the CABRI reactor over the next decade on fuel rods with higher burnups [1]. Additionally, a larger testing program at the NSRR reactor in Japan with high burnup fuel would uncover a similar trend of pre-DNB ruptures in high burnup test rods at lower-than-expected peak enthalpies [2]. Numerous out of pile testing programs involving a variety of innovative mechanical testing techniques have been employed in an attempt to better quantify the failure thresholds of corroded zirconium alloy cladding in these rapid heating and loading conditions [3][4][5]. While previously interim guidance had been issued, in June of 2020 the NRC officially published updated regulatory guidance to account of these pre-DNB failures in Regulatory Guide 1.236 [6]. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

INDEN Fissile Actinide Issues [Slides]

This presentation addresses the outlook for the International Nuclear Data Evaluation Network (INDEN), specifically regarding general issues that have been solved, including Thermal Neutron Constants (TNC) and Prompt Fission Neutron Spectra (PFNS). Other topics discussed include outstanding issues in fissile evaluations, comments on the RPI quasi-differential experiment, and reaction rate testing of trial evaluations in fast assemblies. The presentation concludes by discussing how INDEN interactions strongly helped to find deficiencies in existing evaluations and highlighted potential solutions to existing challenges, that updated trial evaluations for U-235, U-233, Pu-239 are available for testing, that further work is expected on U-235, Pu-239, and U-233, and that performance was tested on the ICSBEP and/or SINBAD benchmarks, during which significant improvement was demonstrated. The INDEN scheme of international collaboration on nuclear data evaluation is working well.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Stability of Molten Core Materials

The purpose of this report is to document a literature and data search for data and information pertaining to the stability of nuclear reactor molten core materials. This includes data and analysis from TMI-2 fuel and INL’s LOFT (Loss of Fluid Test) reactor project and other sources.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Accelerating High-Level Waste Glass Corrosion Research with Big Data

Savannah River National Laboratory (SRNL) is accelerating the research of high-level nuclear waste (HLW) glass corrosion using advanced big data techniques available through the Accelerated Leaching and Testing of GLASS (ALTGLASS) database. Developed, maintained and distributed by SRNL, ALTGLASS gives HLW researchers access to chemical durability data for hundreds of glass compositions, allowing scientists to use an “informatics approach” to analyze existing and new experimental data contained in the database to advance glass corrosion science.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Application of Principal Component Analysis to Electrochemical Reprocessing PM and NMAC

In this report, data from an electrorefiner (ER) for nuclear fuel reprocessing is evaluated for process monitoring (PM) conclusions. This data comes from tests performed at the Idaho National Laboratory in 2022. Multivariate approaches utilizing methods of Principal Component Analysis (PCA) is applied. This is based off established work in process monitoring for fault detection in industrial facilities. This report will discuss the background, methods, and results of the application and some of the conclusions and applications that can be drawn from them. PCA is applied to two different electrorefiner (ER) operations that occurred at Idaho National Laboratory between August and October 2022. The first operation occurred with little incident while the second had several noted faults in the equipment in operational logs. The data from the first run was used to train the data for “normal” operations and applied to both sets of data to determine when operations were in an “off-normal” condition and identify where the fault occurs through PCA. PCA was able to identify off-normal events and identify the cause for off-normal operations. These identified off-normal events matched with the events and their causes in the operational logs. However, small amounts of variance in the data led to false detection of “off-normal” events. Thus, careful selection of training data and a-posteriori conclusions based off operator assessments will both be required for application of PCA to PM applications. This work demonstrated that multivariate approaches and latent variables are applicable to pyroprocessing PM applications and can be further expanded in future work as quality variables such as salt concentration from sensors and sampling become available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗