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DOE OSTI2011

SILENE Benchmark Critical Experiments for Criticality Accident Alarm Systems

In October 2010 a series of benchmark experiments was conducted at the Commissariat a Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE [1] facility. These experiments were a joint effort between the US Department of Energy (DOE) and the French CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems (CAASs). This presentation will discuss the geometric configuration of these experiments and the quantities that were measured and will present some preliminary comparisons between the measured data and calculations. This series consisted of three single-pulsed experiments with the SILENE reactor. During the first experiment the reactor was bare (unshielded), but during the second and third experiments it was shielded by lead and polyethylene, respectively. During each experiment several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor, and some of these detectors were themselves shielded from the reactor by high-density magnetite and barite concrete, standard concrete, and/or BoroBond. All the concrete was provided by CEA Saclay, and the BoroBond was provided by Y-12 National Security Complex. Figure 1 is a picture of the SILENE reactor cell configured for pulse 1. Also included in these experiments were measurements of the neutron and photon spectra with two BICRON BC-501A liquid scintillators. These two detectors were provided and operated by CEA Valduc. They were set up just outside the SILENE reactor cell with additional lead shielding to prevent the detectors from being saturated. The final detectors involved in the experiments were two different types of CAAS detectors. The Babcock International Group provided three CIDAS CAAS detectors, which measured photon dose and dose rate with a Geiger-Mueller tube. CIDAS detectors are currently in use at Y-12 in the newly constructed Highly Enriched Uranium Materials Facility. The second CAAS detector used a 6 LiF TLD to absorb neutrons and a silicon detector to count the charge particles released by these absorption events. Lawrence Livermore National Laboratory provided four of these detectors, which had formerly been used at the Rocky Flats facility in the United States.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Potential Experimental Topics for EGS Collab Experiment 3

To facilitate the success of FORGE, the DOE GTO has initiated a new research effort, the EGS Collab project, which will utilize readily accessible underground facilities that can refine our understanding of rock mass response to stimulation and provide a test bed at intermediate (~10 m) scale for the validation of thermal-hydrological-mechanical-chemical modeling approaches as well as novel monitoring tools. The first two EGS Experiments 1 and 2 are planned be performed under different stress/fracture conditions, and will evaluate different stimulation processes: Experiment 1 will focus on hydrofracturing of a competent rock mass, while Experiment 2 will concentrate on hydroshearing of a rock mass that contains natural fractures. Experiment 3 is scheduled to begin in 2019 will build off the lessons learned in Experiments 1 and 2 and will investigate alternate stimulation and operation methods to improve heat extraction in an EGS reservoir. This paper evaluates potential experiments that could potentially be conducted in Experiment 3. The two technical parameters defining energy extracted from EGS reservoirs with the highest economic uncertainty and risk are the production well flow rates and the reservoir thermal drawdown rate. A review of historical and currently on-going EGS studies has identified that over 1/2 of the projects have identified heat extraction challenges during their operation associated with these two parameters as well as some additional secondary issues. At present, no EGS reservoir has continuously produced flow rates on the order of 80 kg/s. Short circuiting (i.e. early thermal breakthrough) has been identified in numerous cases. In addition, working fluid loss (i.e. the difference between the injected fluid mass and the extracted fluid mass as compared to the injected mass) has been as high as 90%. Finally, the engineering aspects of operating a true EGS multi-fracture reservoir such as repairing/modifying fractures and controlling working fluid fluxes within multiple fractures for the effective EGS fracture management has not been sufficiently studied. To examine issues such as these, EGS Collab Experiment 3 may be conducted in the testbeds prepared for Experiments 1 and 2 by improving the previously performed stimulations, or conducted at a new site performing new stimulations with alternate method. Potential experiments may include using different stimulation and working fluids, evaluating different stimulation methods, using proppants to enhance permeability, and other high-risk high-reward methods that can be evaluated at the 10-m scale environment.

15 GEOTHERMAL ENERGY↗

DOE OSTI2018

Potential Experimental Topics for EGS Collab Experiment 3

To facilitate the success of FORGE, the DOE GTO has initiated a new research effort, the EGS Collab project, which will utilize readily accessible underground facilities that can refine our understanding of rock mass response to stimulation and provide a test bed at intermediate (~10 m) scale for the validation of thermal-hydrological-mechanical-chemical modeling approaches as well as novel monitoring tools. The first two EGS Experiments 1 and 2 are planned be performed under different stress/fracture conditions, and will evaluate different stimulation processes: Experiment 1 will focus on hydrofracturing of a competent rock mass, while Experiment 2 will concentrate on hydroshearing of a rock mass that contains natural fractures. Experiment 3 is scheduled to begin in 2019 will build off the lessons learned in Experiments 1 and 2 and will investigate alternate stimulation and operation methods to improve heat extraction in an EGS reservoir. This paper evaluates potential experiments that could potentially be conducted in Experiment 3. The two technical parameters defining energy extracted from EGS reservoirs with the highest economic uncertainty and risk are the production well flow rates and the reservoir thermal drawdown rate. A review of historical and currently on-going EGS studies has identified that over ½ of the projects have identified heat extraction challenges during their operation associated with these two parameters as well as some additional secondary issues. At present, no EGS reservoir has continuously produced flow rates on the order of 80 kg/s. Short circuiting (i.e. early thermal breakthrough) has been identified in numerous cases. In addition, working fluid loss (i.e. the difference between the injected fluid mass and the extracted fluid mass as compared to the injected mass) has been as high as 90%. Finally, the engineering aspects of operating a true EGS multi-fracture reservoir such as repairing/modifying fractures and controlling working fluid fluxes within multiple fractures for the effective EGS fracture management has not been sufficiently studied. To examine issues such as these, EGS Collab Experiment 3 may be conducted in the testbeds prepared for Experiments 1 and 2 by improving the previously performed stimulations, or conducted at a new site performing new stimulations with alternate method. Potential experiments may include using different stimulation and working fluids, evaluating different stimulation methods, using proppants to enhance permeability, and other high-risk high-reward methods that can be evaluated at the 10-m scale environment.

15 GEOTHERMAL ENERGY↗

DOE OSTI2023

Beam dynamics corrections in the measurement of the anomalous precession frequency at the Muon $g-2$ experiment at Fermilab

The Muon $g-2$ experiment at Fermilab (E989) aims to measure the anomalous magnetic moment of the muon with an accuracy of 140 ppb (parts per billions). This accuracy, obtained by adding in quadrature a statistical and a systematic contribution of comparable value (100 ppb), will allow to reduce the experimental uncertainty (from the previous E821 experiment at BNL \cite{bnl}) of a factor of 4, and represents one of the most precise tests of the Standard Model (SM) theory of elementary particles. The first result on the Run-1 dataset \cite{prl} was released on April 7, 2021, showing a very good agreement with the previous result from BNL experiment, with a slightly better uncertainty. The corresponding experimental average increases the significance of the discrepancy between the measured and Standard Model prediction of 4.2$\sigma$ \cite{white_paper}.\\ \noindent The measured quantity is the muon magnetic anomaly a$_\mu$=$\frac{g_\mu-2}{2}$ where g$_\mu$ is the gyromagnetic factor of the muon. Dirac's equation predicts g$_\mu$ = 2, while radiative corrections, dominated by the QED contribution due to an exchange of a virtual photon, causes a per-mille correction on this quantity. By including all the SM contributions, a$_\mu$ is known at 370 ppb. The E989 experiment measures $a_\mu$ injecting positive muons with momentum of 3.1 GeV/$c$ polarized longitudinally in bunches (called $fills$) with an average rate of 12 Hz, in a storage ring of 14 meters diameter. Due to the parity violation in the weak muon decay, high energy positrons produced are emitted preferably in the muon's spin direction. By counting the number of positrons with energy greater then 1.7 GeV in function of the time, the frequency precession of the muon spin is measured, that together with the measurement of the magnetic field, allows to extract $a_\mu$. The positrons are detected with 24 electromagnetic calorimeters, that measure the energy and the arrival time of the positrons, each made of 54 crystals of lead fluoride (PbF$_2$) read by silicon photomultipliers (SiPM). Together with calorimeters, two tracking detectors are used to make non-destructive measurements of the muon beam characteristic by reconstructing the muon decay position extrapolating backward the decay positrons. The knowledge of the beam motion inside the ring plays a fundamental role in the analysis of $a_{\mu}$, where the measured anomalous precession frequency $\omega_a$ must be corrected for four main beam dynamics effects. A first correction is associated with the presence of an electric field responsible of the vertical focusing on the storage ring, where vertical direction is orthogonal to the orbit and horizontal direction is along the storage ring radius. Due to the oscillations in the vertical direction, the so called vertical betatron oscillations, a second correction is necessary to account for an average angle associated with the muon motion off the ideal orbit. A third correction is caused by lost muons in the ring which have a different spin phase at the injection respect to the decay ones. Finally, due to the correlation between the vertical and horizontal beam motion of the muons and the acceptance of the calorimeter, a correction (called ``phase-acceptance") arises. In Run1, due to the presence of two damaged resistors in one electrostatic quadrupole, this correction had a prominent role. The replacement of the damaged resistors before Run2 decreased this effect by one order of magnitude, and a further reduction in Run-3 was provided by an improved orbit.\\ \noindent The work of this Thesis focuses on the beam dynamics corrections on $\omega_a$. Due to the prominent role in Run-1 a special attention has been put to the phase-acceptance correction\footnote{The effect of this correction on the E821 BNL measurement of the $g-2$ has been evaluated to be within the quoted systematic error.}. Chapter \ref{ch:anomalous_magnetic_moment} introduces the anomalous magnetic moment of the muon. Chapter \ref{ch:early_experiment} describes the history of the Muon $g-2$ experiments. Chapter \ref{ch:theory} discusses the Standard Model prediction and possible new physics scenario. Chapter \ref{ch:muon_g2_experiment} describes the E989 experiment, whose experimental technique and the apparatus are discussed focusing on the improvements needed to reach the final goal on $a_\mu$ measurement. The original work of the Thesis is discussed in the last 5 chapters. Chapter \ref{ch:experiment_technique} presents the analysis technique to extract $\omega_a$, Chapter \ref{ch:beam_dynamics} describes the beam dynamics in the E989 experiment, Chapter \ref{ch:beam_dynamics_correction} discusses the beam dynamics correction to $\omega_a$, and Chapter \ref{sec:pa_corr} focuses on the phase acceptance correction. Finally Chapter \ref{ch:run23_analysis_improvements} contains the Run-2/3 improvements which are expected to allow for an increased precision ons the beam dynamics co...

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Fuel Performance Simulation of High Burnup Fuels in Planned Integral Design Basis Accident Experiments

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Rossi-alpha Analysis of CURIE Experiment

Critical assembly measurement and operations are crucial to the development of benchmark data to support research into criticality safety, radiation-detection development, and the overall application of nuclear technologies. Accurate nuclear data are needed for accurate predictive simulations, and validation using critical experiments is an important part of the nuclear data pipeline. The accuracy of nuclear data are improved using more robust and targeted measurements. In particular, the intermediate energy range of uranium is of great interest. LANL has successfully performed the Zeus series of experiments on the Comet assembly to investigate the intermediate energy range for HEU with various moderators. A successor to these experiments is the benchmark for the Critical Unresolved Region Integral Experiment (CURIE), designed to be sensitive to the unresolved resonance region (URR). This experiment is designed using polytetrafluoroethylene, more commonly known as Teflon, moderators. The CURIE experiment was successfully conducted at the National Criticality Experiments Research Center (NCERC). The Rossi-alpha method was used to evaluate the propensity of the CURIE configurations to sustain fission chains by estimating the prompt neutron decay constant α. This work evaluates the α at delayed critical using Rossi-alpha for different Teflon moderator thicknesses in the CURIE experiment. These results will help improve understanding of the CURIE benchmark experiments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

The CWS Experiments – An Experimental Study of the Effects of Chlorine on Thermal Neutron Absorption

The Chlorine Worth Study (CWS) experiments consist of layers of Zero Power Physics Reactor plates moderated by high density polyethylene, polyvinyl chloride, and chlorinated polyvinyl chloride with the goal to produce a thermal critical experiment. A thick high density polyethylene reflector surrounds the experiments. Aluminum fuel trays and frames were used to conduct the intrinsic heat generation of the plutonium plates away from the core and out to the top plate and the platen, both acting as heat sinks. The goal of the CWS experiments was to provide benchmark critical experiments that were sensitive to chlorine and matched the sensitivity of criticality safety applications for aqueous chloride operations at PF-4 at Los Alamos National Laboratory. The specific aqueous chloride PF-4 applications were solutions of 30 g/L plutonium, 300 g/L plutonium, and 600 g/L plutonium. The CWS experiments were designed using prototype methods developed by the ARCHIMEDES and EUCLID projects at Los Alamos National Laboratory. The CWS experiments were performed in December of 2021 at the National Criticality Experiments Research Center at the Nevada Nuclear Security Site on the Planet critical assembly. It is intended that the CWS experiments will be evaluated and submitted to the ICSBEP in the future as criticality safety benchmarks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

TEX-HEU: Integral Experiments with Highly Enriched Uranium and Polyethylene (IER-297 CED-3b Report)

This report documents the experimental critical configurations and measurements for IER-297, Thermal/Epithermal eXperiments (TEX) using highly enriched uranium (HEU) fuel, moderated and reflected by polyethylene. These experiments establish the baseline TEX-HEU configurations spanning thermal, intermediate, and fast fission energy regimes. The experiment campaign was completed in FY2020 at the National Critical Experiments Research Center at the Nevada National Security Site. The campaign produced six documented critical configurations (0”, ¹/₈” A, ¹/₈” B, ¹/₄”, ¹/₂”, and 1¹/₂”) and one reproducibility measurement (¹/₈” R). Table 1 summarizes these seven experiments along with their period and calculated reactivity, in order of execution. The Comet General Purpose Critical Assembly Machine was used to conduct the TEX-HEU experiments. The HEU fuel consists of U(93+) plates, collectively known as the “Jemima” plates. Each experiment consisted of an alternating stack of these HEU plates with or without polyethylene moderator plates. By varying the thicknesses of the polyethylene plates, the neutron spectrum of the experiment could be fine-tuned to a specific fission energy regime. The experiments were reflected by 1” of polyethylene consisting of a bottom and top plate and outer reflector rings to surround the HEU and polyethylene plates. The fine reactivity control of the system was provided by the thickness of the top reflector plate, which could be varied in increments of ¹/₃₂”.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Design of Experiments Involving Molybdenum in SNL SPRF/CX Facility

This report is part of the collaboration with the Nuclear Criticality Safety Program and constitutes the “CED-2 report” of the experiments proposed in the SPRF/CX facility of SANDIA National Laboratories to contribute to the validation of Molybdenum in thermal energy spectrum. It supplements the study carried out in the CED-1 report by providing a more precise design of the proposed experiments and presents a complete evaluation of the experimental uncertainties. In the CED-1 report, corresponding to the preliminary design of the experiments, IRSN identified several promising configurations that could be implemented in the SPRF/CX facility of SNL. They use metallic molybdenum either in the form of foils between the UO 2 pellets inside BUCCX rods, or in the form of sleeves surrounding 7uPCX rods. In this CED-2 report, a new design using molybdenum sleeves surrounding BUCCX rods is also analyzed and the experimental uncertainties of all the proposed configurations are estimated. Given the constraints to carry out the experiments and the costs associated with the manufacturing of the foils, the experiments using molybdenum foils are discarded. Similarly, the design using molybdenum sleeves around BUCCX rods leads to significant manufacturing costs which could certainly be partially offset by use of existing grids but which would lead to slightly lower values of the integral sensitivities to the cross sections of molybdenum. Consequently, the design using molybdenum sleeves around 7uPCX rods is favored insofar as it presents higher $k_{eff}$ sensitivities to the molybdenum cross sections and that there is still possibility to reduce the number of sleeves to reduce the cost of experiments. Regarding feedback on nuclear data in support of criticality safety assessment, one of the advantages of the work described here is that the proposed experiments are not correlated with the MIRTE and FP experiments using molybdenum. In addition, they lead to higher sensitivity values in the thermal energy range than the existing experiments and partly “cover” the first resonance of the 95 Mo, which is a real advantage for improving knowledge on molybdenum cross sections.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

The PARADIGM Project: Case Study in Balancing Experiment Uncertainty with Design simplicity

Accurate nuclear data are required for simulations of many applications including nuclear criticality safety. Actinide nuclear data at intermediate energies (from 1 to 100s of keV) are imprecise and inaccurate, because of scarce differential data, and an insufficient theory approach to capture the structures expected in the data to yield evaluated nuclear data, and lack of integral data for proper validation. This is a known deficiency but has proved challenging to address. More specifically, only 5% of integral experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) benchmark suite address intermediate energies (Fig. 1). Associated calculated effective multiplication factor, k eff , values for these experiments are far outside the experimental uncertainties and are 25× further from experiment than for fast energies. These differences could either stem from systematic biases in nuclear data, experiments or both. The goal of the PARADIGM (PARallel Approach of Differential and InteGral Measurements) project is to significantly reduce (by more than tens of percent) the uncertainties of intermediate energy actinide nuclear data. The PARADIGM project designed and intends to execute LANSCE (Los Alamos Neutron Science CEnter) and NCERC (National Criticality Experiments Research Center) intermediate experiments in parallel. They will specifically address a high priority nuclear data need—reducing bias and uncertainty in intermediate plutonium nuclear data. The two experiment will achieve that by informing each other and nuclear theory. By doing all these steps in parallel, the timeline to deliver improved nuclear data to users will significantly be reduced. This work will focus on the integral experiment final design and the balance of design and modeling simplicity while minimizing experiment uncertainty.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O 2 fueled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Rossi-alpha Analysis of Thermal/Epithermal eXperiments Optimized for Polyethylene Thermal Neutron Scattering

Accurate nuclear data is the foundation for predictive simulations and design of new experimental in the nuclear community. The criticality safety community has particular interest in benchmarking assemblies with thermal neutron data. To address such needs the Nuclear Criticality Safety Program (NCSP) funded the Thermal/Epithermal eXperiments (TEX) campaigns that were to be completed between Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). Specifically, analysis of plutonium with a thermal neutron spectrum expands on previous work to help validate thermal scattering law data, which can have larger impacts in thermal applications. The first of these experiments were successfully conducted in 2018, but this paper will focus on the 2021 measurement focused on investigating the the thermal scattering law (TSL) for polyethylene. These experiments were successfully conducted at the National Criticality Experiments Research Center (NCERC) using the Planet vertical lift critical assembly machine. Polethylene plates were layered with trays of Zero Power Physics Reactor (ZPPR) 24 plates in a 12" by 12" square. The ZPPR plates were 2" by 3" by 0.125" bearing weapons grade plutonium. The polyethylene moderator was either 2" or 1.6875" thick. This work builds on the Rossi-alpha calculations done by McKenzie et al. for the same detector-assembly system and will only focus on the Rossi-alpha neutron noise method. This work will aim to further validate the results of the experiment through a novel neutron noise python package. Following similar methodology to the previous analysis, analysis of TEX evaluated the prompt neutron decay constant at delayed critical, $α_{DC}$ using Rossi-alpha for different polyethylene moderator thicknesses. These results will help improve understanding of TSL in critical experiments. Alpha (α), is the prompt neutron decay constant of the measured system and allows for the evaluation of a systems propensity to sustain fission chains via prompt neutrons. The single value description of the assemblies allow for comparison between experiments regardless of composition, geometry, and reflectors/moderators. Rossi-alpha measurements were performed on the polyethylene moderated TEX experiments to estimate $α_{DC}$.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2013

Investigating the experiences of New Zealand MRI technologists: Exploring intra-orbital metallic foreign body safety practices

Qualitative research is lacking regarding the experiences of magnetic resonance imaging (MRI) technologists and their involvement in workplace safety practices. This article provides a gateway to explore, describe and document experiences of MRI technologists in New Zealand (NZ) pertaining to intra-orbital metallic foreign body (IMFB) safety practices. This phenomenological study describes the experiences of seven MRI technologists all with a minimum of 5 years' NZ work experience in MRI. The MRI technologists were interviewed face-to-face regarding their professional IMFB workplace experiences in order to explore historical, current and potential issues. Findings demonstrated that aspects of organization and administration are fundamentally important to MRI technologists. Varying levels of education and knowledge, as well as experience and skills gained, have significantly impacted on MRI technologists’ level of confidence and control in IMFB practices. Participants’ descriptions of their experiences in practice regarding decision-making capabilities further highlight the complexity of these themes. A model was developed to demonstrate the interrelated nature of the themes and the complexity of the situation in totality. Findings of this study have provided insight into the experiences of MRI technologists pertaining to IMFB safety practices and highlighted inconsistencies. It is hoped that these findings will contribute to and improve the level of understanding of MRI technologists and the practices and protocols involved in IMFB safety screening. The scarcity of available literature regarding IMFB safety practices highlights that more research is required to investigate additional aspects that could improve MRI technologists’ experiences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

On the flow induced vibration of an externally excited nuclear reactor experiment

An in-pile, drop-in experiment design is presently being designed and studied for the near-term deployment within the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL); this experiment is termed the Miniplate-1 Large-B (MP-1 LB) Experiment. A number of explicit studies are performed during the design- and safety-related stage. Traditionally, a clear and logical methodology has been developed and utilized for analyses such as hydraulics, thermal-loads, mechanical loads, and others for these experiments. Recently a small component from a different experiment assembly mechanically separated while in the reactor’s core. While this experiment didn’t compromise the safety of the reactor, it led to a higher-level question which centered on whether the appropriate level of consideration was being made toward the fluid–structure-interactions of these experiments. The outcome yielded separate flow test experiments of like-for-like geometry in an experimental loop located at Oregon State University which produces experimental data compliant with applicable parts and requirements to ASME’s NQA-1 2008, 2009a standard – suitable for benchmark evaluation. Here, the objectives of this study are to (1) develop a process for handling and interpreting the mechanical response of the test elements during hydraulic testing, to (2) characterize the motion of a specific test element during a flow test which imposes a wide range of hydraulic conditions, and (3) provide objective observations toward the potential safety related implications that are tied to the synthesized data. The outcome of this study has led to a confident process in inspecting the experimental data, synthesized it for interpretation, identified several unique hydraulic characteristics of the experiment design which were previously unknown, and demonstrated that the likelihood for mechanical failure resulting from fluid-structure-interactions in the reactor is far below any criterion for concern of the element’s safety.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Impact of recent ENDF nuclear data update, high initial enrichment and high burnup fuel on critical experiments applicability determination via the integral index c k for burnup credit validation

In 2012, NUREG/CR-7109 reported on the validation of burnup credit calculations involving major and minor actinides and major fission products which was investigated for pressurized and boiling water reactor (PWR and BWR) fuel enrichments up to 5 wt% 235 U and assembly-average burnups up to 60 GWd/MTU. Recently, there has been interest in increasing the maximum enrichment used in PWR fuel as high as 8 wt% 235 U and correspondingly increasing the maximum assembly-average burnups to approximately 75 GWd/MTU. These proposed increases in enrichment and burnup necessitate reinvestigation of the validation basis for k eff calculations for this expanded application space. Additionally, the 2012 study was performed by using the Evaluated Nuclear Data File (ENDF)/B-VII.0 nuclear data with the SCALE 6 covariance library, and the effects of using the newly released ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data and covariance libraries should be evaluated. In this work, published in NUREG/CR-7309 in 2025, the validation assessment was performed consistently with NUREG/CR-7109: modeling irradiated fuel assemblies in the Generic Burnup Credit (GBC)-32 cask defined in NUREG/CR-6747. The TSUNAMI-3D sequence was used to generate sensitivity data for the application model, and the data were compared with sensitivity data from select benchmark models. The integral parameter c k is the metric of similarity used in this study and is consistent with NUREG/CR-7109, where a c k value in excess of 0.8 indicates sufficient similarity for use in validation. A new set of benchmark experiments with sensitivity data has been assembled for this effort. The number of experiments with available sensitivity data is now 2,104, compared to 474 in NUREG/CR-7109. This increase was facilitated by the efforts of the Nuclear Energy Agency to generate sensitivity data for a majority of the experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook to supplement the data available in the Oak Ridge National Laboratory (ORNL) Verified, Archived Library of Inputs and Data (VALID). The complete set of benchmarks considered here includes experiments for low-enriched uranium (LEU), intermediate enriched uranium (IEU), and a mixture of uranium and plutonium (MIX) from the ICSBEP Handbook and VALID, as well as ORNL models of the Haut Taux de Combustion (HTC) experiments and other potentially relevant models not included in VALID. The updated similarity study shows that none of the extended burnup and higher enrichment combinations considered show a significant decrease in the number of potentially applicable experiments, meaning sufficient critical experiments exist for the validation of BUC criticality safety calculations, with initial enrichments up to 8 wt% 235 U and burnups up to 80 GWd/MTU. Additionally, both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries can be used for validation since the number of critical experiments applicable for validation increases for most cases with the most recent nuclear data compared to the previous one. As in previous BUC validation studies, the French HTC experiments are the most similar in a majority of the application cases studied, especially from representative discharge burnups ranging from 40 to 80 GWd/MTU. In conclusion, these results match the conclusions presented in NUREG/CR-7109 regarding validation of the primary actinides in BUC analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗