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DOE OSTI2013

ATRC Neutron Detector Testing Quick Look Report

As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activation spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutron detectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered Neutron Detectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Enhancement of x-rays generated by a guided laser wakefield accelerator inside capillary tubes

Electrons accelerated in the nonlinear regime in a laser wakefield accelerator experience transverse oscillations inside the plasma cavity, giving rise to ultra-short pulsed x-rays, also called the betatron radiation. We show that the fluence of x-ray can be enhanced by more than one order of magnitude when the laser is guided by a 10 mm long capillary tube instead of interacting with a 2 mm gas jet. X-rays with a synchrotron-like spectrum and associated critical energy {approx}5 keV, with a peak brightness of {approx}1x10{sup 21} ph/s/mm{sup 2}/mrad{sup 2}/0.1%BW, were achieved by employing 16 TW laser pulses.

43 PARTICLE ACCELERATORS↗

DOE OSTI2019

Dynamic Frictional Characteristics of TP2 Copper Tubes during Hydroforming under Different Loading and Fluid Velocities

Tube hydroforming (THF) experiments were performed on TP2 copper tubes under different loading velocities and fluid velocities using a self-developed measurement system to investigate dynamic frictional characteristics in the guiding zone. The results show that the coefficient of friction (COF) dynamically changes during forming experiments and decreases with tube deformation. The average descending rate and amplitude of the COF increase with increasing loading velocity. Microscopically, the micro-protrusions on the tubular surface are flattened, and the surface scratches are finer and more uniform, as the loading velocity increases, resulting in a decrease in COF. At the same external loading velocity, the COF increases with increasing fluid velocity and is also extremely sensitive to it. Moreover, improving and predicting the formability of such tubes by accurately adjusting and controlling fluid velocity in THF is valuable and critical for the future.

36 MATERIALS SCIENCE↗

DOE OSTI2015

An integrated equipment for massive segmentation and packaging of control Rod Guide Tubes - 15161

Over time, some utilities have decided to undergo safety upgrades that include replacement of some ageing components like control rod guide tubes (GT's). GT's are safety related components because they control the path of the control rods into and out of the core. These components are usually stored underwater in interim storage pools to decay their radioactivity. When the interim storage facility reaches saturation, treatment and final disposal of the retired Guide Tubes needs to be performed. Based on more than 30 years of experience and lessons learned in the development of segmentation and packaging technologies for nuclear components, Westinghouse has designed, commissioned and will operate a dedicated system for cutting and packaging retired GT's from 900 MW plants, currently stored in a EDF maintenance facility located in the south of France. The first main technical challenge was the waste volume optimization. Two separate waste streams had to be considered for coping with the highly irradiated portion of the GT's (lower part) and the very low level waste sections (upper part). The segmentation plan has been designed to minimize the waste volume of the highly irradiated waste and therefore develop a process which could be adapted to the container size specified by EDF. Metallic baskets have been designed to optimize the packaging and storage density in the containers. All cutting techniques previously used by Westinghouse in similar applications were assessed and underwater mechanical cutting with a band saw was selected as the preferred method. A turning stand positions the GT in front of the band saw. When the cut is completed, the remnant GT is inserted into the next turning stand for the next cut. Each cell of the turning stand positions the GT at the pre-defined cutting height. An integrated water filtration system is used to ensure good visibility during the cutting operations and to minimize water volume and limit the final liquid effluents. The whole process of GT insertion, cutting, packaging and waste conditioning is carried out in a hot cell and controlled remotely. The operator can therefore follow the packaging plan accurately and safely by using a four axis pick and place unit, while keeping personal radiation exposure low. This innovative process allows cutting and packaging large quantities of GT's in a safe and efficient manner. The pilot project covers the segmentation of 93 GT's but it can be further used for cutting larger quantities of retired GT's resulting from the upcoming GT replacement campaigns. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Characterization of electroless nickel-phosphorus plating for ultracold-neutron storage

Electroless nickel plating is an established industrial process that provides a robust and relatively low-cost coating suitable for transporting and storing ultracold neutrons (UCN). Using roughness measurements and UCN-storage experiments we characterized UCN guides made from polished aluminum or stainless-steel tubes plated by several vendors. All electroless nickel platings were similarly suited for UCN storage with an average loss probability per wall bounce of 2.8 • 10 -4 to 4.1 • 10 -4 for energies between 90 neV and 190 neV, or a ratio of imaginary to real Fermi potential η of 1.7 • 10 -4 to 3.3 • 10 -4 . Measurements at different elevations indicate that the energy dependence of UCN losses is well described by the imaginary Fermi potential. Some special considerations are required to avoid an increase in surface roughness during the plating process and hence a reduction in UCN transmission. Increased roughness had only a minor impact on storage properties. In conclusion, based on these findings we chose a vendor to plate the UCN-production vessel that will contain the superfluid-helium converter for the new TRIUMF UltraCold Advanced Neutron (TUCAN) source, achieving acceptable UCN-storage properties with η = 3.5(5)•10 -4 .

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

Retractable Sensor Poster for ANIMMA 2023 Conference

42 ENGINEERING↗

DOE OSTI2020

BWR Spent Nuclear Fuel Acquisition and Testing to Support DOE-NE High Burnup Spent Fuel Data Project

The Office of Spent Fuel and Waste Disposition (SFWD) within the US Department of Energy (DOE) Office of Nuclear Energy (NE) established the Spent Fuel and Waste Science and Technology (SFWST) campaign to conduct research and development (R&D) activities related to the storage, transportation, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste. The SFWST program was created within SFWD to address issues of extended or long-term SNF storage and transportation. Some near-term objectives of SFWST are to use a science-based, engineering-driven approach to: Support the enhancement of the technical bases to support the continued safe and secure dry storage of SNF for extended periods; Support the enhancement of the technical bases for retrieving SNF after extended dry storage; Support the enhancement of the technical bases for transporting high burnup (HBU) fuel and transporting low burnup fuel and HBU fuel after dry storage DOE-NE, in partnership with the Electric Power Research Institute, developed the High Burnup Spent Fuel Data Project to perform a large-scale demonstration and laboratory-scale testing of HBU pressurized water reactor (PWR) fuels (exceeding 45 gigawatt-days per metric ton of uranium [GWd/MTU]). Under this project, 25 sister rods—which are rods that have the same design, power histories, and other characteristics—were removed from assemblies at the North Anna Nuclear Power Station and sent to Oak Ridge National Laboratory (ORNL) in January 2016. ORNL performed detailed nondestructive examination (NDE) on all 25 rods. The NDE consisted of visual examinations, gamma and neutron scanning, profilometry and rod length measurements, and eddy current examinations. After completing the NDE, 10 of the sister rods were delivered to Pacific Northwest National Laboratory (PNNL) in a NAC International, Inc. legal-weight truck cask in September 2018 for destructive examination (DE). To date, SFWD work has focused on the PWR fuel that is part of the Sister Rod Test program. No boiling water reactor (BWR) fuel has been tested in the program, and the data needs that were identified for the PWR fuel have not been collected for BWR fuel. The goal to obtain six to nine BWR rods and test them at ORNL will support closing this important data gap. BWR fuel comprises approximately 56% of the total fuel assemblies currently in storage at nuclear power plants in the United States. BWR nuclear fuel and cladding designs and manufacturing are significantly different from PWRs. Differences include the following: BWR fuel pellets are larger than PWR pellets; Variations of Zircaloy-2 (including liners) are used instead of the Zircaloy-4 cladding materials used in PWRs; Clad manufacturing and stress-relief processes are different between PWRs and BWRs; The fuel rod dimensions are different because larger rod diameters and thicker cladding are used in BWRs; BWR fuel typically has lower internal rod pressures and sees vastly different operating conditions than PWR fuel (i.e., two-phase flow); BWR assemblies are “canned,” meaning each assembly is surrounded by a metal fuel channel; BWR cladding is often composed of an inner pure Zr liner that has widely different mechanical properties than the Zircaloy-2 alloy and exhibits a stronger affinity for hydrogen; The BWR SNF generally has more total hydrogen in the cladding/liner than typical PWR fuel; The construction of the PWR and BWR assemblies is vastly different; BWR rods are solidly attached to the assembly nozzles and experience a much different vibration and shock load than PWR rods, which are “floating” within a grid system attached to guide tubes, and the rods sit loosely on the bottom end plates. These numerous differences will affect the way the BWR SNF responds under dry storage preparation processes (e.g., vacuum drying) and during transportation. The results collected in the PWR experimental program must be compared with a subset of similar data collected on BWR SNF to establish a technical basis for whether the larger PWR database is sufficient to bound the BWR SNF end-of-life conditions as is currently assumed for several fuel/clad properties. Changes that occur in both fuel types at HBU could exacerbate any mechanical property differences. As the fuel burnup increases, several changes occur that might affect the performance of the fuel, cladding, and assembly hardware in storage and transportation. These changes include increased cladding corrosion layer thickness, increased cladding hydrogen content, increased cladding creep strains, increased fission gas release, and the formation of the HBU structure at the surface of the fuel pellets. The Nuclear Regulatory Commission (NRC) limits the current maximum rod-averaged burnup to 62 GWd/MTU due to these changes and the lack of data at higher burnups.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Dynamic modeling of heat pipe integrated thermal battery latent heat storage system experiment validation

A heat pipe integrated thermal battery system has been constructed to investigate a high-temperature latent heat thermal energy storage technology that takes advantage of near isothermal operation of latent heat storage and heat pipes to potentially enable high-energy isothermal heat storage. A dynamic model constructed in Modelica has been validated, showing errors between 2.5 °C–39.7 °C across 10-h to 47-h simulations against experiment results, showing good prediction capability of experiment output, especially against phase change time. Model calibrations showing vessel heat-up capability of 3 kW and heat pipes combining to provide 600 W each during experiment operation validate experiment circumstances including reduced material loading and reduced power capability. The experiment configuration uses an Al-Mg-Zn eutectic metal as the storage material, heated via heat tape wrapped around the vessel and guide tubes to bring the system to operation range (>400 °C) and to simulate charging heat exchange, respectively, with heat rejection occurring through the surfaces of the material and facilitated via guide tubes with less insulation wrapping. The model is available in the open-source repository HYBRID on Github.

25 - ENERGY STORAGE↗

DOE OSTI2018

Pellet-Injector Technology—Brief History and Key Developments in the Last 25 Years

High-speed injection of solid fuel was first proposed in 1954 as a possible solution to the problem of transporting fresh fuel across the confining magnetic fields into the plasma of a fusion reactor. While it took a few decades, the use of cryogenic pellets (typically H 2 and D 2 ) on fusion experiments became common place; most tokamaks and stellarators are now equipped with a pellet injector(s). These devices operate at low temperatures (~10 to 20 K) and most often use a simple light gas gun to accelerate macroscopic-size pellets (~0.4- to 6-mm diameter) to speeds of ~100 to 1000 m/s. Before the advantages of pellet injection from the magnetic high-field side (HFS) of a tokamak were recognized in 1997, development focused on increasing the pellet speed to achieve deeper plasma penetration and higher fueling efficiency. The HFS injection technique typically dictates slower pellets (~100 to 300 m/s) to survive transport through the curved guide tubes that route the pellets to the plasma from the inside wall of the device. Two other key operating parameters for plasma fueling are the pellet-injection repetition rate and time duration—a single pellet is adequate for some experiments and a steady-state injection rate of up to ~50 Hz is appropriate for others. In addition to plasma fueling, cryogenic pellets have often been used for particle transport and impurity studies in fusion experiments (most often with neon pellets). During the past two decades, a few new applications for cryogenic pellets have been developed and used successfully in plasma experiments: (1) one for edge-localized mode mitigation, (2) one for plasma disruption mitigation (requires large pellets that are shattered before injection into the plasma), and (3) another in which pure argon pellets are used to trigger runaway electrons in the plasma for scientific studies. Here in this paper, a brief history and the key developments in this technology during the past 25 years are presented and discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

APR1400 In-Core Instrumentation Penetration Failure by Zirconia Melt

Light water reactors have many penetrations at the reactor vessel lower head which are called ICI (in-core instrumentation tubes), ICM-GT (in-core monitoring guide tubes), CRGT (control rod guide tubes) and drain tubes. For Korean PWRs (pressurized water reactors), there are dozens of bottom-up type ICI penetration tubes (45 for OPR1000 and 61 for APR1400). For the typical BWRs (boiling water reactors), hundreds of penetrations are positioned at the reactor vessel lower head due to the structural complexity at the reactor upper head. These penetrations are regarded as the most vulnerable structures during a severe accident as they could be failed by high temperature of melt (or debris) and its decay heat. That is, tube rupture and/or tube ejection could occur because there exist an empty hole inside the penetration tube which can be a melt flow path, and they are attached to the inside of the reactor vessel by a partial weld with a gap between the tubes and reactor vessel. Therefore, the research on the penetration failure is of importance, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. This paper focuses on the experimental investigation on the tube ejection failure for the APR1400 ICI penetration. The tube ejection experiments were performed as it was exposed to high temperature of zirconia melt and high pressure. The tube ejection failure experiments were performed for the APR1400 in-core instrumentation penetration nozzle. Zirconia melt was used as a simulant of molten corium and interacted with the penetration specimen. The specimen was pressurized up to 2.6 bar during the interaction with melt to induce the tube ejection. The penetration weld was heated up to its melting temperature, and the penetration tube and weld above the reactor vessel surface were eroded severely by the melt. However, even though it was estimated that the weld failure occurred and the reactor vessel with stainless steel cladding was ablated severely by the melt, the tube ejection did not take place, and consequently the integrity for the APR1400 ICI penetration was confirmed in the present experimental conditions. The goal of this research is to improve the penetrations failure models in MAAP5 codes based on the experimental data. To achieve this goal, it is necessary to increase the heat flux and pressure acting on the penetration tube to find the conditions for the tube ejection failure. Moreover, the penetration failure experiments will be performed by delivering melt into the penetration specimen to simulate the melt relocation process in a real severe accident situation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Preliminary critical experiments validation of a lattice physic code LATC - 14415

Core and System Integrated Engine for design and analysis (COSINE) is an integrated nuclear engineering code package, which is being developed by State Nuclear Power Software Development Center (SNPSDC) in China. A lattice physics code named LATC is included in COSINE.LATC is a 2D, steady-state code for modeling PWR fuel. A WIMS-D 69 group micro library based on ENDF/B-VI is used and there are two transportation methods in LATC, such as two steps method (collision probability method for cell homogenization, discrete coordinate method for assembly homogenization) and one step method (MOC). The UO{sub 2} fuel assembly, which may contain guide tubes, burnable poisons such as Integral Fuel Burnable Absorbers (IFBA) / Wet Annular Burnable Absorbers (WABA)/ Pyrexes or control rods etc, can be calculated in LATC. LATC provides the required cross-section data for the core physics simulator analysis code in COSINE. In order to evaluate the quality of the LATC, a strategy of validation is researched and applied to the LATC. Firstly, a validation requirement analysis is performed, and the corresponding critical experiments are collected, and then a critical experimental matrix considering the variety of fuel enrichments, fuel pitch and water to uranium ratio is generated. Finally, a comparison of results between LATC using collision probability and discrete coordinate method and critical experiments are made to make sure that the code works well. According to the validation strategy discussed above, the preliminary validation is carried out and critical experiments are calculated according to the critical experimental matrix and results from LATC code are compared with reference results. The result shows that the calculated results agree well with critical experiments, and do not indicate any significant trends with respect to the various conditions considered. (authors)

42 ENGINEERING↗

DOE OSTI2017

Boron-Based Nanostructures, Stability, Functionality and Synthetic Routes

Boron (B) is one of the most intriguing elements not only because of its position between metals and nonmetals in periodic table but also because of its ability to form an enormous number of allotropes. Apart from several bulk three-dimensional (3D) phases, boron can form 0D clusters, 1D nanotubes and nanowires, and 2D layers. In particular, boron sheets of monoatomic thickness have raised interest as a potential new 2D-material and as a (conceptual) precursor, for example, so-called α-sheets, from which other boron structures - fullerene cages and tubes - might be constructed. In fact, a number of planar B clusters up to tens of atoms, found in experiments, appear as seeds for extended sheets. In this project we developed theoretical methods to guide synthesis, have successfully identified the material substrates (Ag, Au, Cu) to producing the pure boron layers, and further even predicted what atomistic structures should be expected. These guidelines have successfully led to discoveries in several labs and now have grown into an active line of research worldwide.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Muon Beam Dynamics and Spin Dynamics in the g-2 Storage Ring

The goal of the new g-2 experiment at fermilab is a measurement of the anomalous magnetic moment of the muon, with uncertainty of less than 140 ppb. The experimental method is to store a beam of polarized muons in a storage ring with pure vertical dipole field and electrostatic focusing, and to measure the precession frequency. Control of the systematics depends on unprecedented knowledge of the details of the phase space of the muon distribution. That knowledge is derived from direct measurements with scintillating fiber detectors that are inserted into the muon beam for diagnostic measurements, traceback straw tube tracking chambers, as well as the calorimeters that measure energy, time and position of the decay positrons. The interpretation of the measurements depends on a detailed model of the storage ring guide field. This invited talk presents results of studies of the distribution from the commissioning run of the experiment.

43 PARTICLE ACCELERATORS↗

DOE OSTI2013

Alternatives for Helium-3 in Multiplicity Counters

Alternatives to helium-3 are being actively pursued due to the shortage and rising costs of helium-3. For safeguards applications, there are a number of ongoing investigations to find alternatives that provide the same capability in a cost-effective manner. One of the greatest challenges is to find a comparable alternative for multiplicity counters, since they require high efficiency and short collection or die-away times. Work has been progressing on investigating three commercially available alternatives for high efficiency multiplicity counters: boron trifluoride (BF3) filled proportional tubes, boron-lined proportional tubes, and lithium fluoride with zinc sulfide coated light guides. The baseline multiplicity counter used for the investigation is the Epithermal Neutron Multiplicity Counter with 121 helium-3 filled tubes at 10 atmosphere pressure, which is a significant capability to match. The primary tool for the investigation has been modeling and simulation using the Monte Carlo N-Particle eXtended (MCNPX) radiation transport program, with experiments to validate the models. To directly calculate the coincidence rates in boron-lined (and possibly other) detectors, the MCNPX code has been enhanced to allow the existing coincidence tally to be used with energy deposition rather than neutron capture reactions. This allows boron-lined detectors to be modeled more accurately. Variations of tube number and diameter along with variations in the amount of inter-tube moderator have been conducted for the BF3 and boron-lined cases. Tube pressure was investigated for BF3, up to two atmospheres, as well as optimal boron thickness in the boron-lined tubes. The lithium fluoride was modeled as sheets of material with light guides in between, and the number and thickness of the sheets investigated. The amount of light guide, which in this case doubles as a moderator, was also optimized. The results of these modeling and simulation optimization investigations are described and results presented.

Ely, James H.↗

DOE OSTI2023

Retractable Sensors for In-Core Use in Material Test Reactors - conf paper

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 - 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on friction drive wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 - 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology's MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

Retractable Sensors for In-Core Service in Material Test Reactors

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 - 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on counter-rotating wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 - 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology's MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

Retractable Sensors for In-Core Service in Material Test Reactors

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 – 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on counter-rotating wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 – 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology’s MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

Three-dimensional simulations of low foot and high foot implosion experiments on the National Ignition Facility

In order to achieve the several hundred Gbar stagnation pressures necessary for inertial confinement fusion ignition, implosion experiments on the National Ignition Facility (NIF) [E. I. Moses et al., Phys. Plasmas 16, 041006 (2009)] require the compression of deuterium-tritium fuel layers by a convergence ratio as high as forty. Such high convergence implosions are subject to degradation by a range of perturbations, including the growth of small-scale defects due to hydrodynamic instabilities, as well as longer scale modulations due to radiation flux asymmetries in the enclosing hohlraum. Due to the broad range of scales involved, and also the genuinely three-dimensional (3D) character of the flow, accurately modeling NIF implosions remains at the edge of current simulation capabilities. This paper describes the current state of progress of 3D capsule-only simulations of NIF implosions aimed at accurately describing the performance of specific NIF experiments. Current simulations include the effects of hohlraum radiation asymmetries, capsule surface defects, the capsule support tent and fill tube, and use a grid resolution shown to be converged in companion two-dimensional simulations. The results of detailed simulations of low foot implosions from the National Ignition Campaign are contrasted against results for more recent high foot implosions. While the simulations suggest that low foot performance was dominated by ablation front instability growth, especially the defect seeded by the capsule support tent, high foot implosions appear to be dominated by hohlraum flux asymmetries, although the support tent still plays a significant role. For both implosion types, the simulations show reasonable, though not perfect, agreement with the data and suggest that a reliable predictive capability is developing to guide future implosions toward ignition.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗