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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Characterizing The Mechanical Properties of Polyurethane Foams

The purpose is to determine the mechanical properties of polyurethane foams with the intent of using PU foams to fill large void asymmetrical volumes like a glovebox to fixate transferable radiological material. Testing the resiliency of foam technologies when exposed to physicals stressors will be done. Six COTS PU foams considered will be fabricated and tested (iT 7, iT 8, iT 14, iT 23, 3M Fire Barrier, and Hilti) with and without additives at 12.5% wt (Bi, WO{sub 3}, and Na{sub 2}WO{sub 3}). Polyurethanes (PU) are a special class of polymeric materials that have great potential in many applications. Used in automotive, biomedical, construction, and textiles industries. Sought after due to its versatility and superior properties. PU foams have become one of the most prominent PU-based products. Bench-scale testing by SRNL and FIU-ARC confirmed PU foam technology's superior capabilities: Cured in various and desired volumes; No samples generated heat above 90 deg. C during curing reaction; Incorporation of additives (Bi, WO{sub 3}, BiO{sub 3}, Na{sub 2}WO{sub 3}) led to the following shielding: 98% of Am-241, 16% of Cs-137, 9.5% of Co-60. Proved foams' resistance to thermal stressors: Foams lasted 2 hours of direct flame; Excellent thermal insulation; Maintained structural integrity; No smoke or flame propagation. Experiment 1: A tensile tester (MTS Criterion Series 43) will be utilize to evaluate various properties of different PU rigid and flexible foams with and without additives (at 12.5% wt) under tension at 1 mm/s. Experiment 2: A thermogravimetric analysis (TGA) and Dynamic Mechanical Analyzer (DMA) will be employed to determine PU foams' breakdown of physical properties while under a constant heat rate (10 C/min) and strain rate (1%) under a period of time. Foams with additives showed an increase in stiffness and durability during tensile testing with WO{sub 3} producing the best results for Foam iT 7. All the other foam samples showed similar trends. Glass transition temperature (Tg) range for Foam iT 7 with Na{sub 2}WO{sub 3} ranged from 120-150 C while Foam iT 7 blank Tg range was from 90-110 C with a difference of about 30 C showing improvement with addition of additives.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Applications of Transient Grating Spectroscopy to Radiation Materials Science

Radiation damage is an inherently multi-scale phenomenon, with much of the key material evolution occurring at the mesoscale. Current techniques for the direct interrogation of material property changes due to irradiation either function with very small volumes, such as in-situ TEM, cannot cover the full range of defect populations produced, such as in positron annihilation spectroscopy, or requires destruction of the test material, such as in standard or micromechanical testing. In addition, the ultimate quantities of interest to radiation materials scientists are the changes in key material properties with irradiation. These include stiffness, ductility, and thermal conductivity for structural materials, and optical properties such as index of refraction or changes in transmittance spectra for optically active materials. The latter of these are often correlated directly to stress- or thermally-induced changes, themselves a function of the specific defect populations responsible for their change. The rapid study of such a mesoscale phenomenon requires an inherently mesoscale technique, where multiple quantities of interest can be accurately measured in very short times. Determining the sensitivity of any technique with known certainly is of critical importance to qualifying a new technique to measure in-situ material property changes during irradiation. In this paper, we present the new applicability of an existing technique, known as transient grating (TG) spectroscopy, to rapidly and accurately study key problems in radiation materials science. Results quantifying the absolute sensitivity of the technique are presented, alongside molecular dynamics (MD) simulations of the TG spectroscopy process to gauge direct agreement between theory, simulation, and experiment. An example problem of interest is demonstrated as an illustration of what is possible, showing elastic and thermal changes in irradiated silicon carbide (SiC) films. Finally, future work to quantify the effects of experimentally-simulated radiation defects is proposed, in support of building a database to use TG spectroscopy to deconvolve irradiated material property changes in-situ in terms of the defect populations responsible for their evolution. The application of the TG spectroscopy technique to in-situ radiation materials science has the potential to answer long-standing questions in the field, relating to ion beam rastering, the onset of void swelling, and direct measurement of radiation-created point defect concentrations. Before it can do so, the technique must carefully be analyzed and tuned to increase its sensitivity, the certainty of its results, and the ability to predict radiation-induced changes by simulations must be developed. Making such measurements in-situ with the aid of predictive MD simulations has the potential to elucidate mechanisms of radiation damage accumulation, as well as more rapidly qualify radiation-resistant alloys for service in nuclear systems. Work is underway to build this database of experimentally-simulated radiation-produced defects, which will be combined and compared with actual irradiated materials for eventual in-situ deconvolution of the TG signal into the defect populations responsible for irradiation-induced microstructural change. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

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

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

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Leveraging the High Flux Isotope Reactor for nuclear fuel development: a review of experiments, facilities, and capabilities

Materials testing reactors (MTRs) have been used to develop in-core nuclear fuels and materials since the outset of the nuclear power industry. However, the closure of prominent MTRs worldwide and protracted construction timelines for new facilities have increased reliance on existing infrastructure for near-term irradiation testing needs. One facility that can support these needs is the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. HFIR boasts the highest steady-state neutron flux in the Western Hemisphere and, among other roles, has been used to rapidly administer high fluences on fuels and materials for fission and fusion reactor applications. This paper reviews HFIR facilities and infrastructure, fuel-bearing irradiation experiments conducted in HFIR, and select nonfueled experiments that demonstrate advanced techniques transferable to fuels experiments. Collectively, these examples underscore HFIR's potential role as a nuclear fuels testbed supporting both the existing reactor fleet and advanced reactor fuel development.

Fuel qualification↗
DOE OSTI2015

Atomic-scale mechanisms of helium bubble hardening in iron

Generation of helium due to (n,α) transmutation reactions changes the response of structural materials to neutron irradiation. The whole process of radiation damage evolution is affected by He accumulation and leads to significant changes in the material s properties. A population of nanometric He-filled bubbles affects mechanical properties and the impact can be quite significant because of their high density. Understanding how these basic mechanisms affect mechanical properties is necessary for predicting radiation effects. In this paper we present an extensive study of the interactions between a moving edge dislocation and bubbles using atomic-scale modeling. We focus on the effect of He bubble size and He concentration inside bubbles. Thus, we found that ability of bubbles to act as an obstacle to dislocation motion is close to that of voids when the He-to-vacancy ratio is in the range from 0 to 1. A few simulations made at higher He contents demonstrated that the interaction mechanism is changed for over-pressurized bubbles and they become weaker obstacles. The results are discussed in light of post-irradiation materials testing.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Advanced Natural Gas Reciprocating Engine(s)

Energy independence and fuel savings are hallmarks of the nation’s energy strategy. The advancement of natural gas reciprocating engine power generation technology is critical to the nation’s future. A new engine platform that meets the efficiency, emissions, fuel flexibility, cost and reliability/maintainability targets will enable American manufacturers to have highly competitive products that provide substantial environmental and economic benefits in the US and in international markets. Along with Cummins and Waukesha, Caterpillar participated in a multiyear cooperative agreement with the Department of Energy to create a 50% efficiency natural gas powered reciprocating engine system with a 95% reduction in NOx emissions by the year 2013. This platform developed under this agreement will be a significant contributor to the US energy strategy and will enable gas engine technology to remain a highly competitive choice, meeting customer cost of electricity targets, and regulatory environmental standard. Engine development under the Advanced Reciprocating Engine System (ARES) program was divided into phases, with the ultimate goal being approached in a series of incremental steps. This incremental approach would promote the commercialization of ARES technologies as soon as they emerged from development and would provide a technical and commercial foundation of later-developing technologies. Demonstrations of the Phase I and Phase II technology were completed in 2004 and 2008, respectively. Program tasks in Phase III included component and system development and testing from 2009-2012. Two advanced ignition technology evaluations were investigated under the ARES program: laser ignition and distributed ignition (DIGN). In collaboration with Colorado State University (CSU), a laser ignition system was developed to provide ignition at lean burn and high boost conditions. Much work has been performed in Caterpillar’s DIGN program under the ARES program. This work has consisted of both modeling and single cylinder engine experiments to quantify DIGN performance. The air handling systems of natural gas engines dissipate a percentage of available energy as a result of both flow losses and turbomachinery inefficiencies. An analytical study was initiated to increase compressor efficiency by employing a 2-stage inter-cooled compressor. Caterpillar also studied a turbo-compound system that employs a power turbine to recover energy from the exhaust gases for improved engine efficiency. Several other component and system investigations were undertaken during the final phase of the program to reach the ultimate ARES goals. An intake valve actuation system was developed and tested to improve engine efficiency, durability and load acceptance. Analytical modeling and materials testing were performed to evaluate the performance of steel pistons and compacted graphite iron cylinder head. Effort was made to improve the detonation sensing system by studying and comparing the performance of different pressure sensors. To reduce unburned hydrocarbon emissions, different camshafts were designed and built to investigate the effect of exhaust valve opening timing and value overlap. 1-D & 3-D coupled simulation was used to study intake and exhaust manifold dynamics with the goal of reducing load in-balance between cylinders. Selective catalytic reduction with on-board reductant generation to reduce NOx emissions was also engine tested. An effective mean to successfully deploy ARES technologies into the energy markets is to deploy demonstration projects in the field. In 2010, NETL and Caterpillar agreed to include a new “opportunity fuel” deliverable and two field demonstrations in the ARES program. An Organic Rankine Cycle system was designed with production intent incorporating lessons learned from the Phase II demonstration. Unfortunately, business conditions caused Caterpillar to cancel this demonstration in 2011. Nonetheless, Caterpillar partnered with a local dealer to deploy an ARES class engine using syngas from a biomass gasifier as the DE-FC26-01CH11079 primary combustion fuel in Gleason, TN. Upon the successful start-up and commissioning of the demonstration unit, ownership of the system was transferred to the dealer. In order to further our understanding of syngas combustion, a fundamental combustion study on syngas combustion at high pressure and lean condition was conducted through the collaboration with University of Southern California. A Methane program was also developed to rate engine performance for various compositions of syngas using empirical data obtained at CSU. While much work remains in terms of extending and integrating these developments into commercial products, it is evident that engine manufacturers on our own or through private consortium efforts could not have overcome the financial hurdles to drive these improvements into reciprocating engine and system capabilities, helping maintain the natural gas reciprocating engine power generation technology as a strong option for electric power markets, both in the United States and worldwide.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2015

Structural Testing of the Blade Reliability Collaborative Effect of Defect Wind Turbine Blades

Two 8.3-meter (m) wind turbine blades intentionally constructed with manufacturing flaws were tested to failure at the National Wind Technology Center (NWTC) at the National Renewable Energy Laboratory (NREL) south of Boulder, Colorado. Two blades were tested; one blade was manufactured with a fiberglass spar cap and the second blade was manufactured with a carbon fiber spar cap. Test loading primarily consisted of flap fatigue loading of the blades, with one quasi-static ultimate load case applied to the carbon fiber spar cap blade. Results of the test program were intended to provide the full-scale test data needed for validation of model and coupon test results of the effect of defects in wind turbine blade composite materials. Testing was part of the Blade Reliability Collaborative (BRC) led by Sandia National Laboratories (SNL). The BRC seeks to develop a deeper understanding of the causes of unexpected blade failures (Paquette 2012), and to develop methods to enable blades to survive to their expected operational lifetime. Recent work in the BRC includes examining and characterizing flaws and defects known to exist in wind turbine blades from manufacturing processes (Riddle et al. 2011). Recent results from reliability databases show that wind turbine rotor blades continue to be a leading contributor to turbine downtime (Paquette 2012).

17 WIND ENERGY↗
DOE OSTI2026

Pull Force Evaluation of CCS Insulated End Caps

Insulated end caps installed on DC pins of Combined Charging System (CCS) inlets have been identified as potential debris sources within electric vehicle supply equipment (EVSE) connectors, increasing the risk of electric shock and fire hazards. Although standards such as IEC 62196-1:2022 specify mechanical pull-force requirements for these end caps, it remains unclear whether these requirements provide adequate robustness under real-world conditions. To address this concern, the National Charging Experience (ChargeX) Consortium's Hardware Task Force conducted evaluations of insulated end caps used in OEM CCS inlets. This study assesses the performance of insulated end caps used in OEM CCS inlets, specifically those installed on DC pins, using the procedures defined in IEC 62196-1:2022 (Section Sign) 26.7 and SAE J3400, supplemented by additional experimental conditioning. To ensure relevance to real-world conditions, NLR collaborated with charging network operators (CNOs) to guide the selection of charging inlet samples. Material analyses of the insulated end cap samples were conducted to confirm that the tested materials reflected the types and properties commonly observed in the field. The selected samples underwent temperature and humidity conditioning followed by pull-force testing to evaluate the end-cap integrity. Although SAE J3400 inlets were not directly tested due to limited diversity in available field data, the underlying end-cap principles between CCS and SAE J3400 are comparable, allowing the study's insights to be relevant to both technologies. Overall, this study provides a structured evaluation framework to inform potential refinements to mechanical pull-force requirements in charging standards.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2013

Overview of Hanford Single Shell Tank (SST) Structural Integrity

To improve the understanding of the single-shell tanks (SSTs) integrity, Washington River Protection Solutions, LLC (WRPS), the USDOE Hanford Site tank contractor, developed an enhanced Single-Shell Tank Integrity Project (SSTIP) in 2009. An expert panel on SST integrity, consisting of various subject matters experts in industry and academia, was created to provide recommendations supporting the development of the project. This panel developed 33 recommendations in four main areas of interest: structural integrity, liner degradation, leak integrity and prevention, and mitigation of contamination migration, Seventeen of these recommendations were used to develop the basis for the M-45-10-1 Change Package for the Hanford Federal Agreement and Compliance Order, which is also known as the Tri-Party Agreement. The structural integrity of the tanks is a key element in completing the cleanup mission at the Hanford Site. There are eight primary recommendations related to the structural integrity of Hanford Single-Shell Tanks. Six recommendations are being implemented through current and planned activities. The structural integrity of the Hanford is being evaluated through analysis, monitoring, inspection, materials testing, and construction document review. Structural evaluation in the form of analysis is performed using modern finite element models generated in ANSYS. The analyses consider in-situ, thermal, operating loads and natural phenomena such as earthquakes. Structural analysis of 108 of 149 Hanford Single-Shell Tanks has concluded that the tanks are structurally sound and meet current industry standards. Analysis of the remaining Hanford Single-Shell Tanks is scheduled for FY2014. Hanford Single-Shell Tanks are monitored through a dome deflection program. The program looks for deflections of the tank dome greater than 1/4 inch. No such deflections have been recorded. The tanks are also subjected to visual inspection. Digital cameras record the interior surface of the concrete tanks, looking for cracks and other surface conditions that may indicate signs of structural distress. The condition of the concrete and rebar of the Hanford Single-Shell Tanks is currently being tested and planned for additional activities in the near future. Concrete and rebar removed from the dome of a 65 year old tank was tested for mechanics properties and condition. Results indicated stronger than designed concrete with additional Petrographic examination and rebar completed. Material properties determined from previous efforts combined with current testing and construction document review will help to generate a database that will provide indication of Hanford Single-Shell Tank structural integrity.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Retractable Sensor Poster for ANIMMA 2023 Conference

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-pile environment. The test environment within these reactors are extremely harsh and long-term exposure causes sensor decalibration or failure. Most tests only require data points at daily intervals meaning that the sensor does not necessarily need to be in place continuously for the full experiment. One proposed solution to the issue described above is to have a sensor which can be periodically inserted through a guide tube into the test region and retracted; this would act as a reference check on the sensors installed permanently. If a compact and robust enough design can be made, it has the potential to extend the life of the sensor and keep it within calibration. This LRS entry is for a poster summarizing a presentation that will be given at the ANIMMA 2023 conference. The poster will be on display in the Measurement Sciences Laboratory booth at ANIMMA. Most of the material on this poster came from LRS entry INL/CON-23-72797.

42 ENGINEERING↗
DOE OSTI2021

Apparatus for skidding sensitivity testing of energetic materials

A remote-operated apparatus for testing the detonation sensitivity of energetic materials is detailed. Using an air ram and rotating disk, the normal force and transverse velocity of the impact plane are controlled independently, enabling the exploration of varying impact conditions over a wide parameter space. A microcontroller local to the apparatus is used to automate apparatus operation and ensure temporal alignment of the impacting ram head with the rotating disk. Calculation of the firing parameters and issuing of operational commands are handled by a remote computer and relayed to the local microcontoller for execution at the hardware level. Finally, impact forces are taken from fast strain measurements obtained from gauges incorporated into the ram head. Infrared imaging of explosive samples provides insight to the peak thermal temperatures experienced at the sample surface during the impact event.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Study of the influence of heat sources on the out-of-pile calibration curve of calorimetric cells used for nuclear energy deposition quantification

At present the Jules Horowitz Reactor is under construction in Cadarache research center of CEA 'French Alternative Energies and Atomic Energy Commission' center located in the south-east of France. This new Material Testing Reactor (MTR) will be operational in late 2019 and will allow the generation of a new experimental potential (up to 20 irradiation devices simultaneously) and new harsh conditions such as higher neutron fluxes (5.10{sup 14} n.cm{sup -2}.s{sup -1} for E≥1 MeV), faster material ageing and higher nuclear heating (up to 20 W/g for nominal capacity of 100 MW). In nuclear research field, the control and the measurement of the nuclear heating (energy deposition rate per mass unit induced by the interactions of radiations with matter) is crucial to carry out accurate studies on ageing of materials and on the behavior of nuclear fuels under irradiation. Several experiments need to know precisely this key parameter in order to establish dedicated thermal conditions. The measurement of the nuclear heating inside MTRs is realized by three kinds of sensors: single-cell calorimeter, differential calorimeter and gamma thermometer. One scientific objective of the IN-CORE program, between CEA and Aix-Marseille University in 2009, is to improve the nuclear heating measurement. In this context a new multi-sensor device, called CARMEN, was made. This device contains in particular a differential calorimeter which was designed to measure the nuclear heating in the periphery of OSIRIS reactor (a MTR located at Saclay, France) up to 2 W/g and tested during two irradiation campaigns. Results obtained during these campaigns showed that temperatures reached inside the calorimeter are higher than ones obtained during the preliminary out-of-pile calibration experiments. For instance for 1.74 W/g, the in-pile temperature of the calorimeter rod is equal to 305 deg. C against 225 deg. C in laboratory conditions by simulating the nuclear heating by Joule Effect inside the calorimeter cell head. This discrepancy is higher than in previous experiments because the calorimeter owns a high sensitivity. Consequently, a new prototype was created and instrumented by other heat sources in order to impose an energy deposition on the calorimetric cell structure (in particular in the base) and to improve the calibration step in out-of-pile conditions. In this paper, on the first part a detailed description of the new calorimetric sensor will be given. On the second part, the experimental response of the sensor obtained for several internal heating conditions will be shown. The influence of these conditions on the calibration curve will be discussed. Then the response of this prototype will be also presented for different external cooling fluid conditions (in particular flow temperature). In this part, the comparison between the in-pile and out-of-pile experimental results will be performed. On the last part, these out-of-pile experiments will be completed by 2D axisymmetrical thermal simulations with the CEA code CAST3M using Finite Elements Method. After a comparison between experimental and numerical works, improvements of the sensor prototype will be studied (new heat sources). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Standards and Best Practices for Solar Measurements

NREL leads and contributes to the development of radiometric standards and associated best practices through the American Society for Testing Materials (ASTM) International and others, such as the International Energy Agency (IEA) and International Organization for Standardization (ISO). Development of best practices and consensus standards in solar measurement enables industry to develop widely accepted protocols for various stages of solar project development and operations. This reduces barriers to seeking financing and reduces warranty costs. These standards and best practices play an essential role in all aspects of weathering and durability, including standard conditions, methods and instrumentation, accelerated testing, and service lifetime of materials systems.

14 SOLAR ENERGY↗
DOE OSTI2014

Final Report: Part 1. In-Place Filter Testing Instrument for Nuclear Material Containers. Part 2. Canister Filter Test Standards for Aerosol Capture Rates.

A portable instrument has been developed to assess the functionality of filter sand o-rings on nuclear material storage canisters, without requiring removal of the canister lid. Additionally, a set of fifteen filter standards were procured for verifying aerosol leakage and pressure drop measurements in the Los Alamos Filter Test System. The US Department of Energy uses several thousand canisters for storing nuclear material in different chemical and physical forms. Specialized filters are installed into canister lids to allow gases to escape, and to maintain an internal ambient pressure while containing radioactive contaminants. Diagnosing the condition of container filters and canister integrity is important to ensure worker and public safety and for determining the handling requirements of legacy apparatus. This report describes the In-Place-Filter-Tester, the Instrument Development Plan and the Instrument Operating Method that were developed at the Los Alamos National Laboratory to determine the “as found” condition of unopened storage canisters. The Instrument Operating Method provides instructions for future evaluations of as-found canisters packaged with nuclear material. Customized stainless steel canister interfaces were developed for pressure-port access and to apply a suction clamping force for the interface. These are compatible with selected Hagan-style and SAVY-4000 storage canisters that were purchased from NFT (Nuclear Filter Technology, Golden, CO). Two instruments were developed for this effort: an initial Los Alamos POC (Proof-of-Concept) unit and the final Los Alamos IPFT system. The Los Alamos POC was used to create the Instrument Development Plan: (1) to determine the air flow and pressure characteristics associated with canister filter clogging, and (2) to test simulated configurations that mimicked canister leakage paths. The canister leakage scenarios included quantifying: (A) air leakage due to foreign material (i.e. dust and hair) fouling of o-rings, (B) leakage through simulated cracks in o-rings, and (C) air leakage due to inadequately tightened canister lids. The Los Alamos POC instrument determined pertinent air flow and pressure quantities, and this knowledge was used to specify a customized Isaac® (Z axis, Salt Lake City, UT) leak test module. The final Los Alamos IPFT (incorporating the Isaac® leak test module) was used to repeat the tests in the Instrument Development Plan (with simulated filter clogging tests and canister leak pathway tests). The Los Alamos IPFT instrument is capable of determining filter clogging and leak rate conditions, without requiring removal of the container lid. The IPFT measures pressure decay rate from 1.7E-03 in WC/sec to 1.7E-01 in WC/sec. On the same unit scale, helium leak testing of canisters has a range from 5.7E-07 in WC/sec to 1.9E-03 in WC/sec. For a 5-quart storage canister, the IPFT measures equivalent leak flow rates from 0.03 to 3.0 cc/sec. The IPFT does not provide the same sensitivity as helium leak testing, but is able to gauge the assembled condition of as-found and in-situ canisters.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Corrosion and Fatigue Behavior of High-Strength Steel Treated with a Zn-Alloy Thermo-diffusion Coating

High and low cycle fatigue tests were conducted on high-strength steel using four-point bending. The materials tested were ASTM A723 steel in the as-machined condition, grit-blasted condition, MIL-DTL-16232 heavy manganese phosphate-coated condition, and ASTM A1059 Zn-alloy thermo-diffusion coated (Zn-TDC). The ASTM A723 steel base material exhibits a yield strength of ~1000 MPa. The effects of the surface treatments versus uncoated steel were examined. The fatigue life of the Zn-TDC specimens was generally reduced on as-coated specimens versus uncoated or phosphate-coated specimens. Several mechanisms are examined including the role of compressive residual stress relief with the Zn-TDC process as well as fatigue crack initiation from the hardened Zn-Fe alloy surface layer produced in the gas-metal reaction. Additionally, the effects of corrosion pitting on the fatigue life of coated specimens are explored as the Zn-TDC specimens exhibit significantly improved corrosion resistance over phosphate-coated and oiled specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Fibre optic extensometer for high radiation and high temperature nuclear applications

In the framework of the Joint Instrumentation Laboratory (LCI), gathering resources from SCK-CEN (Belgium) and CEA (France), we are developing an optical sensor in order to accurately measure radiation-induced elongation of material placed in the core of a Material Testing Reactor (MTR). This extensometer displays common advantages of Fibre Optic (FO) sensors: high resolution, easy remote sensing and multiplexing, and also compact size which is of particular interest for in pile experiments with little room available. In addition, light weight reduces gamma heating hence limiting the thermal effect. In accordance with the specifications, the sensor has preferably two fixing points defining a gauge length of 10 to 15 mm (as a minimum). The diameter is less than 2 mm. Intense gamma and neutron irradiation as well as high temperatures are the most difficult environment conditions to withstand. Reactor radiation produces huge losses in common optical fibre. The losses can be limited by selecting the fibres (mainly with high purity silica core), the wavelength range (800-1200 nm), and a measurement based on interferometry (largely insensitive to losses in the fibre thanks to the wavelength encoding of the useful signal). Heavy neutron - mainly - and gamma flux such as in MTR, also produce compaction of silica, resulting in a significant drift and preventing the use of commercial FO sensors in such environment. Knowing this issue we revised the basic scheme of Extrinsic Fabry Perot Interferometer (EFPI) in order to limit the effects of compaction. A first sensor prototype fixed on a stainless steel support was tested in the Smirnof test facility in the BR2 MTR in Mol (Belgium). The support was subject to a constant mechanical and thermal stress, and his dimensions were not supposed to vary. This test showed a very low drift of the revisited EPFI design under high irradiation field in comparison with a commercial EFPI. This result has to be confirmed with second generation sensors with an increased robustness. The other difficulty to face is high temperature. Fibre optics with metal coating allows safe operation under temperatures up to 400 deg.C and even higher. But differential dilatation between silica and typical metallic material produces differential elongation in the range of 0.5 10{sup -2} i.e 5000{mu}{epsilon} for an increase in temperature of 300-400 deg. C. Such large elongation has to be considered carefully in the sensor design and its fixing on the sample. Laboratory tests with a second generation of adapted sensors demonstrated their good accuracy: less than 1% error over a displacement of 100{mu}m (< 1 {mu}m error), at room temperature. The displacement measured can reach 250{mu}m (20000{mu}{epsilon} with a gauge length of 12.5 mm) or more. We are currently implementing metal-coated fibre and we are preparing the next in pile irradiation on the BR2 reactor scheduled for mid-2011. Other applications of the sensor can be considered. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Initial Back-to-Back Fission Chamber Testing in ATRC

Development and testing of in-pile, real-time neutron sensors for use in Materials Test Reactor experiments is an ongoing project at Idaho National Laboratory. The Advanced Test Reactor National Scientific User Facility has sponsored a series of projects to evaluate neutron detector options in the Advanced Test Reactor Critical Facility (ATRC). Special hardware was designed and fabricated to enable testing of the detectors in the ATRC. Initial testing of Self-Powered Neutron Detectors and miniature fission chambers produced promising results. Follow-on testing required more experiment hardware to be developed. The follow-on testing used a Back-to-Back fission chamber with the intent to provide calibration data, and a means of measuring spectral indices. As indicated within this document, this is the first time in decades that BTB fission chambers have been used in INL facilities. Results from these fission chamber measurements provide a baseline reference for future measurements with Back-to-Back fission chambers.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Enhanced Design of Radiation Tolerant High-Temperature Structural Health Monitoring Sensors

Acoustic emission sensors are vital in the nuclear industry for real-time structural health monitoring and early detection of material degradation. By capturing high-frequency stress waves emitted from defects like cracks, corrosion, or fatigue, acoustic emission sensors enable non-invasive monitoring of critical components such as reactor vessels, piping, and containment structures. This technology supports predictive maintenance, enhances safety, and ensures regulatory compliance by providing early warnings of potential failures. It is also instrumental in research, particularly in material testing reactors, where it is used to monitor the behavior of fuels and materials under irradiation, by allowing the detection of cracking or other acoustic signals in real time. This enables the evaluation of performance and accident behavior of advanced fuel concepts.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗