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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Expected total counts for the Self-Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in spent fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron counts in the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach in this study consisted in introducing a small neutron detector in the central guide tube of a PWR 17x17 fuel assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types were used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the total neutron counts that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of total neutron counts and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the total neutron counts by increasing the detector size. The study shows that the highest total neutron counts are achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the total neutron counts associated to each detector type will play a major role in the selection of the detector types used for the SINRD measurement. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

The Acoustic Emission signal acquired by the microphones placed in the CABRI test device along the fourteen last R.I.A. experiments: an example of reproducible research in nuclear science

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission. It is located at the Cadarache Research Centre in southern France and it is designed to act as a support to the French nuclear infrastructure. The purpose of the new testing programme termed, 'CABRI International Programme' (CIP) is to study the behaviour of PWR-type fuel rods at high burnup, equipped with an 'advanced' cladding, under Reactivity Initiated Accident (RIA) conditions (such as the scenario of a control rod ejection). Within the framework of this programme, piloted and funded by the French Institute of Nuclear Radioprotection and Safety (IRSN), ten tests are to be conducted with a frequency of two tests per year. The LPRE laboratory of the CEA which is in charge of the Preparation, realisation and breakdown of the test results studies the possibility to set up a new test analysis based on the processing of signals coming from sensors placed within the test equipment. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream of the test device. Studies showed the interest to realize temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena as the rod failure or the test shutdown (i.e. scram). The work presented in this article results from the will to consolidate these studies. Since the main phenomenon to be tracked is the fuel rod failure, the aim would be to highlight specific events which would have been precursors of the rod failure in order to use in the future these signals for further interpretation. Such an antecedent works resumption leads to a better understanding of the experimental needs and constitutes a good initial state to prepare the new very fast digital data acquisition systems. Although all the raw data are accessible in the form of text files, analyses and graphics representations were not straightforward to reproduce from the ancient studies since that, on one hand, people who were in charge of the original work left the laboratory and on the other hand because it is not easy when the time passes, even with our own work, to be able to remember the steps of data manipulations and the exact setup: - During the ancient experiments the use of analog data acquisition systems required to digitize tapes to be able to realize computer treatments. That had had for consequence to lose the initial dating. This one must be correctly edited to do temporal comparisons. - Analyses require functions for calculations whose parameters has to be well-known to reach the same results. We thus wished to manage our workflow in the idea that it can be easily reproducible on all the experiments. The object of the work presented in this article was to put in practice this strong bind between the data, treatments and generation of the document in order not to hesitate to do the iteration principle in action. We do not have to be afraid by the data driven analyses. According to the philosophy of the literate programming, the text of the technical document is woven with the computer code that produces all the printed output as tables, graphs for the study eliminating hence the unrealistic cut and paste. This difficulty is not specific to the nuclear domain. For many years, researchers have been worked out solutions to this mundane issue. And, presently, new technologies and high-level programming languages offer us actual answers. We will firstly present the tools applied in our laboratory to implement this workflow, then we will describe the global perception carried out to continue the study of the Acoustic Emission signals recorded by the two microphones during the fourteen last CABRI R.I.A. test.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Neutron Guide System for Ultracold and Cold Neutrons at the WWR-M Reactor

The results of calculation of fluxes of ultracold (UCNs), very cold, and cold neutrons at the output of neutron guides of the UCN source with superfluid helium at the WWR-M reactor are presented. UCN density ρ{sub 35L} = 1.3 × 10{sup 4} n/cm{sup 3} in the trap of the electric dipole moment (EDM) spectrometer was obtained by optimizing source parameters. This UCN density in the EDM spectrometer is two orders of magnitude higher than the UCN density at the output of the available UCN sources. The flux density of cold neutrons with a wavelength of 2–20 Å at the output of a neutron guide with a cross section of 30 × 200 mm{sup 2} should be as high as 1.1 × 10{sup 8} n/(cm{sup 2} s), while the flux density of very cold neutrons (50–100 Å) at the output of the same neutron guide should be 2.3 × 10{sup 5} n/(cm{sup 2} s). An extensive program of fundamental and applied physical research was mapped out for this source.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Numerical and Experimental Thermal Responses of Single-cell and Differential Calorimeters: from Out-of-Pile Calibration to Irradiation Campaigns

The nuclear radiation energy deposition rate (usually expressed in W.g{sup -1}) is a key parameter for the thermal design of experiments, on materials and nuclear fuel, carried out in experimental channels of irradiation reactors such as the French OSIRIS reactor in Saclay or inside the Polish MARIA reactor. In particular the quantification of the nuclear heating allows to predicting the heat and thermal conditions induced in the irradiation devices or/and structural materials. Various sensors are used to quantify this parameter, in particular radiometric calorimeters also called in-pile calorimeters. Two main kinds of in-pile calorimeter exist with in particular specific designs: single-cell calorimeter and differential calorimeter. The present work focuses on these two calorimeter kinds from their out-of-pile calibration step (transient and steady experiments respectively) to comparison between numerical and experimental results obtained from two irradiation campaigns (MARIA reactor and OSIRIS reactor respectively). The main aim of this paper is to propose a steady numerical approach to estimate the single-cell calorimeter response under irradiation conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Identification of NPP transients using artificial intelligence - 335

If the accidents happen in nuclear power plants (NPPs), the NPP operators identify the transients through observing the behaviors of the major parameters. However, the integrity of the instrumentation signals is not ensured under the accident circumstances and the major parameters essential in identifying the transients can rapidly change although the instruments generate accurate signals. Thus, to successfully control the accidents by taking necessary measures relatively faster, it is very important to provide the operators with the information that can identify the transients and accidents accurately. In this study, the transients of severe accidents in NPPs are identified and classified using artificial intelligence methodologies. Several transients are identified using support vector classification (SVC) and probabilistic neural network (PNN) as artificial intelligence methodologies. The used data for the transients are the simulation data using the modular accident analysis program (MAAP) code for the optimized power reactors (OPR1000s). The time-integrated values of chosen simulated sensor signals are used as the input variables for identifying the transients. As a result of this study, the proposed models identified the transients quite precisely. Therefore, the quite accurate identification of the transients indicates the excellence of these artificial intelligence methods and they will produce useful supporting information for the operators even under extreme transient circumstances in NPPs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

TMI-2 – A Case Study for PWR Instrumentation Performance During a Severe Accident

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focused upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Critical review of gamma spectrometry detection approaches for in-plant surface deposition monitoring

Surface deposition of activated corrosion product on oxide layers of light-water reactor primary system components is the primary source for ex-core radiation fields and personnel radiation exposure. Understanding the deposition mechanism and what factors influence the deposition and release behaviors are crucial for developing effective radiation field reduction measures. One of the available tools to assess the surface deposition is in-plant gamma spectrometry, which has been performed for several decades using either sodium iodide (NaI) or high-purity germanium (HPGe) detectors. Lately, the much more mobile cadmium-zinc-telluride (CZT) detectors are increasingly employed by stations because of their ease in use and handling. However, all of these gamma detectors face the same challenges; namely large-geometry samples of inconsistent sample compositions and sometimes gaps in the information necessary to establish proper efficiency calibrations. This paper reviews current measurements and efficiency calibration approaches taken in the industry. The validity of the measurement results and the feasibility of the data's use in understanding source term behavior is examined. Suggestions are made for the development of a more robust deposit characterization and radiation field monitoring program. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

BWR modernization project: installation of a new digital feedwater control system - 295

A European BWR plant carried out a relevant and pioneering modernization project regarding the Feedwater and Feed Pump Turbine Control System, replacing the legacy analogic control with the new General Electric MarkVIe digital system. This digital instrumentation and control modernization, motivated by obsolescence, had a great impact on the main control room and on the plant operation. Tecnatom collaborated with the nuclear power plant in the project performing the following activities: early installation of the system in the full scope and classroom simulators, Human Factors Engineering (HFE) Verification and Validation (V and V), Simulation Assisted Engineering (SAE), design and implementation of cyber secure communication with the Plant Process Computer (PPC), operating procedures modification and plant operators training. Those activities contributed to the success of the project that was completed during the refueling outage as planned. Scheduling HFE verification before the design freeze allowed the plant to modify the Human Machine Interface (HMI) to comply with the standards and increase safety without overcharging the project. SAE resulted in the reduction of costs and time during the system commissioning through the detection of discrepancies and improvements in the control, covering different plant scenarios. All the activities carried out in the simulators before the implementation of the system in the plant contributed to validate it and to build confidence in the new control among the plant personnel. New digital system communication interfaces with the PPC saved cable laying works and allowed future expansions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

An expert-systems approach to automatically determining flaw depth within Candu pressure tubes - 221

Delayed Hydride Cracking (DHC) is a crack growth mechanism that occurs in zirconium alloys, including the pressure tubes of CANDU reactors. DHC is caused by hydrogen in solution in zirconium components being diffused to any flaws present, resulting in an increased concentration of hydrogen within these flaws. An increased hydrogen concentration can lead to brittleness, followed by cracking, in high-stress regions of a pressure tube. Regular in-service ultrasonic inspection of CANDU pressure tubes aims to locate and classify any flaws that pose a potential for DHC initiation. A common approach to inspection is the use of a bespoke tool containing multiple ultrasonic transducers to ensure that each point on the pressure tube is inspected from a minimum of three angles during a scan. All flaws from within the inspected pressure tubes must be characterized prior to restarting the reactor, thus the time-consuming analysis process lies on the critical outage path. This process is manually intensive and often requires a significant amount of expert knowledge. A modular system to automatically process outage data to provide decision support to analysts has been developed. This system saves time on the critical outage path while providing repeatable and explicable measurements. Part of the analysis process requires the depth of all flaws to be measured, which is often the most time consuming stage of the analysis process. This paper describes an approach that utilizes captured analysts knowledge to perform automatic flaw depth estimation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

GDOE Bearing

Explore the source record for details and available documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hour of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Non-destructive assay of mechanical components using gamma-rays and thermal neutrons

This work presents the results obtained in the inspection of several mechanical components through neutron and gamma-ray transmission radiography. The 4.46 Multiplication-Sign 10{sup 5} n.cm{sup -2}.s{sup -1} thermal neutron flux available at the main port of the Argonauta research reactor in Instituto de Engenharia Nuclear has been used as source for the neutron radiographic imaging. The 412 keV {gamma}-ray emitted by {sup 198}Au, also produced in that reactor, has been used as interrogation agent for the gamma radiography. Imaging Plates - IP specifically designed to operate with thermal neutrons or with X-rays have been employed as detectors and storage devices for each of these radiations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard-fired alumina insulation and a molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple, based on molybdenum/niobium alloys and developed at Idaho National Laboratory, was also tested. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Integrating Human Performance and Technology

Human error is a significant factor in the cause and/or complication of events that occur in the commercial nuclear industry. In recent years, great gains have been made using Human Performance (HU) tools focused on targeting individual behaviors. However, the cost of improving HU is growing and resistance to add yet another HU tool certainly exists, particularly for those tools that increase the paperwork for operations. Improvements in HU that are the result of leveraging existing technology, such as hand-held mobile technologies, have the potential to reduce human error in controlling system configurations, safety tag-outs, and other verifications. Operator rounds, valve line-up verifications, containment closure verifications, safety & equipment protection, and system tagging can be supported by field-deployable wireless technologies. These devices can also support the availability of critical component data in the main control room and other locations. This research pilot project reviewing wireless hand-held technology is part of the Light Water Reactor Sustainability Program (LWRSP), a research and development (R&D) program sponsored by the U. S. Department of Energy (DOE). The project is being performed in close collaboration with industry R&D programs to provide the technical foundations for licensing, and managing the long-term, safe, and economical operation of current nuclear power plants. The LWRSP vision is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current nuclear reactor fleet.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Atmospheric {sup 41}Ar Plume Tracking of the UF Training Reactor Using a Large-Volume NaI Detector

The ability to track and monitor gaseous radioactive materials is important in the event of a radioactive release such as from a nuclear power plant accident, weapons testing, or other unintended release scenarios. The International Monitoring System (IMS) is currently in place to monitor for weapons testing and has been used to monitor nuclear power plant accidents. However, only 80 radionuclide stations for the IMS exist in the world and are stationary, limiting the resolution of any radioactive plumes detected. Furthermore, the IMS radionuclide stations are designed to be sensitive specifically to atmospheric radio-xenon and no other radionuclides. Historically, monitoring radioactive releases has been limited to the detection of radio-xenon and radio-krypton, both arising as fission products from nuclear fuel. After the accident at Three Mile Island Unit 2, scientists were interested in finding the dose the public received from released radioactive material. This was achieved by analyzing radiation monitors in the exhaust stack and filters within the reactor building. furthermore, a series of 20 thermoluminescent dosimeters (TLD's), 8 air sampling station positioned around Three Mile Island, and a helicopter survey which used a NaI(Tl) detector array provided data on radionuclides, specifically {sup 137}Cs, that escaped the confines of the reactor building. The TLD's and air sampling stations were stationary, and therefore limited the resolution of the radioactive release plume. The helicopter survey using the NaI detectors was limited by the technology available at the time. The NaI detector array output was compressed from 1024 channels to 256 channels, limiting the energy resolution of the spectral data. The helicopter used a microwave ranging system to establish position. Since this survey was taken in October of 1982, advancements in technology allow spectral data to be saved in its full 1024 channel format and the advent of GPS can give a more accurate account of the detector position. An updated mobile system for tracking a radioactive release would help to alleviate these issues as well as gain the ability to track a radioactive release in real-time. We investigate the feasibility of mobile tracking through study and measurement of the {sup 41}Ar plume released from the University of Florida Training Reactor (UFTR) research reactor. We use a large-volume NaI(Tl) detector placed on the roof near the reactor. This detector may also be placed and moved on a cart to survey the area surrounding the reactor. We compare the spectral data taken when the reactor is running at full power to the background to analyze the effectiveness of the detector at identifying atmospheric {sup 41}Ar. Finally we discuss the implications of these measurements with regards to validation of plume models and the capability to track releases over a localized area from the source. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗