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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 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Feedback of the Dutch Historical Waste Management Program - 17085

This paper provides a feedback on the Dutch historical waste management program. Over the past decades, the High Flux Reactor located in Petten in the province of North Holland has tested numerous types of materials and fuels irradiation, for both medical and nuclear energy applications. A number of waste containers was accumulated on site, where the storage facility was initially built. Following Dutch political decisions in the 1980's, the contents of the waste storage facility in Petten must be transferred to another location. The consequences of this decision included the transportation of waste containers from Petten to the new location, which is located in the South-West of the Netherlands. This project was initiated and is now running. It includes a complex structure and can be described as follows: each historical-waste container will be opened, its contents segregated in three categories depending on the dose rate emission, characterized and repacked in new containers suitable for transportation and storage at the new site. The approach chosen is based on the final dose rate emitted by the repacked containers, which are sub-divided into 3 categories.. The lowest dose-emitting containers are transported directly to the new storage facility. The 2 other categories are first transported to a third party to be super compacted and subsequently transported to the new storage facility. So far, the segregation method is based on Co-60 gamma emission and nuclide vectors associated to the different materials. A pre-sorting has been established to fit as much as possible IAEA recommendations. Historical waste has then to be classified in families sharing common history. Thus, a set of canisters containing the pieces of the old reactor vessel, which was replaced in 1984 was the first challenge, followed by a set of canisters containing residues of instruments only and pieces coming from the top of the control rods of the reactor which were not directly irradiated in the reactor. So far, the two first sets of canisters are being treated and characterized as well as possible. Reporting and official documents have been filled up to comply to Dutch transport regulations and to define waste acceptance criteria for the storage facility. The first transports of containers with Low Level Waste occurred in October 2015, and another transport took place at the end of 2016. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Data-driven analysis of ultrasonic pressure tube inspection data - 220

Pressure tubes are critical components of the CANDU reactors and other pressurized heavy water type reactors, as they contain the nuclear fuel and the coolant. Manufacturing flaws, as well as defects developed during the in-service operation, can lead to coolant leakage and can potentially damage the reactor. The current inspection process of these flaws is based on manually analyzing ultrasonic data received from multiple probes during planned, statutory outages. Recent advances on ultrasonic inspection tools enable the provision of high resolution data of significantly large volumes. This is highlighting the need for an efficient autonomous signal analysis process. Typically, the automation of ultrasonic inspection data analysis is approached by knowledge-based or supervised data-driven methods. This work proposes an unsupervised data-driven framework that requires no explicit rules, nor individually labeled signals. The framework follows a two-stage clustering procedure that utilizes the DBSCAN density-based clustering algorithm and aims to provide decision support for the assessment of potential defects in a robust and consistent way. Nevertheless, verified defect dimensions are essential in order to assess the results and train the framework for unseen defects. Initial results of the implementation are presented and discussed, with the method showing promise as a means of assessing ultrasonic inspection data. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

In-Containment Signal Conditioning and Transmission via Power Lines within High Dose Rate Areas of Nuclear Power Plants

Signal conditioning and transmission for sensor systems and networks within the containment of nuclear power plants (NPPs) still poses a challenge to engineers, particularly in the case of equipment upgrades for existing plants, temporary measurements, decommissioning of plants, but also for new builds. This paper presents an innovative method for efficient and cost-effective instrumentation within high dose rate areas inside the containment. A transmitter-receiver topology is proposed that allows simultaneous, unidirectional point-to-point transmission of multiple sensor signals by superimposing them on existing AC or DC power supply cables using power line communication (PLC) technology. Thereby the need for costly installation of additional cables and containment penetrations is eliminated. Based on commercial off-the-shelf (COTS) electronic parts, a radiation hard transmitter is designed to operate in harsh environment within the containment during full plant operation. Hardware modularity of the transmitter allows application specific tradeoffs between redundancy and channel bandwidth. At receiver side in non-radiated areas, signals are extracted from the power line, demodulated, and provided either in analog or digital output format. Laboratory qualification tests and field test results within a boiling water reactor (BWR) are validating the proof of concept of the proposed system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Irradiation Testing of Sensors - ASI Program Review 2022

This is a powerpoint presentation for the annual Advanced Sensors and Instrumentation (ASI) program webinar. The following is a summary of the presentation. Advanced instrumentation enables testing of nuclear fuels and materials in support of the US advanced nuclear technology industry. A number of promising sensor technologies are in the pipeline. The early part of sensor development can be done outside of the reactor environment, but full technical readiness requires experience gained from in-core performance testing. Customers usually have only one shot to conduct their irradiation experiments. Because of the high costs test sponsors are frequently reluctant to incorporate unproven technologies in reactor experiments. Therefore, it is vital to demonstrate newly-developed sensors in operational conditions, prior to incorporating them into long-term high-value experiments. Successful completion of these activities will create new monitoring capabilities in high-power test reactors (such as ATR) as well as specialty reactors such as TREAT.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Pile noise experiment in MINERVE reactor to estimate kinetic parameters using various data processing methods

MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from two high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel is additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with self-cooling system (enhanced comparing to the cooling properties inherent with regular rector pool water flows), -proper cooling conditions can be ensured by an appropriate water flow, so the resistance to flow has to be optimised, -the requirement of the minimum resistance to water flow leads to the openwork design of the fuel element separator, which, on the other hand, has to be strong enough to ensure the needed strength for mechanical load due to the fuel weight and forces associated with the water flow, -the possibility of changing beam and flux qualities by rotating the converter or repositioning the converter plates by moving or replacing with another materials. In order to minimize the neutron activation of the fuel in the converter, the possibility was predicted to remove the converter and to replace it with an aluminium dummy for the time when the beam at the channel H2 is not used. This means that both, the converter and the dummy, have to be easily removable from the converter socket. There has to be also the place in the water pool, near the research stand or in technological pool, where the converter can be safely stored (this place have to be proper for operation with plates i.e. changing amount of plates). Thermal and neutron load of the fuel plates in the converter will be inhomogeneous. In order to equalize these loads, the converter should be designed in such way that it would be possible to change the order of fuel plates. Moreover replacing the amount of the plates gives the opportunity to obtain different fluxes of neutrons (quantitatively and qualitatively i.e. energetically). The project of the converter is based on Monte Carlo calculation concerning neutron production and on Computational Fluid Dynamics (CFD) i.e. modelling of converter for thermodynamical aspects. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Development and applications of retro-reflective surfaces for ultrasound in LBE

The Belgian Nuclear Research Centre SCK-CEN is in the process of developing MYRRHA - a fast flux research reactor to replace the aging BR2. MYRRHA is conceptualized as an accelerator driven system cooled by lead bismuth eutectic mixture (LBE). As LBE is opaque to visual light, ultrasonic measurement techniques are employed as the main technology to provide feedback to submerged operations when needed. Conceptually, MYRRHA is a pool type reactor divided in a cold lower plenum and hot upper plenum separated by a diaphragm that forces the main flow through the core. The main flow is cooled by four heat exchangers and driven by two liquid metal pumps. One of the tasks tackled using ultrasound is locating a potentially lost fuel assembly to assist a recovery operation. As all fuel manipulations in MYRRHA are performed in the lower plenum, a potentially lost fuel assembly is located in the lower plenum. Buoyancy will force the lost fuel assembly to float against the diaphragm unless it is still partially inserted in the core. Because of the latter situation, an ultrasonic scan localizing the fuel assembly should be performed from a large distance to avoid a collision with such a partially inserted fuel assembly. Unfortunately, standard machined stainless steel objects, such as a fuel assembly, reflect an ultrasonic pulse in a specular way which induces stringent requirements on the alignment of the ultrasonic sensor with respect to the fuel assembly as we cannot rely on diffuse reflections and/or scattering of the ultrasonic pulse. Moreover, increasing the distance also increases geometric spreading and absorption of the pulse weakening the signal amplitude even faster to noise levels when deviating from perfect alignment. An alternative approach consists in relying on reflections from the known surroundings: a lost fuel assembly will block the line-of-sight to the diaphragm resulting in an anomaly in the reflection - either a shorter than expected time-of-flight of the pulse or a complete absence of a reflection like a shadow. In that way, it suffices to align the sensor with the diaphragm instead of the fuel assembly which is much easier to achieve as the robotics on which the sensor is mounted move parallel with the diaphragm. The alignment requirement in the latter approach can be further relaxed by using a tiling of retro-reflectors on the lower surface of the diaphragm. In that way, alignment becomes less vital and the main source of acoustic energy loss - geometric spread of the beam - is almost completely removed, leaving only absorption losses. In this paper, we present the first results in developing a retro reflectance surface for ultrasound in LBE. We present experimental results for different designs of retro-reflectors in both water and LBE. We discuss both linear and array retro-reflectors of different sizes and investigate the influence of the main relevant ultrasonic parameters such as wavelength and spot size on the strength of the received reflection under different alignment angles. We also demonstrate how retro-reflective surfaces can be exploited when localizing objects using linear and rotating scanning methods. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Implementation and Analysis for APR1400 Soft Control System

Due to the rapid advancement of digital technology, the definite technical advantages of digital control system compared to analog control system are accelerating the implementation of advanced distributed digital control system in the nuclear power plant. One of the major advantages of digital control system is the capability of Soft Control System. The design of Soft Control System for Advanced Power Reactor 1400 (APR1400) plant of Man-Machine Interface System (MMIS) is based on full digital technologies to enhance reliability, operability and maintainability. Computer-based compact workstation has been adopted in the APR1400 Main Control Room (MCR) to provide convenient working environment. This paper introduces the approaches and methodologies of Soft Control System for the Advanced Control Room (ACR). This paper also explains major design features for operation and display of the Soft Control System and its implementation to cope with regulatory requirements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Design and commissioning of a high magnetic field muon spin relaxation spectrometer at the ISIS pulsed neutron and muon source

The high magnetic field (HiFi) muon instrument at the ISIS pulsed neutron and muon source is a state-of-the-art spectrometer designed to provide applied magnetic fields up to 5 T for muon studies of condensed matter and molecular systems. The spectrometer is optimised for time-differential muon spin relaxation studies at a pulsed muon source. We describe the challenges involved in its design and construction, detailing, in particular, the magnet and detector performance. Commissioning experiments have been conducted and the results are presented to demonstrate the scientific capabilities of the new instrument.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactor innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Benefits of digitalizing and employing simulation to increase plant system performance and ensure compliance with technical specifications - 109

The essential services water system in a BWR nuclear power plant was affected by corrosion, soiling and blockages due to ageing. Those episodes affected safety margins hindering the Technical Specifications compliance during periodic system testing. Tecnatom designed, implemented and tested a design modification for the on-line monitoring of this system by integrating it in the existing Digital Control System (DCS). Key system parameters were acquired in real time to be displayed in the Human System Interface, used in performing calculations and stored their historical evolution. The objective was to optimize monitoring and surveillance of the essential service water system. Furthermore, Tecnatom developed an engineering simulator ('What if' simulator) consisting of a hydraulic model of the system. This simulator takes as inputs each heat exchanger performance parameters from the plant DCS, allows the user to change the essential services water system configuration (valves position, UHS level...) and calculates theoretical process values predicting the system real behavior. In a second stage, more system instruments were wired to the DCS for Technical Specifications compliance automatic surveillance regarding opening and closing times of system valves. Operating experience shows important benefits after the digitalization. It enhances on-line information of system operation and automates calculation of load loss factor of each heat exchanger. It eases the evaluation of blockages and soiling in the system allowing optimization of cleaning proceedings. System Surveillance Requirements and Test Requirements reports have been automated optimizing the operation workload. The simulator use enables to identify optimal system configuration increasing safety margins of the system and ensuring Technical Specifications fulfillment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Long-Term Instrumentation, Information, and Control Systems (II&C) Modernization Future Vision and Strategy (Revision 1)

Life extension beyond 60 years for the U.S operating nuclear fleet requires that instrumentation and control (I&C) systems be upgraded to address aging and reliability concerns. It is impractical for the legacy systems based on 1970's vintage technology operate over this extended time period. Indeed, utilities have successfully engaged in such replacements when dictated by these operational concerns. However, the replacements have been approached in a like-for-like manner, meaning that they do not take advantage of the inherent capabilities of digital technology to improve business functions. And so, the improvement in I&C system performance has not translated to bottom-line performance improvement for the fleet. Therefore, wide-scale modernization of the legacy I&C systems could prove to be cost-prohibitive unless the technology is implemented in a manner to enable significant business innovation as a means of off-setting the cost of upgrades. A Future Vision of a transformed nuclear plant operating model based on an integrated digital environment has been developed as part of the Advanced Instrumentation, Information, and Control (II&C) research pathway, under the Light Water Reactor (LWR) Sustainability Program. This is a research and development program sponsored by the U.S. Department of Energy (DOE), performed in close collaboration with the nuclear utility industry, to provide the technical foundations for licensing and managing the long-term, safe and economical operation of current nuclear power plants. DOE's program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy security and environmental security. The Advanced II&C research pathway is being conducted by the Idaho National Laboratory (INL). The Future Vision is based on a digital architecture that encompasses all aspects of plant operations and support, integrating plant systems, plant work processes, and plant workers in a seamless digital environment to enhance nuclear safety, increase productivity, and improve overall plant performance. The long-term goal is to transform the operating model of the nuclear power plants (NPP)s from one that is highly reliant on a large staff performing mostly manual activities to an operating model based on highly integrated technology with a smaller staff. This digital transformation is critical to addressing an array of issues facing the plants, including aging of legacy analog systems, potential shortage of technical workers, ever-increasing expectations for nuclear safety improvement, and relentless pressure to reduce cost. The Future Vision is based on research is being conducted in the following major areas of plant function: (1) Highly integrated control rooms; (2) Highly automated plant; (3) Integrated operations; (4) Human performance improvement for field workers; and (5) Outage safety and efficiency. Pilot projects will be conducted in each of these areas as the means for industry to collectively integrate these new technologies into nuclear plant work activities. The pilot projects introduce new digital technologies into the nuclear plant operating environment at host operating plants to demonstrate and validate them for production usage. In turn, the pilot project technologies serve as the stepping stones to the eventual seamless digital environment as described in the Future Vision.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Cryogenic system with GM cryocooler for krypton, xenon separation from hydrogen-helium purge gas

In the thorium molten salt reactor (TMSR), fission products such as krypton, xenon and tritium will be produced continuously in the process of nuclear fission reaction. A cryogenic system with a two stage GM cryocooler was designed to separate Kr, Xe, and H{sub 2} from helium purge gas. The temperatures of two stage heat exchanger condensation tanks were maintained at about 38 K and 4.5 K, respectively. The main fluid parameters of heat transfer were confirmed, and the structural heat exchanger equipment and cold box were designed. Designed concentrations after cryogenic separation of Kr, Xe and H{sub 2} in helium recycle gas are less than 1 ppb.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Reactivity measurements using the Zolotukhin-Mogilner Method

The zero count probability method (also called Zolotukhin-Mogilner Method, after its inventors) is a method for measuring the reactivity in nuclear reactors, being a competitor of the [more well-known in the West] Feynman-alpha Method and the Rossi-alpha Method. The modus operandi of this method is using the detector channels where there was no counting at all in order to calculate reactor parameters. In deep subcriticality few models have been tested and this work tries out the Zolotukhin-Mogilner Method in one of such scenarios: measurements will be made in environments below -3,5 k pcm in the zero-potency water-moderated reactor IPEN/MB-01 which is fuelled by UO{sub 2} enriched by 4.3%. These extremely low reactivity environments are required because the chance of no counts on the detector must be significant: otherwise, the method would demand a large time of acquisition of data. Besides that, the method is very simple and straightforward. One of the advantages of this method is that it needs very little data reduction, since the essential data is directly given by the measuring apparatus. The detection will be in charge of modern BF{sub 3} detectors. It will be assumed that, in these deep subcritical systems, the function K{sub eff} = f(N-Nkp) has a linear portion in its first part, resulting of the decomposition of it in a Taylor series. The value of alpha is related to the reactivity ρ with linear dependency. The results will be compared with recent studies of the two main methods described above. Presented for the first time in the 60's, this tool has seen little use in the west hemisphere. This work shows its use in the measurements of the nuclear reactor IPEN/MB-01, as well as the code developed for its employment. It will be the first time this method is used in the south hemisphere. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Study on main control room and human-system interface of generation III in China - 313

This paper focuses on the study on Main Control Room (MCR) and Human-System Interface (HSI) of Generation III in China. An intelligent HSI has been developed to enhance the safety and availability of Nuclear Power Plant (NPP) by improving operational reliability. The key elements of the proposed HSI such as Advanced Alarm System, State Oriented Procedures (SOP), Large Display Panel (LDP), Computerized-base procedures..., that affects what information the operator sees or the system's response to a control input must be empirically evaluated to ensure that the new design does not compromise HSI effectiveness. Based on CPR1000's experience in China, this paper presents the approach used as well as the most relevant aspects of this kind of project for the next generation. This approach will be used in new Generation III nuclear power plant, and also used in modernizing I and C system in currently operative nuclear power plants, in addition to meeting safety requirements and the plant's operational requirements, to improve cost-effective plant and human performance and to reduce likelihood of human errors, to gain maximum benefit of the implemented technology and to increase the performance, resulting in improved plant safety, availability, reliability, and cost-effective operation. The proposed HSI has been validated and demonstrated using the CPR1000 simulator of National Key Laboratory. After sufficient validation, the characteristic design features of the proposed HSI will be reflected in the design of the main control room of the next generation reactors in China. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2007

Design Concept for a Nuclear Reactor-Powered Mars Rover

A report presents a design concept for an instrumented robotic vehicle (rover) to be used on a future mission of exploration of the planet Mars. The design incorporates a nuclear fission power system to provide long range, long life, and high power capabilities unachievable through the use of alternative solar or radioisotope power systems. The concept described in the report draws on previous rover designs developed for the 2009 Mars Science laboratory (MSL) mission to minimize the need for new technology developments.

Elliott, John↗