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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague. The Training Reactor VR-1 is a pool type (light water) reactor based on UO{sub 2} low enriched uranium. It has a nominal power of 1 kW, and can be operated for a short period up to 5 kW. The arrangement of the reactor pool reactor facilitates access to the core, setting and removal of various experimental samples and detectors, and safe and easy handling of fuel assemblies. The reactor is equipped with two horizontal channels (radial and tangential) and 10 vertical channels, of varying diameters, which can be loaded into various core positions, and one pneumatic transfer system. It is also equipped with several specifically designed educational instrumentation systems that can be used to supply complementary measurements and characterization around the reactor. The reactor is operated by the Department of Nuclear Reactors of the Faculty of Nuclear Sciences and Physical Engineering of the Czech Technical University in Prague. The two detectors were placed in-core through one of the vertical insertion channel. They were coupled to remote placed (5 m BNC cable) classical nuclear charge sensitive electronics. Detection properties of both sensors, including: pulse height spectra of U-235 fission fragments (response linearity with neutron flux, count rate, gamma background, were evaluated varying the power of the reactor from 0.005 W to 500 W. The evolution of the counting rate of the thinned optical grade detector as a function of counting rate of a gas ionization chamber used currently for reactor monitoring shows the very good linearity of the detector over the 5 decades. Similar results were obtained with the PIM detector. Additionally fast transient current signals of the detectors were recorded on a digital storage oscilloscope (DSO) using broad-band amplifier and with a simple bias-T, showing potential use of such sensors for neutron counting with no need of an amplification stage, since non-amplified signals from fission fragments exceeded 4 mV in amplitude. Therefore, one can think of simple neutron counting system by feeding diamond detectors signals directly to the low threshold discriminators. The results obtained on the VR1 will be described and discussed in detail in the paper and associated presentation. The results demonstrate that diamond micro-fission chambers can be used for in-core neutron monitoring, where robust, simple and compact devices are required.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Development, calibration and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Quantitative void fraction measurement with an eddy current flowmeter for generation IV Sodium cooled Fast Reactor

This study was carried out to understand the response of an eddy current type flowmeter in two phase liquid-metal flow. We use the technique of ellipse fit and correlate the fluctuations in the angle of inclination of this ellipse with the void fraction. The effects of physical parameters such as coil excitation frequency and flow velocity have been studied. The results show the possibility of using an eddy current flowmeter as a gas detector for large void fractions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development, calibration, and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Division at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Pulsed-Neutron Experiments at the Inherently Safe Subcritical Assembly

The pulsed neutron technique is a powerful, dynamic method to assay the reactivity of a multiplying system. This work presents the novel application of the pulsed neutron technique to the Inherently Safe Subcritical Assembly, an experimental configuration accepted by the International Criticality Safety Benchmark Evaluation Project Handbook. The experiments were replicated with COG11.3, a continuous-energy Monte Carlo code. The pulsed neutron data were analyzed using the Sjöstrand and Gozani area-ratio methods and by extracting the prompt neutron decay constant. Subsequent static k-eigenvalue and 𝛼-eigenvalue simulations were also performed for the same configurations. Neutron detector dead-time effects from the experiments were corrected using the Backwards Extrapolation Method and shown to have a negligible impact on the estimated reactivities. The results highlight that capturing time-dependent effects like delayed neutron precursor buildup are essential to accurately reproduce experimental results. They also highlight the importance of shielding the detectors from generator source neutrons in deeply subcritical configurations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Extensibility of the fission surface power (FSP) system from the moon to Mars

Fission reactors have great near-term potential to power human and robotic missions/outposts on the surface of the Moon and Mars (and potentially other planets, moons, and asteroids). The ability to provide a power-rich environment that is independent of solar intensity, nights, dust storms, etc., is of significant (perhaps enabling) importance to the further expansion of humans into our solar system. NASA's Reference Fission Surface Power (FSP) System is a 40 kWe system that has been primarily designed for lunar applications. This paper examines the extensibility of the FSP design and technology for potential missions on Mars. Possible impacts include the effects of changes in heat sink, gravity, day-night cycles, mission transit time, communication delay, and the chemistry of the regolith and atmosphere. One of the biggest impacts might be differences in the potential utilization of in-situ materials for shielding. Another major factor is that different missions will likely require different performance requirements, e.g. power, lifetime and mass. This paper concludes that the environmental differences between potential mission locations will not require significant changes in design and technologies, unless performance requirements for a specific mission are substantially different than those adopted for the FSP The primary basis for this conclusion is that the FSP has been designed with robust materials and design margins.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Measurement of neutron spectra in the experimental reactor LR-0

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important task is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values. After the successful filter installing and a series of measurements, first test neutron radiography attempts with test samples could been carried out. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Simulation and control of a passively cooled small modular reactor Samet - 142

This work regulates a small modular reactor (SMR) with traditional control methodologies and analyzes its response to typical disturbances. An analytical model for the passively cooled NuScale SMR is developed to represent the essential dynamics. The MATLAB/Simulink based model consists of reactor core, steam generator, pressurizer, hot leg riser and downcomer, and turbine representations. The point kinetics equations with a single combined neutron precursor group and the lumped parameter representations with single-phase natural circulation account for the neutronics and thermohydraulics in the reactor core. A lumped parameter, moving-boundary approach forms the basis of the nonlinear model adopted for the once-through helical-coil heat exchangers in which boundaries between regions of different fluid states (i.e., subcooled, boiling, and superheated) vary over time. A proportional-integral-derivative controller employing a sliding-average temperature strategy commonly used for pressurized water reactors is implemented. The responses of the SMR to two different disturbances are analyzed and discussed in detail. For the first scenario, a 5% step increase in the load is applied to the system, which results in a change in the steam valve opening. For comparison, two different simulations under this same disturbance are run: one including and one excluding the control systems. The second scenario tests the effectiveness of the control system to increase (or decrease) the reactor thermal power to a certain level within a desired time period. The simulation results show that important system variables are kept at the desired values by the proposed control strategy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Estimation of LOCA break size using cascaded support vector regression - 334

A failure of emergency core cooling system (ECCS) during a certain small break loss of coolant accident (SBLOCA) can cause a severe core uncovery and fuel failure. The safety injection systems (SISs) might not function properly in case of a SBLOCA, due to the slight change of pressure in the pipes. Early trend of the major parameters has to be observed and the accurate information has to be given to the nuclear power plant (NPP) operators by precisely identifying the break size to effectively manage accidents. In this study, the objective is to provide to operators the information on the break size in a short time by considering the accident situations such as hot-leg LOCA, cold-leg LOCA, steam generator tube rupture (SGTR). A cascaded support vector regression (CSVR) is used in order to estimate the break size. The simulation data set was obtained from the optimized power plant reactor 1000 (OPR1000) using modular accident analysis program (MAAP) code. And genetic algorithm (GA) was used to select the input variables of the CSVR model and optimize its parameters. As a result of this study, the CSVR model estimate very well the break size of LOCAs. If the operators can predict the break size in the LOCA, they can response quickly and properly to LOCA circumstances to prevent the core uncovery and fuel failure. Also, it will be possible to more efficiently manage beyond design basis accidents. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Nuclear Power Plant Control Room Modernization - Simulation Based Design

As many plants age and equipment modifications are implemented, the industry has the opportunity to address the issue of control room obsolescence through technology application and enhanced engineering techniques using existing systems. This paper addresses the practice of Simulation Based Design (SBD) for control room modernization driven by plant modifications and upgrades.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

NRC technical basis for evaluation of its position on protection against common cause failure in digital systems used in nuclear power plants - 378

Digital technology has advantages over analog systems, including automated monitoring and alerts for standby safety functions, and predictive algorithms to maintain critical safety systems. Additionally, digital technology generally has higher reliability and can be designed to reduce single point vulnerabilities. For these reasons many nuclear plants have applied digital technology to safety and non-safety related applications, including reactor protection system, feedwater and turbine controls, etc. with a corresponding significant improvement in trip reduction. Nonetheless, digital instrumentation and control (I and C) systems also present potential new vulnerabilities that need to be assessed, including potential failures due to increased complexity of digital systems, the introduction of unique failure modes due to software (including software common cause failure (CCF)), and limited operating history of digital systems in nuclear safety related applications compared to analog systems. The fact that software is intangible means that common methods, such as analysis or testing, used for detecting CCF may not be effective when applied to software. Consequently, digital technology is perceived to pose a potential risk from the introduction of undetected systematic faults that could result in CCF. Despite the I and C system upgrades and modifications performed to date, the U.S. Nuclear Regulatory Commission (NRC) and industry stakeholders have identified the need to modernize the regulatory infrastructure to efficiently address risks associated with the use of digital technology for nuclear safety applications and address regulatory uncertainties. The NRC's current position on CCF is guided by the staff requirements memorandum (SRM) on SECY 93-087. The SRM provides specific acceptance criteria for the evaluation of CCF, which the staff implemented in the Branch Technical Position (BTP) 7-19. However, industry stakeholders have proposed using methods to characterize the likelihood of software CCF and eliminate it from further consideration in a defense-in-depth and diversity analysis. The NRC's current position does not consider these alternatives, and thus corresponding acceptance criteria is not currently available. The work discussed in this paper assesses the underlying technical basis associated with CCF, provides technical support for updating the NRC position and considers proposed methods for addressing potential CCF in digital systems while enhancing efficiency, clarity, and confidence. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Accident tolerant, non-oxidizable MOX fuel fabrication

Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Westinghouse Small Modular Reactor nuclear steam supply system design

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Improving Operational Efficiency, Compliance and Community Perception with Real-time Data - 19260

The nuclear and medical radiation industry has always faced a significant burden in effectively balancing exposure and productivity. Maintenance operations and development of new techniques require significant staffing resources to monitor and review personnel exposures and changes in operational radiation levels. Currently utilised monitoring technology and review techniques remain unchanged since the 1950's, relying upon mostly analogue instrumentation and paper reporting. After utilising isolated-area monitoring during a reactor decommissioning project it became apparent that current systems created false assurance as the units continually failed without warning due to 'dirty' power. These instrument failures could have easily led to an unnecessary exposure while operators continued to assume they were being safely monitored. This paper is a philosophical discussion on how current shortcomings could be addressed, security increased, and exposure risks minimised, by utilising real-time monitoring and automated reporting. The community and media expectations of reported data and nuclear monitoring systems continue to be coloured by sensational technology seen in movies and emerging technology available through smart homes and smart cities initiatives. For the general public to be more engaged with real radiation levels and effects, the nuclear industry needs to more effectively communicate safe levels. Organisations can have regain control of exposure levels discussions through the controlled release of background and expected exposure level data. An increase in real world monitoring data communicated in a controlled and more effective way will aid in the general public accepting, and being educated about, the minimal increase above natural background levels of the nuclear industry and other radiation facilities. This is particularly as Australia is in the process of implementing its Nuclear Waste Facility, with community acceptance the number one inhibitor of being able to move forward. Storage, mobile assets, and processing all present unique management, security, exposure, and telemetry challenges for a facility operations manager. Active monitoring and automated review of the monitoring data can alert operators to any unexpected events before any resultant increases in radiation levels cause an unplanned exposure to operators or the environment. Integration of radiation monitoring data with global positioning telemetry enables better organisational awareness of mobile asset location and early detection of any potential shifting and reduction in shielding integrity of sources and encapsulated waste during transport. Additionally, the real-time monitoring can be set-up to enhance an organisations' or nation's security capabilities. For instance, the real time monitor will send an automated alert in the case of an unexpected movement of a source through a monitored facility or for an unscheduled movement of a monitored vehicle. There are further benefits to be gained in control, ALARA principles and effluent management by increasing real-time monitoring capabilities through the stages of the waste cycle. Increasing operational effectiveness while delivering operational cost savings sounds like an attractive proposition. With increased access to operational area radiation data comes drastically improved real-time management of exposures for new, and experimental practices. Increased live data can be utilised to identify optimum practices for reducing operational and environmental exposures or releases. This increased operational data effectively comes with the added benefit of a reduction in the required health physics personnel resources for general operations. By reducing exposure and the detachment of required health physics resources from routine operations, there is greater availability for these resources to support maintenance operations and project-based work. The review of this increased data will create internal capability to identify low-- and high-risk activities, resulting in the ability to reduce the number of highly skilled workers in an industry where the availability of this talent is declining. During my time as the Radiation Safety and Nuclear Security Manager at a research facility, I implemented a real-time monitoring system. The system identified a secondary risk that had previously been unidentified and posed a higher exposure and security risk than the radiation practice being monitored. What stops industry from implementing technology that encourages agile work practices, while minimising security risks and personnel exposure? (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Calculation of Nuclear Reactor Cooling Tower Performance With Limited Data Streams

Monitoring of cooling tower performance in a nuclear reactor facility is necessary to ensure safe operation; however, instrumentation for measuring performance characteristics can be difficult to install and may malfunction or break down over long duration experiments. This paper describes employing a thermodynamic approach to quantify cooling tower performance, the Merkel model, which requires only five parameters, namely, inlet water temperature, outlet water temperature, liquid mass flowrate, gas mass flowrate, and wet bulb temperature. Using this model, a general method to determine cooling tower operation for a nuclear reactor was developed in situations when neither the outlet water temperature nor gas mass flowrate are available, the former being a critical piece of information to bound the Merkel integral. Furthermore, when multiple cooling tower cells are used in parallel (as would be in the case of large-scale cooling operations), only the average outlet temperature of the cooling system is used as feedback for fan speed control, increasing the difficulty of obtaining the outlet water temperature for each cell. To address these shortcomings, this paper describes a method to obtain individual cell outlet water temperatures for mechanical forced-air cooling towers via parametric analysis and optimization. In this method, the outlet water temperature for an individual cooling tower cell is acquired as a function of the liquid-to-gas ratio (L/G). Leveraging the tight tolerance on the average outlet water temperature, an error function is generated to describe the deviation of the parameterized L/G to the highly controlled average outlet temperature. The method was able to determine the gas flowrate at rated conditions to be within 3.9% from that obtained from the manufacturer’s specification, while the average error for the four individual cooling cell outlet water temperatures were 1.6 °C, -0.5 °C, -1.0 °C, and 0.3 °C.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Rapid assessment of structural and compositional changes during early stages of zirconium alloy oxidation

A multimodal chemical imaging approach has been developed and applied to detail the dynamic, atomic-scale changes associated with oxidation of a zirconium alloy (Zircaloy-4). Scanning transmission electron microscopy, a gas-phase reactor chamber attached to an atom probe tomography instrument, and synchrotron-based X-ray absorption near-edge spectroscopy were employed to reveal morphology, composition, crystal, and electronic structure changes that occur during initial stages of oxidation at 300 °C. Oxidation was carried out in 10 mbar O 2 gas for short exposure times of 1 and 5 min. A multilayered oxide film with a cubic ZrO adjacent to the oxide/metal interface, a nanoscopic transition region with a graded composition of ZrO 2-x (where 0 < x < 1), and tetragonal ZrO 2 in the outermost oxide were formed. Partitioning of the major alloying element (tin) to the oxide/metal interface and heterogeneously within the oxide accompanied the development of the layered oxide. Our work provides a rapid, high-throughput approach for detailed characterisation of initial stages of zirconium alloy oxidation at an accelerated time scale, with implications for several other alloy systems.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Investigation of therapeutic potentials of some selected medicinal plants using neutron activation analysis

Series of attempts were made to investigate concentrations of trace elements and their therapeutic properties in various medicinal plants. In this study, samples of some commonly used plants were collected from Bauchi State, Nigeria. They includes leaves of azadirachta indica (neem), Moringa Oleifera (moringa), jatropha curcas (purgin Nut), guiera senegalensis (custard apple) and anogeissus leiocarpus (African birch). These samples were analyzed for their trace elements contents with both short and long irradiation protocols of Instrumental Neutron Activation Analysis (INAA) at Nigerian Research Reactor-1 (NIRR-1) of Ahmadu Bello University, Zaria, Nigeria. The level of trace elements found varies from one sample to another, with some reported at hundreds of mg/Kg dry weight. The results have been compared with the available literature data. The presence of these trace elements indicates promising potentials of these plants for relief of certain ailments.

60 APPLIED LIFE SCIENCES↗