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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 OSTI2017

Model-free adaptive control for MHTGR-based nuclear steam supply systems - 144

Modular high temperature gas-cooled reactor (MHTGR) uses helium as coolant and graphite as both moderator and structural materials. Due to its inherent safety feature, MHTGR has been regarded as an important candidate for next generation of nuclear plants. By coupling with the once-through steam generator (OTSG), MHTGR can be utilized to form the nuclear steam supply system (NSSS) for producing superheated steam, which can be adopted for nuclear plants with high efficiency. Power-level control of is important for the stable and efficient operation of the MHTGR-based NSSS, and it is necessary to develop the power-level control law that is robust to physical and thermal parameters and can guarantee the closed-loop stability in normal and even abnormal conditions. In this paper, a model-free adaptive power-level control law is developed for the MHTGR-based NSSS, which is free from physical and thermal-hydraulic parameters, takes the simple forms of proportional-integral (PI) or proportional-differential (PD) feedback law, and can provide globally asymptotic closed-loop stability. Then, it is applied to the NSSS control of the two modular high temperature gas-cooled nuclear plant HTR-PM. Numerical simulation results in normal power maneuver and some typical abnormal cases show that this new control law can provide satisfactory closed-loop stability as well as transient performance. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

The overview in safety review of human factors engineering and control room design in chinese AP1000 nuclear power plant - 174

The first AP1000 nuclear power plant is constructed in Sanmen county in Zhejiang Province in China, which has creative and distinctive design characteristics. Human factors engineering disciplines are applied to the design of the AP1000. In addition to the elements of the program review model(including human factors engineering program management, operating experience review, functional requirements analysis and allocation, task analysis, staffing, integration of human reliability analysis with human factors engineering, human system interface design, procedure development, training program development, human system interface verification and validation, design implementation, human performance monitoring),the minimum inventory of controls, displays, and alarms present in the main control room and at the remote shutdown workstation. These contents mentioned above are reviewed according to standards, rules and regulations. This article introduces review process, and several important issues during the reviewing are described in detail. These issues include many elements, such as operating experience review, task analysis, human system interface design, verification and validation,and so on. This paper emphasizes on the main control room design(including environment and layout). Background noise in the actual main control room may exceed too much to design value. Another important issue is about main control room habitability systems(VES) changes to satisfy post-actuation performance requirements in AP1000 design change proposal. All the wall panel displays will be closed in this design change proposal. Review process and proposal are described in the paper. Since the AP1000 unit in China is the first constructed all over the world, verification and validation are especially important and necessary. The plan of verification and validation, the results summary report of verification and validation, the report of human engineering discrepancies are focus attention. Verification and validation in human factors engineering include HSI task support verification, HFE design verification and integrated system validation. This paper introduces the review process of verification and validation, the issues found in verification and validation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Implementing a control application on an FPGA Platform - 356

Today, many I and C system suppliers propose FPGA-based I and C platforms. In that context, EDF has set up a project to assess the use of such platforms for safety-related systems and to enhance EDF's ability to manage this technology for the upgrade of existing I and C systems or for new constructions. One aim of the project is to assess the adequacy of these platforms for the development of control functions. This paper presents the development of a case study on a Radiy FPGA-based platform in a Hardware-In-the-Loop configuration, and the comparison with a Dymola Model-In-the-Loop and a Siemens T2000 S7 Hardware-In-the-Loop approach. The case study is a part of the level control of steam generators of French PWR NPP. The control flow diagram uses complex analog elementary blocks coded with fixed point format for the analog values. The use of IP cores of the FPGA supplier was avoided to enable a complete mastership of the VHDL code, to provide evidence of compliance with safety-related requirements, and to allow porting this code to another FPGA platform. The case study has highlighted the specificity of using a FPGA, and has shown that a complex control application in a FPGA-based PLC presents no major differences in precision and response time with a processor-based PLC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Introduction of Diverse Manual Actuation Design for APR1400

APR1400 is an Advanced Light Water Reactor (ALWR) design with enhanced plant safety and advanced design features. One unit is in commercial operation and 11 are currently under construction, 4 abroad and 7 in Korea. APR1400 adopts full digitalized Man-Machine Interface System (MMIS) to perform the function of control, indication and alarming for safety and non-safety systems APR1400 MMIS is designed to operate the plant from the Main Control Room (MCR) under all anticipated operating conditions including design basis accidents. This advanced MMIS is designed to meet the latest codes, standards and regulatory requirements including the defense-in-depth and diversity requirements, which is intended to mitigate the consequence of a total failure of the digital safety system caused by a software Common Cause Failure (CCF) coupled with anyone of the Design Basis Events. The redundancy of the safety system could be defeated during this CCF condition and the protective function could be rendered inoperable.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Sensitivity Analyses of Natural Convection in the AP1000 Passive Containment Cooling System Following LBLOCA Using CFD

The AP1000 power plant has implemented multiple layers of passive safety systems to enhance reactor safety and ensure system integrity during postulated scenarios, such as loss of coolant accident (LOCA) and main steam line break (MSLB). Among these, the passive containment cooling system (PCCS) is a safety-related system designed to prevent the containment from exceeding the design limits of both pressure and temperature following a postulated design basis accident, by transferring heat from the steel containment vessel to the atmosphere. During LOCA, natural air convection plays a vital role in removing heat from the steel containment vessel to the atmosphere. This is integrated with the condensate film from the Passive Containment Cooling Water Storage Tank (PCCWST). Air natural convection is also the only heat removal mechanism after the dry-out of the PCCWST and during the loss of shutdown decay heat removal (LOSDHR) event. Therefore, it is essential to investigate the thermal hydraulics behavior of the containment under transient conditions to ensure its safety and integrity. This paper presents a simplified CFD/ANSYS FLUENT model developed to analyze the impact of different parameters, such as air relative humidity, air temperature, and steel containment temperature, on the natural convection capability to remove the decay heat following LOCA. The model aims to examine the thermal behavior of the containment and provides a comprehensive understanding of the system's natural convection capability during postulated accidents. The results obtained from the model can be used to improve the safety and integrity of the containment system and provide valuable insights for future research in the field.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

FIDES-II/P2M Simulation Exercise on AN3 and AN10 bump tests: main results

This paper presents simulations of bump tests conducted as part of the P2M (Power to Melt and Maneuverability) project within the OECD/NEA FIDES-II framework (Framework for Irradiation Experiments). To prepare fuel performance codes for analyzing the P2M experiments, an international Simulation Exercise (SE) was initiated under the P2M project. The objective of this exercise was to model the AN3 and AN10 bump tests carried out in the 1980s at the DR3 reactor (3rd RISØ Fission Gas Project), during which instrumented fuel rodlets (equipped with thermocouples and pressure sensors) were subjected to prolonged power transients and power dips. The ongoing simulation exercise involves seven organizations from industry, research, and regulatory bodies, each using different fuel performance codes. This paper provides detailed descriptions of the test cases and simulation results from six fuel performance codes (ALCYONE, BISON, FAST, FEMAXI, FRAPCON, and TRANSURANUS), with a particular focus on fission gas release kinetics.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Nuclear heating measurements by in-pile calorimetry: prospective works for a microsensor design

Since 2009 works have been performed in the framework of joint research programs between CEA and Aix-Marseille University. The main aim of these programs is to design and develop in-pile instrumentations, advanced calibration procedure and accurate measurement methods in particular for the new Material Testing Reactor (MTR) under construction in the South of France: Jules Horowitz Reactor (JHR). One major sensor is a specific radiometric calorimeter, which was studied out-of-pile from a thermal point of view and in-pile during irradiation campaigns. This sensor type is dedicated to measurements of nuclear heating (energy deposition rate per mass unit induced by interactions between nuclear rays and matter) inside experimental channels of MTRs. This kind of in-pile calorimeter corresponds to heat flux calorimeter exchanging with the external cooling fluid. This thermal running mode allows the establishment of steady thermal conditions inside the sensor to carry out online continuous measurements inside the reactor (core or reflector). Two main types of calorimeters exist. The first type consists of a single cell calorimeter. It is divided into a sample of material to be tested and a jacket instrumented with two thermocouples or a single thermocouple (Gamma Thermometer). The second, called a differential calorimeter, is composed of two superposed twin cells (a measurement cell containing a sample of material, and a reference cell to remove the heating of the cell body) instrumented with four thermocouples and two electrical heaters. Contrary to a single-cell calorimeter, a differential calorimeter allows the compensation of the parasite nuclear heating of the sensor body or jacket. Moreover, it possesses interesting advantages: thanks to the heaters embedded in the cells, three different measurement methods can be applied during irradiations to quantify nuclear heating. The first one is based on the use of out-of-pile calibration curves obtained by generating a heat source by the Joule Effect inside each calorimetric cell. The second one is a zero method consisting in cancelling the difference in cell responses with an additional energy into the reference cell. The last measurement method is based on current additions in the two calorimetric cells. However, one drawback of the existing differential calorimeter is the size of the sensor: a great length equal to 220 mm and a diameter equal to 18 mm. This current size leads to measurement limitations. This paper will begin with a presentation of these measurement limitations from a bibliographic state. Each limitation will be detailed and in particular in the case of a high nuclear heating level expected, for instance, inside the JHR's core at its highest nominal power. The second part of the paper will develop the scientific skills of each partner in heat sciences, micro technology and nuclear physics necessary to design a new calorimetric micro-system: the advantages of studied microelements such as micro-thermocouples, micro- fluxmeters and micro-heaters will be presented. The last part will discuss preliminary designs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Technical Challenges and Gaps in Integration of Advanced Sensors, Instrumentation, and Communication Technologies with Digital Twins for Nuclear Application

This paper explores integrating advanced sensor, instrumentation, and communication technologies with digital twin technologies for nuclear energy application. Digital twins and digital-twin-enabling technologies are expected to integrate with future nuclear reactor designs and have the potential to impact currently operating nuclear power plants. Greater digital integration, advanced instrumentation and control systems, and advanced operations and maintenance practices are all associated with digital-twin-enabling technologies. Advanced sensors, instrumentation, and communication technology are expected to comprise important elements of the infrastructure required to develop and operate a nuclear digital twin system. This paper identifies and discusses challenges and gaps in developing and implementing advanced sensors and instrumentation and communication technology to be integrated with a digital twin in current and advanced reactor applications. It is important to address some challenge and gap to enable a successful near-term deploying advanced sensors, instrumentation, and communication technologies integrated with digital twins.

Yadav, Vaibhav↗
DOE OSTI2016

Laboratory Experiments and Modeling for Interpreting Field Studies of Secondary Organic Aerosol Formation Using an Oxidation Flow Reactor

This grant was originally funded for deployment of a suite of aerosol instrumentation by our group in collaboration with other research groups and DOE/ARM to the Ganges Valley in India (GVAX) to study aerosols sources and processing. Much of the first year of this grant was focused on preparations for GVAX. That campaign was cancelled due to political reasons and with the consultation with our program manager, the research of this grant was refocused to study the applications of oxidation flow reactors (OFRs) for investigating secondary organic aerosol (SOA) formation and organic aerosol (OA) processing in the field and laboratory through a series of laboratory and modeling studies. We developed a gas-phase photochemical model of an OFR which was used to 1) explore the sensitivities of key output variables (e.g., OH exposure, O 3 , HO 2 /OH) to controlling factors (e.g., water vapor, external reactivity, UV irradiation), 2) develop simplified OH exposure estimation equations, 3) investigate under what conditions non-OH chemistry may be important, and 4) help guide design of future experiments to avoid conditions with undesired chemistry for a wide range of conditions applicable to the ambient, laboratory, and source studies. Uncertainties in the model were quantified and modeled OH exposure was compared to tracer decay measurements of OH exposure in the lab and field. Laboratory studies using OFRs were conducted to explore aerosol yields and composition from anthropogenic and biogenic VOC as well as crude oil evaporates. Various aspects of the modeling and laboratory results and tools were applied to interpretation of ambient and source measurements using OFR. Additionally, novel measurement methods were used to study gas/particle partitioning. The research conducted was highly successful and details of the key results are summarized in this report through narrative text, figures, and a complete list of publications acknowledging this grant.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

An automatic multi-precursor flow-type atomic layer deposition system

Designs for two automated atomic layer deposition (ALD) flow reactors are presented, and their capabilities for coating additively manufactured (AM) metal prints are described. One instrument allows the coating of several AM parts in batches, while the other is useful for single part experiments. To demonstrate reactor capabilities, alumina (Al 2 O 3 ) was deposited onto AM 316L stainless steel by dosing with water (H 2 O) vapor and trimethylaluminum (TMA) and purging with nitrogen gas (N 2 ). Both instruments are controlled by custom-programmed LabVIEW software that enables in situ logging of temperature, total pressure, and film thickness using a quartz crystal microbalance. An initial result shows that 150 ALD cycles led to a film thickness of ~55 nm, which was verified with Rutherford backscattering spectroscopy. This indicates that the reactors were indeed depositing single atomic layers of Al 2 O 3 per ALD cycle, as intended.

47 OTHER INSTRUMENTATION↗
DOE OSTI2011

Numerical and experimental calibration of calorimetric sample cell dedicated to nuclear heating measurements

Online nuclear measurements inside experimental channels of MTRs are needed for experimental works (to design set-ups) and for numerical works (input data) in order to better understanding complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuel. In this paper we focus only on one kind of measurements: nuclear heating performed by means of a radiometric calorimeter. The aims of numerical and experimental works are firstly to optimize the sensor response for new energy deposit ranges and then to miniaturize this sensor for JHR irradiation conditions A first calorimeter, developed previously by the CEA, is studied. It corresponds to a graphite differential calorimeter divided into two twin cells (a reference cell, and a sample one). It is used with a non adiabatic mode or heat flow mode. Experimental calibration of the sample cell is presented. In that case, energy deposit is simulated by Joule effect and the sample cell is inserted into a bath at a regulated temperature and controlled flow. The response of the sensor is shown versus electrical power imposed for two flow regimes (intensive or moderated forced convection). These experimental results are compared to numerical works and improvements are discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Ultrasonic imaging in liquid sodium: topological energy for damages detection

Under-sodium imaging at high temperature is an important requirement in sodium cooled fast reactors during structural inspection. It aims at checking the integrity of immersed structures and assessing component degradation. The work presented in this paper focuses on designing an advanced ultrasound methodology for detecting damages such as deep crack defects. For that purpose, a topological imaging approach was implemented. This method takes advantage of all available prior knowledge about the environment through the integration of differential imaging and time reversal techniques. The quality of the topological energy distribution (the image) is further enhanced by applying a time gating related to each reconstructed pixel. Numerical and experimental results are presented using this method in order to confirm its reliability. These images are compared to a B-scan to emphasize the localization performances of this method. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

An intelligent scanning algorithm for lost fuel rod recovery in the MYRRHA lower plenum

This paper presents a system for visualizing the interior structure of a vessel filled with a liquid metal. Due to the opaque and conductive nature of the medium, regular visualization systems based on cameras, electromagnetic or laser based techniques can't be used. The operation, strengths and weaknesses of an existing set-up are investigated by creating a scenario to find and extract a rectangular object inside the vessel in the shortest possible time. The main focus of this paper is not on designing a new system but to investigate the bottlenecks that are faced and need to be kept in mind in future designs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Test Article and Test Plan for Heat-Pipe Gap-Conductance Testing

The SPHERE Facility at INL is established to test phenomena related to heat pipes in nuclear applications. The current capability of the SPHERE testbed is operation of liquid alkali-metal-cooled heat pipes in a variety of atmospheres. The controlled atmosphere can be used to prevent oxidation of system parts and also to affect the thermal conductivity of gases for any interfacial gaps within the heat-pipe assembly. Currently available modes are vacuum, air, nitrogen, helium, and argon. Controlled mixtures of helium and argon are available to vary the thermal conductivity of the gas. The primary test article will be based on the existing 7-hole block design used for shakedown testing of the SPHERE Facility. The SPHERE test stand will be used with test articles to establish the conductivity between fuel elements as well as between the heat pipe and other layers of the fuel-rod assembly. Existing and new test blocks will be used and modified with additional instrumentation to establish the thermal conductivity between the gap of the block wall and the heat pipes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Development of a Radiation-Tolerant Front End Digitizer

High-radiation and high-temperature environments represent notoriously detrimental conditions for electronics. Yet sensors and instrumentation are critically important for the safe operation of nuclear reactors and for the validity of the scientific experiments conducted within them. Traditionally, sensors in such extreme environments operate with few active components, resulting in small signals (mV–V, nA–µA range) that must be transmitted through long cable runs before digitization by sensitive data acquisition systems located in clean, less-harsh environments. Long cables are prone to picking up noise through electromagnetic interference and distortion caused by their parasitic impedances, resulting in signal degradation. The Front End Digitizer (FREND) system described and demonstrated herein seeks to address these issues by enabling the placement of a radiation-resistant analog front end within reactor containment, or other high-radiation environments. FREND will enable early signal pre-amplification, digitization, and optical encoding within the high-radiation field to transmit sensor signals over an optical fiber to the non-nuclear environment of an instrument room. Early signal pre-amplification and digitization maintain signal integrity, whereas optical transmission renders the system blind to induced electromagnetic noise over the long cable runs. This report describes the FREND initial prototype, which can multiplex and optically encode data from up to four sensor inputs, though, in principle, this can be extended to any number of sensor inputs. The end result is an extremely flexible, high-fidelity system.

42 ENGINEERING↗
DOE OSTI2017

The effect of support functions in a computerized procedure system on interface management tasks - 235

By adapting digital and computer technology, the design of main control rooms (MCR) in nuclear power plants has been changed. When operators operate a nuclear power plant, the tasks using soft controls in advanced MCRs include primary tasks and secondary tasks. Primary tasks are to control nuclear power plant devices and secondary tasks are to find appropriate screens or devices by screen navigations and selections before performing the primary task. The secondary task including interface management tasks is newly introduced in advanced MCRs. In this paper, the tasks are analyzed based on the simulation data of APR1400 which is one of the digitalized nuclear power plants. For analyzing tasks, the unit tasks are divided into three types: navigation, manipulation, and acknowledgement. The primary task represents the manipulation of devices and the secondary task indicates the navigating screens, opening control panels, closing control panels, and confirming the acknowledging alarms. To analyze the effect of support functions of a computerized procedure system (CPS), one of the functions which is expected to have a significant effect on the secondary tasks was considered, that is 'related screen link button.' Two factors, 'ratio of screen alteration' and 'ratio of using screen link button', were used to see the effect of 'related screen link button'. From the task analysis result, it was observed that the secondary tasks which is one of the unique features in advanced MCRs have a quite much portion than that of the primary tasks. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Temperature compensated transfer function for probability of detection (POD) in high temperature ultrasonic NDE using low temperature signals - 249

Ultrasonic piezoelectric based transducers offer the potential for in-coolant NDT measurements and imaging required at high temperature for the safe operation of advanced small modular reactors (SMRs). Current work estimates a significant 97% reduction in the probability of detection (POD) for an artificial flaw size of 2{lambda} due to degradation of the transducer performance at elevated temperature. The source of variability for POD in the current work is the temperature dependence of material matrix of piezoelectric material PZT-5A used in the sensor. The physics based model developed using finite element method quantifies reduction in the scattering amplitude for the side drilled hole (SDH) from 15 C to 195 C. Using model assisted probability of detection (MAPOD) approach, sensor temperature effect on the POD is demonstrated for SDHs in the low carbon steel specimen. Scattering amplitudes for the room temperature simulated data is validated with the experimental data. A temperature correction factor and transfer factors are reported to transform the simulated temperature effect from physics based model to the experimental data for estimation of high temperature POD using low temperature ultrasonic measurements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗