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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 757 records · Page 42
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DOE OSTI2015

Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR), under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the calorimeter were carried out. A preliminary analysis shows that the numerical results overestimate the measurements by about 20 %. A new approach has been developed in order to estimate the nuclear heating by two methods (energy deposition or KERMA) by considering the whole complete geometry of the sensor. This new approach will contribute to the interpretation of the irradiation campaign and will be useful to improve the out-of-pile calibration procedure of the sensor and its thermal response during irradiations. The aim of this paper is to present simulations made by using MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) for the nuclear heating inside the different parts of the calorimeter (head, rod and base). Calculations into two steps will be realized. We will use as an input source in the model new spectra (neutrons, prompt-photons and delayed-photons) calculated with the Monte Carlo code TRIPOLI-4{sup R} inside different experimental channels (water) located into the OSIRIS periphery and used during the CARMEN-1P irradiation campaign. We will consider Neutrons- Photons-Electrons and Photons-Electrons modes. We will begin by a brief description of the differential-calorimeter device geometry. Then the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements will be introduced. The energy deposition due to the prompt-gamma, delayed-gamma and neutrons, the neutron-activation of the device will be considered. The different components of the nuclear heating inside the different parts of the calorimeter will be detailed. Moreover, a comparison between KERMA and nuclear energy deposition estimations will be given. Finally, a comparison between this total nuclear heating Calculation and Experiment in graphite sample will be determined. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Operational performance risk assessment in support of a supervisory control system - 251

A supervisory control system is being developed for multiunit advanced small modular reactors to minimize human interventions during normal and abnormal operations. In the supervisory control system, control action decisions are made based on a probabilistic risk assessment that employs event trees and fault trees. Although traditional probabilistic risk assessment tools are implemented, their scope is extended to normal operations, and the application is reversed to assess the success of non-safety related systems and to enable continued operation of the plant. This extended probabilistic risk assessment approach is called operational performance risk assessment (OPRA). OPRA helps to identify available paths, combine control actions for maintaining plant conditions within operational limits, and to quantify the likelihood of success of these operational trajectories to optimize the selection of alternative actions without activating reactor protection system. In this paper, a case study of OPRA in a supervisory control system is demonstrated for the Advanced Liquid Metal Reactor (ALMR) Power Reactor Inherently Safe Module (PRISM) design, specifically the power conversion system. The scenario investigated involved a condition in which the feedwater control valve that was observed to be drifting to the closed position. Alternative plant configurations that would allow the plant to continue to operate at full or reduced power were identified using OPRA. Dynamic analyses were performed with a thermal-hydraulic model of the ALMR PRISM system using Modelica to evaluate the magnitude of safety margins. Successful recovery paths for the selected scenario were identified and quantified using the supervisory control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Conceptual Design of Temperature-Controlled Fueled-Salt Irradiation Experiment to Support Demonstration of Advanced Nuclear Reactors

The irradiation testing of fuel-bearing molten salts is critical to supporting the development and demonstration of molten salt reactors (MSRs). These experiments can inform several important reactor design and safety parameters, including source term modeling, evolution of thermophysical properties with burn-up, and the degradation of structural materials under reactor relevant conditions. Idaho National Laboratory is designing an instrumented and heated high temperature molten salt-fueled irradiation capsule to study the behavior of the fuel salt during in-pile irradiation. This paper details the neutronics, thermal, and mechanical analysis performed to-date. Parametric studies are performed to assess a range of material, different experiment dimensions, two in-reactor positions, and the fuel enrichment used. Principal recommendations are to opt for a peripheral reactor position to alleviate neutronic constraints, a thin salt annulus to alleviate thermal constraints, and high-temperature alloys that provide additional safety margins.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Summary of LWRS Research in Addressing RPV Research Gaps in NRC EMDA Report

Reactor Pressure Vessels (RPVs) are critical components in nuclear reactors, housing the reactor core and coolant under extreme conditions of temperature, pressure, and radiation. These harsh environments contribute to the degradation of RPV materials over time, presenting challenges for extending reactor operations beyond their original design lifespans. The NRC Expanded Materials Degradation Assessment (EMDA) report volume 3 have been instrumental in guiding research to support extending the operational life of light water reactors (LWRs) up to 80 years. It provides a comprehensive framework to address technical challenges related to aging and degradation mechanisms in RPVs. The LWRS program has played a key role in advancing this research, supporting projects such as the UCSB ATR-2 Experiment, material testing from Zion and Palisades reactors, and the development of advanced mini-compact tension testing techniques. These efforts have been crucial in identifying and addressing gaps in our understanding of RPV aging, contributing to the successful subsequent license renewals of eight LWR units in the U.S. The EMDA report volume 3, built on the Phenomena Identification and Ranking Table (PIRT) analysis from earlier versions of the EPRI Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs), provides a detailed assessment of RPV degradation mechanisms. However, as EPRI has updated the MDM and IMT, it is important to revisit research priorities and methodologies to reflect these changes. The revised MDM and IMT may introduce new factors affecting long-term RPV performance and safety, highlighting the need for continued research and updated guidance to ensure the reliable and safe operation of reactors beyond 80 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Modeling of ultrasound transmission through a solid-liquid interface comprising a network of gas pockets

Ultrasonic inspection of sodium-cooled fast reactor requires a good acoustic coupling between the transducer and the liquid sodium. Ultrasonic transmission through a solid surface in contact with liquid sodium can be complex due to the presence of microscopic gas pockets entrapped by the surface roughness. Experiments are run using substrates with controlled roughness consisting of a network of holes and a modeling approach is then developed. In this model, a gas pocket stiffness at a partially solid-liquid interface is defined. This stiffness is then used to calculate the transmission coefficient of ultrasound at the entire interface. The gas pocket stiffness has a static, as well as an inertial component, which depends on the ultrasonic frequency and the radiative mass.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Migration of Older to New Digital Control Systems in Nuclear Power Plant Main Control Rooms

The United States (U.S.) Department of Energy (DOE) Office of Nuclear Energy (NE) has the primary mission to advance nuclear power by resolving socio-technical issues through research and development (R&D). One DOE-NE activity supporting this mission is the Light Water Reactor Sustainability (LWRS) program. LWRS has the overall objective to sustain the operation of existing commercial nuclear power plants (NPPs) through conducting R&D across multiple “pathways,” or R&D focus areas. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies pathway conducts targeted R&D to address aging and reliability concerns with the legacy instrumentation and control (I&C) and related information systems in operating U.S. NPPs. This work involves (1) ensuring that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) implementing digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Under the LWRS Advanced II&C pathway, Human Factors experts at Idaho National Laboratory (INL) have been conducting R&D in support of NPP main control room (MCR) modernization activities. Work in prior years has focused on migrating analog I&C systems to new digital I&C systems (). In fiscal year 2016 (FY16), one new focus area for this research is migrating older digital I&C systems to new and advanced digital I&C systems. This report summarizes a plan for conducting a digital-to-digital migration of a legacy digital I&C system to a new digital I&C system in support of control room modernization activities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Development of thermal transit flow measurement for small modular reactors - 140

Safe, reliable, and efficient operation of nuclear reactors requires timely and accurate measurement of primary coolant parameters such as temperature, flow, pressure, and level. These measurements may be challenging in small modular reactors (SMRs) with integral vessel design. For SMRs with the integral design, sensors may be located in confined spaces or embedded within pressure vessels. The absence of primary coolant loops is another fundamental design attribute that will affect the location of sensors, sensor interfaces, and types of sensors suited for process parameter measurements. Further, the number and location of pressure vessel penetrations may be constrained compared to traditional nuclear power plants (NPPs). As a result, conventional NPP temperature, flow, pressure, and level sensors may not be appropriate for use in SMRs. This paper addresses the development of new sensor technology for hybrid temperature, flow, and level measurement. The authors demonstrate the alternative use of resistance temperature detectors (RTDs) where cross-correlation of process temperature variations between RTDs is used to calculate process flow velocity. The new sensor and measurement technology will also incorporate RTD response time and self-heating characteristics to provide flow and level measurement in accident conditions. This capability would simplify reactor design and enable optimal placement of reactor pressure vessel penetrations leading to increased operational reliability and safety. The cross-correlation of temperature signals and adaptation for flow measurement in a laboratory setting is presented here. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Real-time supervisory control implementation of SmAHTR power plant - 143

This paper discusses an implementation of a supervisory control system for a nuclear power plant model using Small modular Advanced High Temperature Reactors (SmAHTR). The simulation operates in real-time using a network of computer agents, implemented on Raspberry Pi's (RPi's), where each major task is performed by one or more RPi's running calculations in parallel. The system architecture contains local, module, master, and operator actions to automate all aspects of the plant. Local level controllers run individual control loops to ensure the desired reference signals are met. The module level controllers set the local level reference signals within each SMR unit and detect sensor and actuator failures to adjust those reference signals as needed. Master level controllers balance plant power by calculating required heat power from each reactor and electric power from each generator and sending that data to the module controllers. Finally, the operator level contains an operator interface to summarize data and faults gathered through the controller levels, with options to review plant and reactor details as desired. This control implementation allows the plant to run autonomously, load follow a desired electrical power curve, and detect a growing number of fault scenarios. Ultimately, this type of supervisory control system will reduce labor-intensive surveillance and testing and allow fewer operators to more safely monitor an entire plant. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

A large scale system approach for stability analysis of multi-unit SMR plants - 248

In this paper we present a framework based on large scale system theory for analyzing the stability of coupled advanced high temperature molten salt small modular reactors. It has been suggested that small modular reactors be coupled through a common molten salt energy repository, called a salt vault. Using vector Lyapunov functions the connective stability of multiple reactors coupled through a salt vault can be guaranteed. In this theory, scalar Lyapunov functions are assigned to portions of the composite system state space. Functions are constructed for each small modular reactor (SMR), each Brayton cycle, and the salt vault. These functions are then aggregated, becoming components of a vector Lyapunov function. If this vector function possesses special properties, the composite system is stable. Bounds on the connective stability of a multi-reactor plant can be found in terms of a free parameter, in this case the mass of molten salt in the salt vault. Examining bounds on the connective stability allows one to use control theory as a tool for making performance and economic design decisions related to the size of the salt vault. This technique can produce a constraint that is used in a design optimization problem for finding the smallest-possible salt vault that ensures stability. It can also be used as an indicator of interconnection and its effect on system behavior. This, in turn, could be used to guide multi-unit control system design to ensure that highly coupled systems behave favorably. The theory for multi-unit control and operation is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Verification of severe accident management guideline (SAMG) entry condition for OPR1000 - 367

The current severe accident prevention and mitigation strategies for OPR 1000 have emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs), respectively. The main objective of SAMG is to prevent a release of radioactive materials from releasing into environment and mitigate severe accident phenomena during severe accident. The current SAMG entry condition is met when the core exit temperature (CET) reaches 650 deg. C in OPR 1000. Peak cladding temperature (PCT), which represents core integrity more precisely than CET does, cannot be used since it is not measurable in the main control room (MCR). Current SAMG entry condition for every kind of accident scenarios is not reasonable, since entry condition is determined without considering operator action time. Therefore, SAMG entry condition need to be verified with consider operator action time. Thus, an analysis methodology for SAMG entry condition was developed to verify the SAMG entry condition. A severe accident DB were also developed in order to analyze the SAMG entry condition. The analysis results of DB show that available operator action time from the SAMG entrance to RV fail is shorter than the required time in most cases. As a result, the proper SAMG entry condition was suggested through the results of DB analysis. When the SAMG entry condition is changed from CET to suggested condition, the available action time is increased about 350 to 4300 sec. Also, the suggested entry condition was verified through the simulation which considered operation action time and reflecting the suggested entry condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of Performance Validation Tool for the NPCS for APR1400

Availability of nuclear power plant is highly dependent on the quality of major control systems including NSSS Control Systems. It has been experienced that a small flaw hidden among the control functions resulted in large operational transient or even reactor trip. The proactive quality control of these systems is not easy because of the dynamical characteristics of control algorithms. In a stand-alone or open loop state, it is not possible to perform integral system tests because they need control loads to form an appropriate closed control loop. In the construction plant, Power Ascension Test (PAT) is the first chance to witness the overall performance of the control systems. Because the PAT is the last activity of plant construction schedule, the cost of corrective actions is very high if any functional defect is found in this stage. To minimize risk caused by late detection of control system defects, we developed a validation tool as a proactive test device for the NSSS Process Control System (NPCS) for the SKN Unit 3 which is the first APR1400 plant in Korea. In this paper, the features of the dynamic performance test tool, test methodology, and the test results are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Temperature-programmed deoxygenation of acetic acid on molybdenum carbide catalysts

Temperature programmed reaction (TPRxn) is a simple yet powerful tool for screening solid catalyst performance at a variety of conditions. A TPRxn system includes a reactor, furnace, gas and vapor sources, flow control, instrumentation to quantify reaction products (e.g., gas chromatograph), and instrumentation to monitor the reaction in real time (e.g., mass spectrometer). Here, we apply the TPRxn methodology to study molybdenum carbide catalysts for the deoxygenation of acetic acid, an important reaction among many in the upgrading/stabilization of biomass pyrolysis vapors. TPRxn is used to evaluate catalyst activity and selectivity and to test hypothetical reaction pathways (e.g., decarbonylation, ketonization, and hydrogenation). Furthermore, the results of the TPRxn study of acetic acid deoxygenation show that molybdenum carbide is an active catalyst for this reaction at temperatures above ca. 300 °C and that the reaction favors deoxygenation (i.e., C-O bond-breaking) products at temperatures below ca. 400 °C and decarbonylation (i.e., C-C bond-breaking) products at temperatures above ca. 400 °C.

09 BIOMASS FUELS↗
DOE OSTI2020

Sensor Drift Estimation for Reactor Systems by Fusing Multiple Sensor Measurements

A nuclear power plant is instrumented with a variety of sensors that monitor its variables to estimate the state and initiate safety actions, if needed. We address the problem of estimating drifts or errors in sensor measurements due to factors such as calibration changes. We propose an information fusion method that uses measurements from other sensors to generate an estimate of a sensor measurement, and its difference from an actual measurement provides an error estimate. We present two fusers based on the ensemble of trees and support vector machine that are trained using sensor measurements collected at an emulated test loop of a pressurized water reactor under no-drift conditions. We present error estimates for a differential pressure sensor of the heat exchanger of the primary coolant system, under twenty controlled scenarios using the test loop. Both positive and negative errors are captured by both methods in scenarios involving calibration drifts, blocking, air gap and electromagnetic interference. The root mean square errors of estimated drifts are typically within 2% percent of the maximum.

Rao, Nageswara↗
DOE OSTI2017

A SUMMARY COMPARISON OF DESIGN EVALUATION TECHNIQUES

The United States’ fleet of Nuclear Reactors is considering modernizing their control rooms and instrumentation as an effort to avoid component obsolescence, keep up with competing industries, and cater to a new work force among others. Multiple technologies have potential to improve the control room. In attempts to evaluate the available options researchers employ a variety of measures to ensure the best candidate is used in a modernizing effort. The NRC is in charge of ensuring any new design has been thoroughly vetted before approving a license. Laid out in NUREG-0711 “Human Factors Engineering Program Review Model” is the review criteria for design validation. As there are a variety of measures currently used to evaluate candidate technologies, this paper seeks to identify weaknesses in the common measures used to design and validate technology in control room modernization efforts. Identifying where measurements are not meeting the criteria will help the control room modernization research platform identify where measurement development needs to occur. This will help ensure that current performance measurements are producing as reliable results as possible to select the right technology to integrate into nuclear operating control rooms.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

The study on overall I and C system scheme for floating nuclear power plant ACP100S and ACP25S - 189

Floating nuclear power plant (FNPP) is built on the offshore platform and it is defined as a kind of Small and Modular Reactor (SMR) due to its power level. The FNPP can provide power supply, fresh water and high temperature steam and other products. So it can be used as special energy sources, such as regional power supply, district heating and offshore oil exploitation. At present, two types of FNPP, named ACP100S and ACP25S, are developed by China National Nuclear Corporation (CNNC). In this paper, the overall technical parameters and overall technical scheme of ACP100S and ACP25S are introduced firstly. On the basis of these schemes, the different arrangements of nuclear power plant which are used for meeting the different energy demands are discussed. Due to the different configuration of nuclear power plant, the requirements for the Instrument and Control (I and C) system are different, so the paper discusses some overall configurations of I and C system under the several typical arrangements models, i.e., the single reactor and the single turbine, the double reactors and double turbines, and the double reactors and the single turbine. Furthermore, the characteristics of these overall I and C configurations are analyzed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Study of the open loop and closed loop oscillator techniques

This paper presents some aspects of a five year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques. The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this paper we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign to measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Experimental and numerical validation of an ultrasonic visualization system for nuclear reactor application

This paper deals with the design of ultrasonic imaging systems for the next generation of nuclear reactors cooled by liquid metal. Indeed, a generation IV research nuclear reactor is being developed by the Belgian Nuclear Research Center (SCK-CEN) in the frame of the MYRRHA project (for Multipurpose hYbrid Research Reactor for High-tech Applications). This sub-critical/critical reactor is cooled by Lead-Bismuth Eutectic (LBE). The opacity of this liquid metal requires the development of an ultrasonic visualization system for internal inspection and object detection. But due to the peculiar conditions met in the core of the reactor, velocity as well as temperature gradients are expected and are likely to affect directly the ultrasonic propagation. The objective of this work is to validate the ultrasonic imaging strategy by tests performed in a dedicated test rig and by numerical simulations using a ray-tracing method. The experimental investigations have been performed on a specific water facility reproducing conditions similar to those encountered in the core of the MYRRHA reactor. These conditions include the propagation over large distance, and the presence of temperature and velocity gradients. In the MYRRHA reactor application, the distance to be travelled by the acoustic waves of the visualization system is about 5 m, including a reflection. The acoustic absorption, the scattering losses, the beam divergence and the transmitted energy during reflection have been determined as a function of the travelled distance. The experimental values are compared with the literature for validation. The presence of temperature and velocity gradients in the core of the reactor is due to the coolant circulation. These gradients are about 5 K over 0.1 m and 1 m/s over 0.2 m, respectively, and are reproduced in the facility for investigating their influence on the propagation of ultrasounds. The experimental data are used for improving and validating a ray-tracing algorithm. This algorithm is then applied to the full scale application for validating the ultrasonic visualization system for nuclear reactor application. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗