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At least 829 records · Page 46

High-Density Plasma Reactors: Simulations for Design

The development of improved and more efficient plasma reactors is a costly process for the semiconductor industry. Until five years ago, the Industry made most of its advancements through a trial and error approach. More recently, the role of computational modeling in the design process has increased. Both conventional computational fluid dynamics (CFD) techniques like Navier-Stokes solvers as well as particle simulation methods are used to model plasma reactor flowfields. However, since high-density plasma reactors generally operate at low gas pressures on the order of 1 to 10 mTorr, a particle simulation may be necessary because of the failure of CFD techniques to model rarefaction effects. The direct simulation Monte Carlo method is the most widely accepted and employed particle simulation tool and has previously been used to investigate plasma reactor flowfields. A plasma DSMC code is currently under development at NASA Ames Research Center with its foundation as the object-oriented parallel Cornell DSMC code, MONACO. The present investigation is a follow up of a neutral flow investigation of the effects of process parameters as well as reactor design on etch rate and etch rate uniformity. The previous work concentrated on silicon etch of a chlorine flow in a configuration typical of electron cyclotron resonance (ECR) or helical resonator type reactors. The effects of the plasma on the dissociation chemistry were modeled by making assumptions about the electron temperature and number density. The electrons or ions themselves were not simulated.The present work extends these results by simulating the charged species.The electromagnetic fields are calculated such that power deposition is modeled self-consistently. Electron impact reactions are modeled along with mechanisms for charge exchange. An bipolar diffusion assumption is made whereby electrons remain tied to the ions. However, the velocities of tile electrons are allowed to be modified during collisions and are not confined to a Maxwellian distribution. The interaction between the neutral flow and plasma is examined, and results for etch rate uniformity from the previous research and the present plasma simulations are compared.

Hash, David B.↗

Radiation hardened RF transceiver for in-containment environment applications using commercial off the shelf components - 154

The ability to use wireless devices and other electronics in a radiation environment such as Nuclear Power Plant containment is mostly limited by the radiation sensitivity of these electronics. The benefits derived from wireless technology such as the reduction of the amount of cable needed for new sensors, the reduction of the labor associated with cable routing and the reduction or elimination of containment penetrations for new sensors is significant and well understood; unfortunately these benefits have not been fully accomplished yet because of numerous challenges, one being the radiation environment. Functional operation and radiation effects of a transceiver, which utilizes an integrated circuit system-on-chip, with a micro controller, serial driver, phase-locked loop, crystal oscillators and an RF amplifier, will be described. The second type of transceiver that will be discussed uses discrete components such as JFETs, MOSFETs and op-amps to perform the signal conditioning and RF transmission is also described. Operation capabilities and radiation effects for these two approaches as well as energy sources will be described and compared in detail. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Structural Design and Modeling of MARVEL Primary Coolant System Using the ASME Section III, Division 5, Code

This paper presents the structural design and supporting analysis for the Microreactor Applications Research Validation and Evaluation (MARVEL) primary coolant system (PCS) using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III, Division 5, rules. MARVEL is a liquid metal–cooled microreactor intended to provide experimental capabilities for the rapid testing and development of microreactor technologies. The PCS utilizes high-temperature sodium-potassium liquid metal as the primary coolant and operates at a design temperature of 570°C, necessitating the consideration of creep-related failure mechanisms. The base metal for the PCS is 316H stainless steel, and the weldments are made with a 16-8-2 filler. The design approach incorporates the current base code rules along with ASME code cases N-924, N-861, and N-862 to address primary load, ratcheting, and creep-fatigue evaluations, respectively. The reactor’s operation involves complex thermal and mechanical interactions due to natural convective flow and differential thermal expansion between components. In conclusion, this paper discusses the structural engineering challenges encountered, such as managing thermal stresses in the distribution plenum and guard vessel, and outlines the strategies implemented to meet the code requirements, including design modifications and operational constraints.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

CORC $^{\circledR}$ cable terminations with integrated Hall arrays for quench detection

ReBCO superconducting cables have the potential to enable compact thermonuclear fusion reactors that operate at magnetic fields exceeding 20 T and allow operation at temperatures far exceeding the boiling point of liquid helium, potentially allowing for demountable magnets. Normal zone detection remains a challenge, and while novel quench detection techniques are an active area of research, few are non-invasive, provide real-time quench detection, and have been demonstrated with current ramp rates relevant for fusion reactors. To address this problem, a CORC$^{\circledR}$ cable termination is developed with integrated Hall sensors to monitor current redistribution as a proxy for quench detection. The methodology exploits the current sharing and layered topology in CORC$^{\circledR}$ cables, and allows quench detection using a localized sensor instead of co-wound voltage wires or optical fibers. Furthermore, experiments are presented where current redistribution is measured from induced quenches, and in a 0.2 meter CORC$^{\circledR}$ sample it is found that the Hall sensors detect normal zone transitions with a similar magnitude and temporal resolution as voltage measurements. To emulate the conditions of dynamic poloidal and central solenoidal fields, experiments are repeated with ramp rates up to 10 kA s -1 that demonstrate the potential to detect normal zone development over a range of experimental parameters.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Criticality Safety SU based USL Calculation for UCl3-NaCl Fuel Salt Operations

The Molten Chloride Reactor Experiment (MCRE) is a fast spectrum, molten salt fueled reactor that is planned to be constructed and operated in the Laboratory for Operation and Testing in the U.S. (LOTUS) testbed, formerly known as the ZPPR cell, at Idaho National Laboratory (INL). MCRE will provide valuable data to support design, licensing, and operation of full scale commercialized molten salt reactor designs.

99 - GENERAL AND MISCELLANEOUS↗

Demonstration of Pronghorn’s Subchannel Code Modeling of Liquid-Metal Reactors and Validation in Normal Operation Conditions and Blockage Scenarios

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

RBMK Graphite Serviceability at the Final Stage of Operation

A procedure for evaluating the technical conditions and residual service life of the RBMK-1000 graphite masonry and a program of research have been developed in order to evaluate the serviceability and the possibility of service life extension of the RBMK graphite stack at the final stage of operation. The serviceability criteria for the graphite stack, periodicity and methods of controlling them, which, in combination with predictive calculations have made it possible to validate the possibility of safe operation of the RBMK-1000 graphite stack above the assigned service life allowed to substantiate the possibility of safe operation of the graphite stack RBMK-1000 over a designated period of time, were determined.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Diversity for safety and security of NPP I and C: Post NUREG/CR 7007 stage - 306

The paper summarizes authors' experience on diversity aspects of NPP I and C systems. It also discusses diversity analysis aspects according to NUREG/CR 7007 and other relevant standards. Diversity implementation challenges related to regulatory authorities, developers, auditors and operation engineers and managers are formulated and discussed. The paper presents a case study and developed tool to support diversity analysis and assessment using post NUREG/CR 7007 approach. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Human Factors Considerations for Anticipated Uses of Autonomy in Advanced Reactor Technologies

The objective of this report is examine anticipated advanced reactor designs that incorporate high levels of autonomy in operation of advanced reactor systems or facilities, and to begin identifying the human factors engineering (HFE) considerations relevant to those designs. The insights gained will help identify areas where U.S. Nuclear Regulatory Commission (NRC) may need to develop or update review guidance to adequately address HFE considerations essential for safe operation. The adequacy of existing guidance, along with recommendations for updates or new guidance, may be addressed in future reports. We use the term “landscape analysis” to describe the review, analysis, and characterization of publicly available information on the anticipated future state of the industry regarding particular technologies or operational concepts. This landscape analysis considers the potential range of advanced reactor automation concepts that may be used for monitoring and controlling nuclear systems and facilities, including the potential for highly automated, near autonomous, or fully autonomous operation of advanced nuclear reactors.

99 - GENERAL AND MISCELLANEOUS↗

Operation of a 50-kW th chemical looping combustion test facility under autothermal conditions

The United States Department of Energy is developing transformational fossil fuel combustion technologies that result in more efficient power cycles and/or substantial reductions in GHG emissions. Chemical Looping Combustion (CLC) has the potential to achieve a lower cost of electricity than a supercritical pulverized coal power plant with an amine absorption system. Although the concept and research on chemical looping has a long history, CLC has numerous technical issues that must be addressed prior to a commercial debut. This effort is directed toward the following major issues for CLC: (1) demonstrating continuous solids handling, operability, and reactor performance at autothermal conditions and (2) preliminary assessments of oxygen carrier attrition and relative material make-up costs. These are critical issues that should be addressed under realistic conditions prior to directing resources toward larger scale demonstrations. This paper will also describe NETL’s small 50-kWth natural gas circulating CLC test facility located in Morgantown, WV. Despite its small size, this experimental test unit has operated at autothermal conditions for 11 h using a copper-iron-alumina oxygen carrier developed at NETL. Here, this paper will describe the test facility and operating experiences using NETL’s “Gen 2.0” oxygen carrier material. Natural gas conversion to CO2 ranged from 70 to 90% over approximately 75 oxidation-reduction cycles during the autothermal operating period. Estimates for the process specific attrition rate during autothermal operation are also presented. Finally, estimates for the oxygen carrier makeup cost, specific to the NETL process configuration, are several orders of magnitude greater than the NETL target. Future work should focus on reducing the oxygen carrier material cost and improving the attrition resistance of the oxygen carrier material under autothermal test conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Computerized procedure systems: procedure edition and maintenance challenges - 232

Very few nuclear power plants are using computerized procedures to guide the control room operation. A CPS implementation consists of two processes: digitalization and computerization. The former deals with the transformation of paper-based procedure into a digital data; the latter is referred to as the addition to that data of extra information that is meaningful for decision-making. This information may include real-time data (plant parameter values and binary process values to indicate equipment states), provide access to trends and historical data, etc. In order to be efficient and to optimize the digitalization and computerization processes, these set of aids should be implemented at the same time as operating procedures are digitalized There are references of plants having successfully implemented Computerized Procedures Systems (CPS) for emergency procedures, but when questioned about normal or abnormal operating procedures, operators and plant management seem reluctant to take that step. Given that a CPS is meant, by design concept, to increase operation crew efficiency by facilitating the procedures execution and reducing human error, why are we reluctant to enjoy CPS benefits in the most common plant status (i.e. normal operation)? Although there are many answers to this question, one is shared by most of the users: CPSs are not designed for editing and maintaining procedures easily and reliably. Normal operating procedures, in opposition to the emergency procedures, are often reviewed. Nowadays, these edition activities are performed easily with a word processor software and using paper, but what is required when dealing with a CPS? (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Status Report on Regulatory Criteria Applicable to the Use of Artificial Intelligence (AI) and Machine Learning (ML)

Although the interest in the use of artificial intelligence (AI) and machine learning (ML) in nuclear energy is increasing rapidly, at present their implementation is limited. This rapid increase in interest is not surprising considering that implementing AI and ML technology would allow for continuous monitoring, facilitate the implementation of predictive maintenance with optimized staffing plans, enable automation and autonomy opportunities that could drastically reduce fixed operation and maintenance costs, and provide training for operations and maintenance. Other industries are using AI for construction, and in the nuclear arena AI could provide great benefit in decommissioning activities. The ability of AI and ML to operate in real time vastly increases their potential impact. Before AI can be used in design, operations, or as a regulatory tool, the specifics on the regulations applicable to the use of AI for nuclear power applications need to be established. The difficulty is that the specific use cases will dictate the applicability of regulations. For example, even within the application domain associated with operations, the regulations might vary if the AI is used to create a virtual reference for plant operations or is used for training, optimization of maintenance intervals, prioritization of maintenance activities, etc. Different still is if the AI is to be used for design or setting technical specifications, which will introduce additional requirements. US Nuclear Regulatory Commission (NRC) licensing reviews are based on an applicant’s design meeting its performance assessment based on (1) safety goals and objectives, (2) deterministic and/or probabilistic analysis of accident scenarios, and (3) quantitative assessment of design alternatives against the safety goals and objectives using accepted engineering tools, methodologies, and performance criteria. The current regulatory framework does not explicitly address AI or autonomous control. However, as implementing AI technology will require the use of a digital platform, it must meet the requirements of an instrumentation and control (I&C) system. The regulatory requirements for AI, which will be incorporated into the I&C system, will be very dependent on how it is used (i.e., its functionality, safety classification, etc.). The licensing process is primarily risk-based with the identification of components and systems as nonsafety, important to safety, or safety related. A risk-informed approach allows further gradation of components and systems based on risk metrics such as core damage frequency or large early release fractions. Thus, the use cases and the risk categorization of impacted systems and components will determine the regulatory requirements. Regardless of how AI is used it presents new opportunities for risk-informing operating, maintenance, and regulatory decisions. Trustworthiness, transparency, and the ability to validate and verify the results will be paramount in showing that the systems and plant still meet their performance requirements. This report describes the results of research to identify regulatory implications of AI technologies and their uses. Specifically, this report reviews current regulatory guidance relevant to the application of AI for design (including design changes or new designs including advanced reactors), construction, operations, training, maintenance, research, testing, and as a regulatory tool. AI can be automated at different levels from purely informative purposes to autonomous controls. The focus of this review included determination of constraints on the application of AI technology, identification of any regulatory gaps or uncertainties, and clarification of anticipated technical basis information likely to be important for regulatory acceptance of these technologies. Currently, any use of AI at nuclear power plants is focused on nonsafety-related applications. The NRC and other regulatory bodies are evaluating providing guidance to address gaps rather than create new regulations to address the use of AI and ML. This approach seems to be the best to encourage AI development without adding regulatory uncertainty.

97 MATHEMATICS AND COMPUTING↗

Development and Scale-Up of Copper-Based Chemical Looping with Oxygen Uncoupling

The University of Utah has been researching chemical looping with oxygen uncoupling of coal since 2007, focusing on copper-based carriers and a system configuration based on two interconnected circulating fluidized bed reactors. R&D has combined fundamental, lab-scale systems with operation of CLOU reactor systems at various scales and computational simulations for performance evaluation and reactor scale-up. Development has progressed from small lab scale through bench scale with a 10 kW dual bed research reactor to semi-pilot scale with a recently-inaugurated 220 kW chemical looping combustion process development unit. One of the challenges of scaling up copper-based CLOU technology is acquiring enough oxygen carrier material to support operation in larger reactor systems. The UofU CLC PDU, for example, requires 150-200 kg of inventory, so the oxygen carrier must be readily available in large quantities at comparatively low cost. Because no supplier capable of delivering 1+ tons of copper-based oxygen carrier material could be identified, an inexpensive oxygen carrier material was developed in-house and the carrier production process was scaled to produce large quantities of material sufficient for the CLC PDU. This presentation describes development of the CLOU process at the University of Utah, including evaluation of fundamental processes key to performance of CLOU, modeling and scale-up of the process, large-scale production of copper based oxygen carrier material and experience to date with the world’s largest dedicated CLOU reactor.

Whitty, Kevin J.↗

An operation verification strategy against network attack of DCS in nuclear power plant - 323

Digital Instrumentation and control (I and C) system serves as the 'senses and nervous system' of a nuclear power plant (NPP). It plays an important role in the safe and economic operation in the NPPs. Digital I and C system has a series of advantages compared with analogue ones, but the introduction of the operation system platform, the application software, and the network communication structure are causing security risk to the power plant. Different kinds of procedures have already been adopted against these risks such as the communication isolation, but the security attacks still occur from time to time in the industrial fields around the world. This paper presents two computer security methods based on the thought of software verification and validation with step-by-step operation identification. First, three scenarios of security attack are proposed according to the structure of a distributed control system (DCS). Then a method with two network bypasses and a security computer are adopted to obtain the operation commands, the actuation signal, and the operation feedback in case of the first two scenarios by identifying the consistency of the obtained data. Furthermore, the method is optimized by combining the operation procedure and the camera shot to identify the anomaly of the information flow, thus to improve the effectiveness against the computer security issues. In the end, the influence and feasibility of this method is discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

The influence of changes in the VVER-1000 fuel assembly shape during operation on the power density distribution

A new approach to calculation of the coefficients of sensitivity of the fuel pin power to deviations in gap sizes between fuel assemblies of the VVER-1000 reactor during its operation is proposed. It is shown that the calculations by the MCU code should be performed for a full-size model of the core to take the interference of the gap influence into account. In order to reduce the conservatism of calculations, the coolant density and coolant temperature feedbacks should be taken into account, as well as the fuel burnup.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Destructive and Non-Destructive Characterization of various LLW/ILW at NRG - 19078

The Nuclear Research and consultancy Group is located at Petten, the Netherlands, and operates the High Flux Reactor (HFR). The HFR is used for isotope production for medical purposes and material research. During its exploitation several types of radioactive waste are produced, which partly have accumulated at the reactor site. In addition, NRG operated the Low Flux Reactor (LFR), a small 30 kWmax Argonaut type reactor which was used for research and education purposes. Over the past decade several projects on waste management and decommissioning were carried out requiring various approaches, including the radiochemical characterisation of the different emerging waste streams as required by the Dutch regulator. This paper focuses on the practices and technical aspects of the destructive and non-destructive characterization of various types of low- and intermediate-level waste arising from a number of nuclear activities performed at the premises of NRG, i.e. the Dutch legacy waste management program, the treatment of ion exchange resins, and the decommission of the Low-Flux Reactor. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A reactive separation process for pre-combustion CO 2 capture employing oxygen-blown coal gasifier off-gas

In this paper, we present an experimental study of a reactive separation system, consisting of a membrane reactor (MR) and an adsorptive reactor (AR) operating in tandem, with the MR’s reject stream serving as the AR’s feed. We investigates the feasibility of applying this MR-AR system for high-purity H 2 production and simultaneous CO 2 capture via the water gas shift (WGS) reaction in the context of Integrated Gas Combined Cycle (IGCC) power generation employing oxygen-blown gasifier syngas from biomass and coal. We previously studied this MR-AR system for the IGCC process employing air-blown gasifier syngas, for which it demonstrated good performance, attaining high conversion exceeding equilibrium, producing ultra-pure H 2 for power generation, and a CO 2 stream ready for sequestration. In this study, we focus on oxygen-blown gasifier off-gas that contains no N 2 , with composition distinctly different from the air-blown gasifier syngas which has a large N 2 content. We employ a carbon molecular sieve membrane (CMSM), a commercial sour-shift WGS catalyst, and a hydrotalcite (HTC) adsorbent. We carried out experiments to determine membrane performance and to identify promising operating conditions in an IGCC-relevant environment. The CMSM proved robust during a long-term (~344 hr run) experimental run under high temperature and pressure maintaining a high He/N 2 selectivity (~170). Multi-cycle runs were carried out during which the MR-AR system displayed superior performance to that of a PBR, by producing a high purity H 2 product directly usable in power generation. Therefore, the findings from this study demonstrate the ability of the MR-AR system to operate stably for a broad range of gasifier off-gas compositions, and indicate its potential for integration into IGCC plants for power generation with CO 2 capture.

01 COAL, LIGNITE, AND PEAT↗