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NASA NTRS1979

Phase 1 of the First Small Power System Experiment (engineering Experiment No. 1). Volume 2: System Concept Selection

The development of a modular solar thermal power system for application in the 1 to 10 MWe range is presented. The system is used in remote utility applications, small communities, rural areas, and for industrial uses. Systems design and systems optimization studies are conducted which consider plant size, annual capacity factors, and startup time as variables. Investigations are performed on the energy storage requirements and type of energy storage, concentrator design and field optimization, energy transport, and power conversion subsystems. The system utilizes a Rankine cycle, an axial flow steam turbine for power conversion, and heat transfer sodium for collector fluid.

Holl, R. J.↗

DOE OSTI2016

Coordinated control design and verification of the HTR-PM plant - 18485

Through multi-modular scheme, i.e. the superheated steam flows produced by multiple MHTGR-based nuclear supplying system (NSSS) modules are combined together to drive a common thermal load, the inherent safety feature of MHTGR can be applicable to the large-scale nuclear plants at any desired power ratings. Since the plant power control technique of classical single-modular plants cannot be directly applied to multi-modular plants, it is necessary to develop the power control method of the multi-modular plants. In this paper, the plant power control strategy of the two-modular HTR-PM plant, i.e. the first under-constructed multi-modular high temperature gas-cooled nuclear plant in the world is given, which includes both the feedback loops and control algorithms. Then, the control strategy is verified through numerical simulation, and the simulation results in the cases of plant power step and ramp show the satisfactory performance of this control strategy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Separation and conversion of carbon dioxide to syngas using a porous ceramic dual membrane in a thermo-electrochemical reactor

A thermo-electrochemical reactive capture apparatus includes an anode and a cathode, wherein the anode includes a first catalyst, wherein the cathode includes a second catalyst, a porous ceramic support positioned between the anode and the cathode, an electrolyte mixture in pores of the ceramic support, and a steam flow system on an outer side of the cathode. The outer side of the cathode is opposite an inner side of the cathode and the inner side of the cathode is adjacent to the ceramic support. In addition, the electrolyte mixture is configured to be molten at a temperature below about 600° C.

Campbell, Patrick↗

DOE OSTI2016

Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator

The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have been completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

High-Fidelity Simulation of Flow-Induced Vibrations in Helical Steam Generators for Small Modular Reactors

Flow-induced vibration (FIV) is a widespread problem in energy systems as they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occur. Given the importance of reliable components in the nuclear industry, FIV has long been a major concern in the safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from FIV and related failures. In this paper we discuss the use of the computational fluid dynamics code Nek5000 coupled to the structural code Diablo to simulate the flow in helical coil heat exchangers and associated FIV. In particular, one-way coupled calculations are performed, where pressure and tractions data are loaded into the structural model. The main focus of this paper is on validation of this capability. Fluid-only Nek5000 large eddy simulations are first compared against dedicated high-resolution experiments. Then, one-way coupled calculations are performed with Nek5000 and Diablo for two data sets that provide FIV data for validation. These calculations were aimed at simulating available legacy FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient since the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral pressurized water reactors. We discuss validation of two-way coupled experiments and benchmarks toward the simulation of fluid elastic instability. We briefly discuss the application of these methods to grid-to-rod fretting.

Flow-Induced Vibration↗

DOE OSTI2016

Experimental Study for Two Phase Flow Instability in the PCHE Type Steam Generator in SMR

Conventional U-tube type heat exchangers have some limitations when applied to Small Modular Reactor (SMRs). In particular, the size of steam generators is important in the SMR, as it greatly affects the size, and thus the cost, of the reactor vessel.. Therefore, the circuit heat exchanger (PCHE) is suggested as an alternative. PCHE, a compact type heat exchanger manufactured by HEATRIC, has several advantages such as the reduced size, wide operation range until 900? and 900 bar and high power density due to the large heat transfer area over volume ratio. Recently, it is newly suggested as an application of a steam generator. MIT and KAIST have already conceptually suggested PCHE as a steam generator. Under boiling condition, flow instability is one of the critical issues-resulting in unstable mass flow rate, sudden temperature change and system control failure. According to the flow mechanism, there are various types of instabilities in the heat exchanger and each flow tendency is different. Typical examples are Ledinegg instability of the static instability and pressure-drop oscillation and density-wave oscillation of the dynamic instability. The characteristic of Ledinegg instability is that the flow under unstable condition changes into the stable flow. The characteristic of dynamic instability is that the flow fluctuates continuously. Pressure-drop oscillation has the period of tens of seconds and the system pressure also highly oscillates with the period of flow instability. In comparison, the density-wave oscillation has the period of few seconds and the system pressure is relatively stable. Previous studies on flow instability are focused on the single channel flow and helical tube type heat exchanger. There has been very limited experimental research on flow instability using PCHE. In KAIST, the test facility was constructed to study the thermal hydraulics in PCHE. The objectives of our test facility are: (1) to identify the occurrence of instability under two phase and super-heated steam flow in PCHE, and (2) to examine whether it can operate normally using throttling effect. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS1973

A high-speed photographic system for flow visualization in a steam turbine

A photographic system was designed to visualize the moisture flow in a steam turbine. Good performance of the system was verified using dry turbine mockups in which an aerosol spray simulated, in a rough way, the moisture flow in the turbine. Borescopes and fiber-optic light tubes were selected as the general instrumentation approach. High speed motion-picture photographs of the liquid flow over the stator blade surfaces were taken using stroboscopic lighting. Good visualization of the liquid flow was obtained. Still photographs of drops in flight were made using short duration flash sources. Drops with diameters as small as 30 micrometers (0.0012 in.) could be resolved. In addition, motion pictures of a spray of water simulating the spray off the rotor blades and shrouds were taken at normal framing rates. Specially constructed light tubes containing small tungsten-halogen lamps were used. Sixteen millimeter photography was used in all cases. Two potential problems resulting from the two-phase turbine flow (attenuation and scattering of light by the fog present and liquid accumulation on the borescope mirrors) were taken into account in the photographic system design but not evaluated experimentally.

Barna, G. J.↗

DOE OSTI2017

Numerical Analysis of Flow Instability in Once-through Steam Generator

Two-phase flow instability is a critical area to consider in the design of steam generators. Oscillation of the flow rate and the system pressure caused by heating may cause several problems and is unfavorable in an extreme case because it causes a burnout of the heat transfer surface. A large amplitude variation in the heat transfer tube wall temperature generally occurs under density wave oscillation (DWO) conditions, and a continuous cycling tube wall temperature may lead to damage to the tube, causing thermal fatigue problems. Therefore, instability of the flow should be avoided, and a two-phase flow system should be designed to have a sufficient margin to avoid this. A computer program was developed for predicting a two-phase flow instability in a steam generator heated by liquid sodium, and the results of numerical studies for several parameters are presented in this paper.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2011

Steam distribution and energy delivery optimization using wireless sensors

The Extreme Measurement Communications Center at Oak Ridge National Laboratory (ORNL) explores the deployment of a wireless sensor system with a real-time measurement-based energy efficiency optimization framework in the ORNL campus. With particular focus on the 12-mile long steam distribution network in our campus, we propose an integrated system-level approach to optimize the energy delivery within the steam distribution system. We address the goal of achieving significant energy-saving in steam lines by monitoring and acting on leaking steam valves/traps. Our approach leverages an integrated wireless sensor and real-time monitoring capabilities. We make assessments on the real-time status of the distribution system by mounting acoustic sensors on the steam pipes/traps/valves and observe the state measurements of these sensors. Our assessments are based on analysis of the wireless sensor measurements. We describe Fourier-spectrum based algorithms that interpret acoustic vibration sensor data to characterize flows and classify the steam system status. We are able to present the sensor readings, steam flow, steam trap status and the assessed alerts as an interactive overlay within a web-based Google Earth geographic platform that enables decision makers to take remedial action. We believe our demonstration serves as an instantiation of a platform that extends implementation to include newer modalities to manage water flow, sewage and energy consumption.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2017

Flow-induced vibration analysis of a helical coil steam generator experiment using large eddy simulation

This paper describes a numerical study of flow-induced vibration in a helical coil steam generator experiment conducted at Argonne National Laboratory in the 1980s. In the experiment, a half-scale sector model of a steam generator helical coil tube bank was subjected to still and flowing air and water, and the vibrational characteristics were recorded. The research detailed in this document utilizes the multi-physics simulation toolkit SHARP developed at Argonne National Laboratory, in cooperation with Lawrence Livermore National Laboratory, to simulate the experiment. SHARP uses the spectral element code Nek5000 for fluid dynamics analysis and the finite element code DIABLO for structural analysis. The flow around the coil tubes is modeled in Nek5000 by using a large eddy simulation turbulence model. Transient pressure data on the tube surfaces is sampled and transferred to DIABLO for the structural simulation. The structural response is simulated in DIABLO via an implicit time-marching algorithm and a combination of continuum elements and structural shells. Tube vibration data (acceleration and frequency) are sampled and compared with the experimental data. Currently, only one-way coupling is used, which means that pressure loads from the fluid simulation are transferred to the structural simulation but the resulting structural displacements are not fed back to the fluid simulation

CFD↗

DOE OSTI2017

Dual Rate Adaptive Control for an Industrial Heat Supply Process Using Signal Compensation Approach

The industrial heat supply process (HSP) is a highly nonlinear cascaded process which uses a steam valve opening as its control input, the steam flow-rate as its inner loop output and the supply water temperature as its outer loop output. The relationship between the heat exchange rate and the model parameters, such as steam density, entropy, and fouling correction factor and heat exchange efficiency are unknown and nonlinear. Moreover, these model parameters vary in line with steam pressure, ambient temperature and the residuals caused by the quality variations of the circulation water. When the steam pressure and the ambient temperature are of high values and are subjected to frequent external random disturbances, the supply water temperature and the steam flow-rate would interact with each other and fluctuate a lot. This is also true when the process exhibits unknown characteristic variations of the process dynamics caused by the unexpected changes of the heat exchange residuals. As a result, it is difficult to control the supply water temperature and the rates of changes of steam flow-rate well inside their targeted ranges. In this paper, a novel compensation signal based dual rate adaptive controller is developed by representing the unknown variations of dynamics as unmodeled dynamics. In the proposed controller design, such a compensation signal is constructed and added onto the control signal obtained from the linear deterministic model based feedback control design. Such a compensation signal aims at eliminating the unmodeled dynamics and the rate of changes of the currently sample unmodeled dynamics. A successful industrial application is carried out, where it has been shown that both the supply water temperature and the rate of the changes of the steam flow-rate can be controlled well inside their targeted ranges when the process is subjected to unknown variations of its dynamics.

Adaptive signal approacg↗

NASA NTRS2016

In-Situ Measurement of the Height of Condensed Water in Steam Pipes with Dynamic Flow

A method based on the use of enhanced filtered Hilbert envelope of the wave signal was developed in order to monitor the height of condensed water through the wall of steam pipes having dynamic flow conditions. A prototype testbed was designed and fabricated in this study to simulate the dynamic flow conditions including the air stream flowing above the water and bubble induced disturbance. A dual-transducer was used to perform the test as a basis for the multiple transducers system to facilitate the detectability and reliability for long term monitoring of the condensed water height in dynamic conditions. The results demonstrated that the method of measuring the water height using multiple-transducer system employing the developed novel signal processing technique is an efficient and accurate tool for practical applications.

water level in pipe↗

DOE OSTI2016

Measurement of Void Fraction Distribution in Steam-Water Two-Phase Flow in a 44 Bundle at 2 MPa

To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam-water two-phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition (Up to 2.8 MPa, 230 deg. C), focusing on the low flow rate condition after the reactor scram. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift-flux models implemented in the two fluids codes, such as TRACE code. We are still in the middle of the data analyses and here we only report a part of the results. For the present bundle rod geometry, the void fraction shows easily gathered at the center area surrounded by 4 rods and the void fraction in the sub - channels between two adjacent rods are low. Under the same steam vapor flow rate condition, void faction increase with the decrease of the water flow rate. Under the same water flow rate, the void fraction increases with the increase of the steam vapor flow rate. In recent future, we will perform more experiments under further low water flow rate condition, including stagnation condition, to confirm this changing tendency. In the present experimental range, no effect of pressure on void fraction was observed. A 3-layer WMS, with a vertical distance between the adjacent layers of 1.3 mm, was used to measure the local instantaneous void fraction in this research. By using the correlation coefficient method, we can determine the time delay between the bubble detections from the upper and the lower layer sensors. Then the speed of the bubble, and its length, can be calculated. We are working on the establishment of the databank for the slug-churn bubbles information. (authors)

42 ENGINEERING↗

DOE OSTI2012

The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). The tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Characterization of steam impulse turbine for two-phase flow

The characterization of a steam impulse Terry turbine operating with single- and two-phase air–water ingestion has been performed; the air mass fraction ranged from 1.0 (single-phase air) down to 0.05. The two-phase characteristics were obtained over a range of turbine inlet pressures and turbine rotational speeds. Here, the torque produced and the mixture flow rate through the turbine were recorded. Based upon experimental air/water measurements, the turbine Affinity Laws have been modified to include the effects of two-phase flow.

42 ENGINEERING↗

DOE OSTI2019

Simulation of Boiling Two-Phase Flow in a Helical Coil Steam Generator Using the Spectral Element Code Nek-2P

To enable the design of a light water small modular reactor, the boiling flow inside a helical coil steam generator has been simulated with the two-fluid model in Nek-2P. Nek-2P is the multiphase branch of the spectral element code Nek5000. Details of the implementation of the two-fluid model and the included closure models are discussed. The presented closure models include interactions for momentum, heat, and mass transfer between phases. Models for the drag, lift, and turbulent dispersion forces are included. The complete model is fully consistent in the limits of both phasic volume fractions approaching zero and is able to simulate flows of dispersed vapor, continuous liquid, dispersed liquid, continuous vapor, or any combination thereof. The closure models and their implementation in Nek-2P have been validated by comparing to experimental data for a boiling flow, demonstrating excellent agreement. Results from the simulation of the helical coil indicate strong phasic separation driven by the effects of buoyancy and inertia. Finally, significant differences were observed in the results compared to simulations performed using Star-CCM+, although these differences were somewhat expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Once-Through Steam Generator Model Analysis Using Python and Advanced Optimization Tools (Summer Internship Report)

This study focuses on the parametric analysis of design parameters for a once-through steam generator (OTSG) model, using python and advanced optimization tools to facilitate applications such as the flowing autoclave steam generator (FASG) test cases. Building on previous research involving another OTSG with a different design, this project aims to enhance our understanding of how steam generators (SGs) behave and how their outputs are influenced by changes in design. The reason for this design change is to allow for more precise modeling and optimization of SG performance, to provide a comparative analysis between the two designs, and to set up the model for integration with the FASG test case. The OTSG python-model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor-type small modular reactor system. Design studies involve changing the model’s input design parameters to observe the resulting effects on the output of the system. By using advanced optimization tools, such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory, detailed design parametric studies and model optimization were performed. Six input parameters—pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid), respectively, of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (i.e., ±1%, ±5% and ±10% relative changes) for 600 samples. The analysis provides valuable insights into SG optimization and can be used for sensor placement optimization to effectively monitor and obtain experimental data in other tests.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2024

Flow Instabilities in boiling channels and their suppression methodologies—A review

Small modular reactors (SMRs) are gaining significant attention as a promising solution for clean and sustainable nuclear-power generation. However, the operation of SMRs is subject to various challenges, including two-phase flow instabilities. Flow instability has the potential to trigger flow-induced vibration and cyclic fluctuations in local thermal stress. These instabilities frequently manifest because of the complex interplay among a multitude of factors, encompassing thermal-hydraulic conditions, the geometric configuration of the steam generator, and operational parameters. These conditions could subsequently lead to premature critical heat flux, equipment malfunctions, and other safety concerns. Here, the endeavor to address steam-generator flow instabilities is of utmost importance in augmenting the sustainability and efficiency of contemporary energy production. This study offers a comprehensive review of instabilities in two-phase flow, with a particular focus on the influential factors impacting the stability of flow boiling. Furthermore, it delves into the processes of identifying, characterizing, and ameliorating these instabilities, emphasizing pivotal findings, methodologies employed, and avenues for prospective research. The primary parameters of concern encompass the efficient transfer of thermal energy, the optimization of mass-flow rates, and the establishment of favorable boundary conditions, all in the context of steam generator design to alleviate instability for water-cooled SMRs. These discernments bear substantial ramifications for enhancing reactor performance and ensuring operational safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗