SIMULASI KARAKTERISTIK ALIRAN DAN SUHU FLUIDA PENDINGIN (H2O) PADA TERAS REAKTOR NUKLIR SMR (SMALL MODULAR REACTOR)

ROTASI ◽  
2013 ◽  
Vol 15 (4) ◽  
pp. 33
Author(s):  
Anwar Ilmar Ramadhan ◽  
Indra Setiawan ◽  
M. Ivan Satryo
Keyword(s):  
Fluid Flow ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Axial Direction ◽  
Operating Conditions ◽  
Fluid Temperature ◽  
Wide Range ◽  
Fuel Reactor ◽  
Performance Of Heat Transfer

Safety is an issue that is of considerable concern in the design, operation and development of a nuclear reactor. Therefore, the method of analysis used in all these activities should be thorough and reliable so as to predict a wide range of operating conditions of the reactor, both under normal operating conditions and in the event of an accident. Performance of heat transfer to the cooling of nuclear fuel, reactor safety is key. Poor heat removal performance would threaten the integrity of the fuel cladding which could further impact on the release of radioactive substances into the environment in an uncontrolled manner to endanger the safety of the reactor workers, the general public, and the environment. This study has the objective is to know is profile contour of fluid flow and the temperature distribution pattern of the cooling fluid is water (H2O) in convection in to SMR reactor with fuel sub reed arrangement of hexagonal in forced convection. In this study will be conducted simulations on the SMR reactor core used sub channel hexagonal using CFD (Computational Fluid Dynamics) code. And the results of this simulation look more upward (vector of fluid flow) fluid temperature will be warm because the heat moves from the wall to the fluid heater. Axial direction and also looks more fluid away from the heating element temperature will be lower.

Analysis of Passive Residual Heat Removal System at Primary Side when Station Blackout Occurs

2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Author(s):  
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao
Keyword(s):  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Primary Circuit ◽  
Station Blackout ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.

HIGH TEMPERATURE HYBRID NANODIAMOND SENSOR SYSTEMS

2014 ◽  
Vol 2014 (HITEC) ◽  
pp. 000034-000039 ◽  
Author(s):  
John R. Fraley ◽  
Lauren Kegley ◽  
Stephen Minden ◽  
Jimmy L. Davidson ◽  
David Kerns
Keyword(s):  
High Temperature ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Chemical Species ◽  
Chemical Vapor ◽  
Smart Sensors ◽  
Sensor Systems ◽  
High Temperature Electronics ◽  
Wireless Telemetry ◽  
Wide Range

In recent years, high temperature semiconductors have been utilized in wireless telemetry systems for use in military and commercial applications, wherein a high temperature environment combined with other factors such as rotating machinery or weight-constraints preclude the use of conventional silicon based wireless telemetry or wired sensor solutions. Present systems include those which can measure temperatures, pressures, vibrations, and strains. By combining the advanced electronics developed for these systems with novel sensor elements created using chemical vapor deposition (CVD) nanodiamond technology, a wide range of other high temperature sensing systems can be enabled. The unique properties of the diamond sensors have proven in principle the capability to sense, with quantifiable signal, a wide variety of parameters under extreme conditions including very high temperatures and pressures. It has been clear for some time that diamond would be the ideal material of choice for solid-state sensors, but only in recent years has the advent of CVD diamond (as opposed to natural or HPHT [high pressure, high temperature] formation) opened the door for its practical development into harsh environment sensor systems. By combining these diamond sensor elements with high temperature electronics and high temperature packaging approaches, smart sensors can be developed to measure parameters ranging from gas chemical species on the surface of Venus, to neutron flux rates outside of a nuclear reactor core. The research presented here is centered around the use of hybrid diamond sensors for neutron detection applications in Nuclear Thermal Propulsion systems. The current technology state and development needs for these hybrid high temperature diamond smart sensors will be highlighted to potentially encourage future R&D from the high-temperature electronics community.

A Natural Convection Fin with a Solution-Determined Nonmonotonically Varying Heat Transfer Coefficient

1981 ◽  
Vol 103 (2) ◽  
pp. 218-225 ◽  
Author(s):  
E. M. Sparrow ◽  
S. Acharya
Keyword(s):  
Heat Transfer ◽  
Heat Transfer Coefficient ◽  
Natural Convection ◽  
Transfer Coefficient ◽  
First Principles ◽  
Numerical Solutions ◽  
Flow Direction ◽  
Operating Conditions ◽  
Fluid Temperature ◽  
Wide Range

A conjugate conduction-convection analysis has been made for a vertical plate fin which exchanges heat with its fluid environment by natural convection. The analysis is based on a first-principles approach whereby the heat conduction equation for the fin is solved simultaneously with the conservation equations for mass, momentum, and energy in the fluid boundary layer adjacent to the fin. The natural convection heat transfer coefficient is not specified in advance but is one of the results of the numerical solutions. For a wide range of operating conditions, the local heat transfer coefficients were found not to decrease monotonically in the flow direction, as is usual. Rather, the coefficient decreased at first, attained a minimum, and then increased with increasing downstream distance. This behavior was attributed to an enhanced buoyancy resulting from an increase in the wall-to-fluid temperature difference along the streamwise direction. To supplement the first-principles analysis, results were also obtained from a simple adaptation of the conventional fin model.

scholarly journals Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

2020 ◽  
Vol 2020 ◽  
pp. 1-12 ◽  
Author(s):  
Linrong Ye ◽  
Mingjun Wang ◽  
Xin’an Wang ◽  
Jian Deng ◽  
Yan Xiang ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Coupling Method ◽  
High Fidelity ◽  
Monte Carlo Code ◽  
Coupling Analysis ◽  
Fuel Reactor ◽  
Nuclear Reactor Core ◽  
Type Fuel ◽  
Plate Type

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

Predictive Modeling of Acoustic Signals From Thermoacoustic Power Sensors (TAPS)

2016 ◽  
Author(s):  
Christopher M. Dumm ◽  
Jeffrey S. Vipperman ◽  
Jorge V. Carvajal ◽  
Melissa M. Walter ◽  
Luke Czerniak ◽  
...  
Keyword(s):  
Predictive Modeling ◽  
Acoustic Waves ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Signal Propagation ◽  
Acoustic Signals ◽  
Operating Conditions ◽  
Liquid Sodium ◽  
Thermoacoustic Engines ◽  
Self Powered

Thermoacoustic Power Sensor (TAPS) technology offers the potential for self-powered, wireless measurement of nuclear reactor core operating conditions. TAPS are based on thermoacoustic engines, which harness thermal energy from fission reactions to generate acoustic waves by virtue of gas motion through a porous stack of thermally nonconductive material. TAPS can be placed in the core, where they generate acoustic waves whose frequency and amplitude are proportional to the local temperature and radiation flux, respectively. TAPS acoustic signals are not measured directly at the TAPS; rather, they propagate wirelessly from an individual TAPS through the reactor, and ultimately to a low-power receiver network on the vessel’s exterior. In order to rely on TAPS as primary instrumentation, reactor-specific models which account for geometric/acoustic complexities in the signal propagation environment must be used to predict the amplitude and frequency of TAPS signals at receiver locations. The reactor state may then be derived by comparing receiver signals to the reference levels established by predictive modeling. In this paper, we develop and experimentally benchmark a methodology for predictive modeling of the signals generated by a TAPS system, with the intent of subsequently extending these efforts to modeling of TAPS in a liquid sodium environment.

Assessing Advanced Test Reactor Process States Using Acoustic Measurement Infrastructure

2017 ◽  
Author(s):  
Vivek Agarwal ◽  
James A. Smith
Keyword(s):  
Nuclear Reactor ◽  
Signal Transmission ◽  
National Laboratory ◽  
Reactor Core ◽  
Operating Conditions ◽  
Acoustic Measurement ◽  
Primary Coolant ◽  
Test Reactor ◽  
Operation Cycle ◽  
Measurement Infrastructure

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentations. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps, a data-acquisition system, and signal processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the primary coolant pumps are collected and processed. The characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement). This paper demonstrated different in acoustic signature of the ATR under different operating conditions. In particular, ATR acoustic baseline is captured during typical operation cycle and during power axial locator mechanism operation cycle. The difference in two acoustic baseline is significant and highlights salient difference that are critical in the design and development of acoustically telemetered sensors.

Effects of Various Parameters on Supercritical-Pressure Heat Transfer and Limits for Heat Transfer Correlations Obtained in Vertical Bare Tubes

2017 ◽  
Author(s):  
Hakim Maloufi ◽  
Hanqing Xie ◽  
Andrew Zopf ◽  
William Anderson ◽  
Christian Langevin ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Heat Fluxes ◽  
Operating Conditions ◽  
Transfer Coefficients ◽  
Wide Range ◽  
Heat Transfer Correlations ◽  
Bare Tube

Currently, there is a number of Generation-IV SuperCritical Water-cooled nuclear-Reactor (SCWR) concepts under development worldwide. These high temperature and pressure reactors will have significantly higher operating parameters compared to those of current water-cooled nuclear-power reactors (i.e., “steam” pressures of about 25 MPa and “steam” outlet temperatures up to 625 °C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated, as the steam will be flowing directly to a steam turbine. In support of developing SCWRs studies are being conducted on heat transfer at SuperCritical Pressures (SCPs). Currently, there are very few experimental datasets for heat transfer at SCPs in power-reactor fuel bundles to a coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations developed with bare-tube data can be used as a conservative approach. Selected empirical heat-transfer correlations, based on experimentally obtained datasets, have been put forward to calculate Heat Transfer Coefficients (HTCs) in forced convective in various fluids, including water at SCPs. The Mokry et al. correlation (2011) has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in Normal and Improved Heat-Transfer (NHT and IHT) regimes. However, it is known that a Deteriorated Heat-Transfer (DHT) regime appears in bare tubes earlier than that in bundle flow geometries. Therefore, it is important to know if bare-tube heat-transfer correlations for SCW can predict HTCs at heat fluxes beyond those defined as starting of DHT regime in bare tubes. The Mokry et al. (2011) correlation fits the best SCW experimental data for HTCs and inner wall temperature for bare tubes at SCPs within the NHT and IHT regimes. However, this correlation might have problems with convergence of iterations at heat fluxes above 1000 kW/m2.

scholarly journals CALCULATION OF RADIOACTIVE SOURCE TERM RELEASE FROM FLEXBLUE NPP

2021 ◽  
Vol 23 (2) ◽  
pp. 63
Author(s):  
Muhammad Budi Setiawan ◽  
Pande Made Udiyani
Keyword(s):  
Nuclear Power ◽  
Noble Metals ◽  
Nuclear Reactor ◽  
Source Term ◽  
Reactor Core ◽  
Fission Products ◽  
Operating Conditions ◽  
Normal Operating ◽  
Small Modular Reactor ◽  
Mechanistic Approach

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).

Validation of a Novel Model for Shell and Tube Heat Exchangers Undergoing Crude Oil Fouling

2010 ◽  
Author(s):  
Francesco Coletti ◽  
Sandro Macchietto
Keyword(s):  
Heat Exchanger ◽  
Crude Oil ◽  
Operating Conditions ◽  
Process Conditions ◽  
Model Parameters ◽  
Fluid Temperature ◽  
Oil Companies ◽  
Wide Range ◽  
Shell And Tube ◽  
Oil Fouling

Fouling in refinery heat transfer units is a major problem that affects plant’s economics, operability, safety and environmental impact. Traditional heat exchanger design methodologies based on fixed values for the fouling resistance (e.g. TEMA fouling factors) have drawn several critiques in the past 40 years and were found responsible for exacerbating fouling rather than mitigating it. The fouling factors approach is, in fact, highly empirical and neglects fouling dynamics and its dependency on process conditions. The ability of capturing such dependency is therefore pivotal to overcome traditional design limitations. A novel dynamic, distributed model for a multi–pass shell–and–tube heat exchanger undergoing crude oil fouling was recently proposed by Coletti and Macchietto. The model takes into account the exchanger geometry and configuration, the variation of fluid temperature, velocity, physical properties and fouling rate along the length of each unit and captures the interactions between the fouling layer growth and the fluid–dynamics by solving a moving boundary problem. In this paper, the model is validated over a wide range of operating conditions (i.e. temperatures and flowrates) with data from four different industrial units (2 single and 2 double shells). Geometries and process conditions used are those of two refineries belonging to major oil companies (ExxonMobil and Shell). Some model parameters are estimated for each exchanger using measurements during the first 60 days after a mechanical cleaning. The model is then used in a fully predictive mode for subsequent times. Results indicate that for all units the outlet temperatures (in °C) are predicted over extended periods (i.e. 4–16 months) with an excellent accuracy of ±1% for the tube-side and ±2% for the shell-side. It is concluded that the model can be used with confidence on a wide range of operating conditions to calculate reliable temperatures and fouling resistances.

scholarly journals Patchy nuclear chain reactions

2021 ◽  
Vol 4 (1) ◽  
Author(s):  
Eric Dumonteil ◽  
Rian Bahran ◽  
Theresa Cutler ◽  
Benjamin Dechenaux ◽  
Travis Grove ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Rensselaer Polytechnic Institute ◽  
Operating Conditions ◽  
Chain Reactions ◽  
Branching Random Walks ◽  
Stochastic Fluctuations ◽  
Nuclear Chain ◽  
Reactor Operating ◽  
Experimental Findings

AbstractStochastic fluctuations of the neutron population within a nuclear reactor are typically prevented by operating the core at a sufficient power, since a deterministic (i.e., exactly predictable) behavior of the neutron population is required by automatic safety systems to detect unwanted power excursions. In order to characterize the reactor operating conditions at which the fluctuations vanish, an experiment was designed and took place in 2017 at the Rensselaer Polytechnic Institute Reactor Critical Facility. This experiment however revealed persisting fluctuations and striking patchy spatial patterns in neutron spatial distributions. Here we report these experimental findings, interpret them by a stochastic modeling based on branching random walks, and extend them using a “numerical twin” of the reactor core. Consequences on nuclear safety will be discussed.

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