scholarly journals Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

2018 ◽  
Vol 170 ◽  
pp. 04018
Author(s):  
Michael A. Reichenberger ◽  
Daniel M. Nichols ◽  
Sarah R. Stevenson ◽  
Tanner M. Swope ◽  
Caden W. Hilger ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Nuclear Reactor ◽  
Modular Design ◽  
Reactor Core ◽  
Operating Conditions ◽  
Sensor Nodes ◽  
Detector Array ◽  
State University ◽  
The Core ◽  
Kansas State University

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

scholarly journals Gas-cooled nuclear reactor core shaping using heat exchange intensifiers

2019 ◽  
Vol 5 (1) ◽  
pp. 75-80
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Heat Exchange ◽  
Pressure Drop ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Temperature Fields ◽  
Ansys Fluent ◽  
Safety Requirements ◽  
Cooling Channels ◽  
The Core ◽  
Nuclear Reactor Core

The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping. The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed. The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields. The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels. Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop. The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).

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.

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.

Nodal Expansion and its Application in AP1000 Nuclear Reactor Core Monitoring System

2013 ◽  
Vol 448-453 ◽  
pp. 1907-1911
Author(s):  
Wei Zhi Jia ◽  
Rui Wang ◽  
Yun Zhou
Keyword(s):  
Monitoring System ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
The Core ◽  
Nodal Model ◽  
Nuclear Reactor Core ◽  
Nodal Expansion Method ◽  
Full Core

As the core monitoring system of AP1000, BEACON always uses a full-core nodal model for core monitoring based on the ANC-NEM nodal model. The theory behind the nodal expansion method is discussed, and the application of the method in BEACON is described. Finally, an ANC-NEM calculation simulation is proposed.

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.

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).

scholarly journals Effects of various types of reflectors on the 99Mo production in the VVER-Ts reactor targets

2020 ◽  
Vol 6 (2) ◽  
pp. 89-92
Author(s):  
Oleg Yu. Kochnov ◽  
Pavel A. Danilov
Keyword(s):  
Neutron Flux ◽  
Reactor Core ◽  
Raw Material ◽  
The Core ◽  
Assessment Tasks ◽  
Materials Used ◽  
Core Characteristics ◽  
Water Space ◽  
Reactivity Margin ◽  
Effective Multiplication

The effects from introducing various types of reflectors in the VVR-Ts reactor core on the 99Мо production were analyzed. Earlier the effects of only the beryllium reflector on the VVR-Ts reactor core characteristics, such as reactivity margin, neutron flux in experimental channels, and activity of the accumulated 99Мо, were calculated. The calculations are based on a generated precision model of the core which comprises one experimental channel where targets are irradiated for the 99Мо production. The model was built using the SCALE code. The code allows a fairly broad range of calculations to be performed, from criticality estimation to radiological assessment tasks. As the result of the computational analysis of the model, such characteristics were obtained as the effective multiplication factor, the power density in the 99Мо production targets, the neutron flux in the target raw material, and the quantity of the produced 99Мо after 120 hours of irradiation. The data was compared with the results of similar calculations of the VVR-Ts reactor core parameters. Further, the list of the materials used extensively as the reactor core reflector or moderator was formed based on reference literature. A number of models were obtained and analyzed on its basis, in which the water space on the core periphery was substituted for the investigated materials.

scholarly journals Condition monitoring of an advanced gas-cooled nuclear reactor core

2007 ◽  
Vol 221 (6) ◽  
pp. 833-843 ◽  
Author(s):  
Y Pang ◽  
L Giovanini ◽  
M Monari ◽  
M Grimble
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Fault Detection And Isolation ◽  
Unknown Input ◽  
Unknown Input Observer ◽  
Fuel Channel ◽  
The Core ◽  
Nuclear Reactor Core ◽  
Monitoring Module

A critical component of an advanced gas-cooled reactor station is the graphite core. As a station ages, the graphite bricks that comprise the core can distort and may eventually crack. Since the core cannot be replaced, the core integrity ultimately determines the station life. Monitoring these distortions is usually restricted to the routine outages, which occur every few years, as this is the only time that the reactor core can be accessed by external sensing equipment. This paper presents a monitoring module based on model-based techniques using measurements obtained during the refuelling process. A fault detection and isolation filter based on unknown input observer techniques is developed. The role of this filter is to estimate the friction force produced by the interaction between the wall of the fuel channel and the fuel assembly supporting brushes. This allows an estimate to be made of the shape of the graphite bricks that comprise the core and, therefore, to monitor any distortion on them.

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