A Set of Fuzzy Systems to Automate the Manual Procedures for Reactor Power Level Changes

2000 ◽  
pp. 40-51
Author(s):  
Byung Soo Moon
Keyword(s):  
Fuzzy Systems ◽  
Power Level ◽  
Reactor Power

scholarly journals CALCULATION OF CONTROL AND ACCEPTABLE LEVELS OF RADIATION FACTORS FOR NUCLEAR PROPULSIONS

2019 ◽  
Vol 5 (2) ◽  
pp. 76-82
Author(s):  
B. I. Zhabrunov ◽  
A. A. Kern ◽  
A. S. Tazov ◽  
B. V. Kutashov
Keyword(s):  
Lower Limit ◽  
Radiation Safety ◽  
Power Level ◽  
Test Procedure ◽  
Mean Value ◽  
Measured Data ◽  
Reactor Power ◽  
Radiation Factor ◽  
Measuring Range ◽  
Acceptable Error

In accordance with requirements of regulatory and guideline documents on radiation safety for controlled radiation factors for the purposes of operating control, controlled and acceptable levels are established. Any excess of these levels requires the determination of the causes and implementation of actions designed to eliminate the excess. The paper presents the method of calculation of these levels and establishing the levels in practice at the present time, disadvantages of accepted regulations are analyzed. It was shown that existing documents do not take into account some circumstances that define the radiation safety test procedure. In a number of measured control points of the radiological situation and staff radiation exposure, the values of controlled parameters are independent of reactor system mode. In the same points that show the dependence of measured data on a reactor power level, values of controlled parameters may also depend on a mode of pumps and purification system. Furthermore, real-time measurements review has showed that beyond the range of lower limit of measuring range of verification means in the range with nonspecified error, the measured data variance is described by mean value and acceptable error. At the same time, a mean value may be a lower order to lower limit of measuring range. Setting a value of controlled level equal to a sum of a mean of double or tripled root-mean-square deviation depending on the accepted confidence level, a possibility of earlier detection of controlled level excess emerges. In this situation, an exact absolute value of a controlling parameter is not essential as that radiation factor level poses no hazard to life. It is important to capture the onset of significant increase of radiation factor i.e. change of radiological situation.

scholarly journals The Local Seismoacoustic Wavefield of a Research Nuclear Reactor and Its Response to Reactor Power Level

2020 ◽  
Vol 92 (1) ◽  
pp. 378-387
Author(s):  
Omar E. Marcillo ◽  
Monica Maceira ◽  
Chengping Chai ◽  
Christine Gammans ◽  
Riley Hunley ◽  
...  
Keyword(s):  
Energy Transport ◽  
Nuclear Reactor ◽  
Power Level ◽  
National Laboratory ◽  
Reactor Power ◽  
Operational Conditions ◽  
Oak Ridge ◽  
Research Nuclear Reactor ◽  
Full Power ◽  
Power Operation

Abstract We describe the seismoacoustic wavefield recorded outdoors but inside the facility fence of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (Tennessee). HFIR is a research nuclear reactor that generates neutrons for scattering, irradiation research, and isotope production. This reactor operates at a nominal power of 85 MW, with a full-power period between 24 and 26 days. This study uses data from a single seismoacoustic station that operated for 60 days and sampled a full operating reactor cycle, that is, full-power operation and end-of-cycle outage. The analysis presented here is based on identifying signals that characterize the steady, that is, full-power operation and end-of-cycle outage, and transitional, that is, start-up and shutdown, states of the reactor. We found that the overall seismoacoustic energy closely follows the main power cycle of the reactor and identified spectral regions excited by specific reactor operational conditions. In particular, we identified a tonal noise sequence with a fundamental frequency around 21.4 Hz and multiple harmonics that emerge as the reactor reaches 90% of nominal power in both seismic and acoustic channels. We also utilized temperature measurements from the monitoring system of the reactor to suggest links between the operation of reactor’s subsystems and seismoacoustic signals. We demonstrate that seismoacoustic monitoring of an industrial facility can identify and track some industrial processes and detect events related to operations that involve energy transport.

ENHS reactor power level enhancement possibilities

2008 ◽  
Vol 50 (2-6) ◽  
pp. 279-285
Author(s):  
Arnaud Susplugas ◽  
Ehud Greenspan
Keyword(s):  
Power Level ◽  
Reactor Power

scholarly journals Feasibility study of the university of Utah TRIGA reactor power upgrade - part II: Thermohydraulics and heat transfer study in respect to cooling system requirements and design

2013 ◽  
Vol 28 (4) ◽  
pp. 352-361
Author(s):  
Philip Babitz ◽  
Dongok Choe ◽  
Tatjana Jevremovic
Keyword(s):  
Heat Transfer ◽  
Cooling System ◽  
Power Level ◽  
Flow Simulation ◽  
Nucleate Boiling ◽  
Reactor Power ◽  
Convection Current ◽  
Triga Reactor ◽  
Higher Power ◽  
The University

The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much more pronounced at higher power levels with vortex shedding also occurring. A departure from nucleate boiling analysis showed that while nucleate boiling begins near 210 kW it remains in this state and does not approach the critical heat flux at powers up to 500 kW. Based on these studies, two upgrades are proposed for extended operation and possibly higher reactor power level. Together with the findings from Part I studies, we conclude that increase of the reactor power is highly feasible yet dependable on its purpose and associated investments.

Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on the HTR-10

2008 ◽  
Author(s):  
Yujie Dong ◽  
Fubing Chen ◽  
Zuoyi Zhang ◽  
Shouyin Hu ◽  
Lei Shi ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Power Level ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Code System ◽  
Safety Features ◽  
Forced Cooling ◽  
High Temperature Gas

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

Conceptual design and analysis of a megawatt power level heat pipe cooled space reactor power system

2020 ◽  
Vol 144 ◽  
pp. 107576
Author(s):  
Wenwen Zhang ◽  
Dalin Zhang ◽  
Chenglong Wang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
...  
Keyword(s):  
Power System ◽  
Heat Pipe ◽  
Conceptual Design ◽  
Power Level ◽  
Reactor Power ◽  
Space Reactor
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