Usage of multiple fission cells for neutron flux measurements during rod-insertion method

The measurements of physical parameters of the TRIGA reactor and Nuclear power plant Krško (NEK) reactor cores have been in the past performed on hand of the neutron flux signal obtained from uncompensated ionization cells and by employment of the a digital meter of reactivity (DMR). At the TRIGA reactor only one ionization cell is currently used for flux measurements. During the insertion of one control rod the neutron flux distribution is significantly altered affecting the flux measurements of inserting different control rods. The problem is presently solved by assigning a correction factor to each control rod what introduces an additional uncertainty. In the present paper the implementation of four fission cells for reactivity measurements is presented. In this way determining the correct gamma background and its subtraction, performed by DMR algorithms, becomes less important as previously by using ionization chambers. The larger number of detectors also reduces the flux redistribution effects on the signal during individual control rod movements.

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Experimental Investigations of Control Rod Drive Mechanism

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-15196 ◽

2013 ◽

Author(s):

Guangyao Lu ◽

Zhaohui Lu ◽

Wenyuan Xiang ◽

Yonghong Lv ◽

Wenyou Huang ◽

...

Keyword(s):

Nuclear Power ◽

Nuclear Reactor ◽

Reactor Core ◽

Experimental Investigations ◽

Control Rod ◽

Drive Mechanism ◽

Start Up ◽

Power Regulation ◽

Set Up ◽

Control Rods

The control rod drive mechanism (CRDM) is installed on the CRDM socket in reactor pressure vessel (RPV). Directed by Rod Control and Rod Position Indicating System (RGL), CRDM can impel the control rods move up and down in the nuclear reactor core, which implements the functions of reactor start-up, power regulation, power maintaining, normal reactor shutdown and abnormal (accident) shutdown. CRDM was developed by China Nuclear Power Research Institute (CNPRI). Several design improvements were conducted to solve the problems appeared in the operation of nuclear power station. Test bench was also set up and cold tests were carried out to investigate the characteristics of CRDM. The cold tests included lifting experiment, inserting experiment, rod drop experiment. And studies were carried out to analyze the signals of lifting coil, moving coil, stationary coil and the vibration signals. The test results show that the design of CRDM is reasonable and the operation is reliable.

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Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽

10.18502/ken.v1i1.465 ◽

2016 ◽

Vol 1 (1) ◽

Author(s):

Syaiful Bakhri

Keyword(s):

Control System ◽

System Reliability ◽

Nuclear Power ◽

Power Plants ◽

Thermal Power ◽

Fault Tree ◽

Reactor Core ◽

Pressurized Water ◽

Control Rod ◽

Control Rods

The Rod Control System is employed to adjust the position of the control rods in the reactor core which corresponds with the thermal power generated in the core as well as the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive employs magnetic stepping-type mechanism. This type of mechanism consists of a pair of circular coils and latch-style jack with the armature. When the electric current is supplied to the coils sequentially, the control-rods, which are held on the drive shaft, can be driven upward or downward in increments. This sequential current control drive system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or known also as the Rod Control System (RCS). The purpose of this paper is to investigate the RCS reliability of APWR using the Fault Tree Analysis (FTA) method since the analysis of reliability which considers the FTA for common CRDM can not be found in any public references. The FTA method is used to model the system reliability by developing the fault tree diagram of the system. The results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.

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Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

Applied Radiation and Isotopes ◽

10.1016/j.apradiso.2011.05.019 ◽

2011 ◽

Vol 69 (11) ◽

pp. 1621-1624 ◽

Cited By ~ 4

Author(s):

Stefan Merz ◽

Mile Djuricic ◽

Mario Villa ◽

Helmuth Böck ◽

Georg Steinhauser

Keyword(s):

Neutron Flux ◽

Triga Reactor ◽

Flux Measurements

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Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements

Applied Radiation and Isotopes ◽

10.1016/j.apradiso.2015.06.026 ◽

2015 ◽

Vol 104 ◽

pp. 34-42 ◽

Cited By ~ 14

Author(s):

Tanja Kaiba ◽

Gašper Žerovnik ◽

Anže Jazbec ◽

Žiga Štancar ◽

Loïc Barbot ◽

...

Keyword(s):

Neutron Flux ◽

Triga Reactor ◽

Control Rod ◽

Flux Redistribution

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Calculation on the Distribution of Ex-Core Neutron Flux during Reactor Start-Up

Applied Mechanics and Materials ◽

10.4028/www.scientific.net/amm.672-674.375 ◽

2014 ◽

Vol 672-674 ◽

pp. 375-378

Author(s):

Chun Yu Liu ◽

Benbicha Mohamed Elghazali

Keyword(s):

Neutron Flux ◽

Physical Model ◽

Pressure Vessel ◽

Reactor Core ◽

Mcnp Code ◽

Control Rod ◽

Start Up ◽

Control Rods ◽

Moving Process ◽

Detection Effect

The distribution of neutron flux is simulated by MCNP code from reactor start-up to criticality when the control rods are drawn different length from reactor core on ex-core detector area of the WWER reactor of TianWan. Because physical model built is very large, in order to save calculation time, the moving process of control rod is simplified. The results of calculation show that The neutron mainly distributed in the range of 0-400cm outside the pressure vessel. The value of the relative neutron flux ex-core is maximum in the range of 110cm to 180cm, so detection effect is better when the detector is set in this region.

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The reactor research and test facility

Space engineering and technology ◽

10.33950/spacetech-2308-7625-2020-4-69-79 ◽

2020 ◽

pp. 69-79

Author(s):

Andrey S. KIRILLOV ◽

Aleksandr P. PYSHKO ◽

Andrey A. ROMANENKO ◽

Valery I. YARYGIN

Keyword(s):

Measurement System ◽

Nuclear Power ◽

Power Plants ◽

Test Facility ◽

Physical Parameters ◽

Pumping System ◽

Current State ◽

History Of ◽

Special Measurement ◽

Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

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