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

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.

Experimental Investigations of Control Rod Drive Mechanism

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.

Study on the Rod Drop Performance of High-Temperature Gas-Cooled Reactor

2014 ◽  
Author(s):  
He Yan ◽  
Xingzhong Diao
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Insertion Time ◽  
Operational Mode ◽  
Control Rod ◽  
Rod System ◽  
Key Factor ◽  
Control Rods ◽  
High Temperature Gas

In this paper, the theoretical study and experimental investigation on the rod drop performance of high-temperature gas-cooled reactor (HTGR) pebble-bed module have been presented. The control rod drive mechanisms (CRDMs), serving as the first shutdown system of the reactor, are positioned above the reactor pressure vessel. When the reactor is operated at the power regulation mode, the control rods are pulled up-and-down in their channels around the reactor core. The CRDM provides a fail-safe operational mode for the control rod system. If the reactor emergency shutdown is required the control rods could drop into their channels by gravity. Thus the key factor, emergency insertion time of the whole control rod stroke, which represents the inherent safety of the CRDM, is crucially important and should be measured precisely. In the final objective of ensuring reliability of the CRDM, a full size drive line had been built and tested to obtain the overall performance function of the CRDM. Every component of the CRDM test line was simulated at the scale 1:1, including a 15 meters high test bench that was used as the substitution of the pressure vessel. At current stage, the rod drop performance had been experimental investigated at ambient temperature and pressure. The emergency insertion time of an 8 meters stroke was measured to be less than 50 seconds. A mathematical model of CRDM also had been developed. The rod motion characteristic equations show that the rod dropping speed approaches to a constant during the emergency insertion. The theoretical results are in agreement with the test results.

scholarly journals Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽  
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

<p class="NoSpacing1"><span lang="IN">The Rod Control System is </span>employed<span lang="IN"> to adjust the position of the control rods in the reactor core </span>which corresponds with <span lang="IN">the thermal power generated in the core </span>as well as <span lang="IN">the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive </span>employs <span lang="IN">magnetic stepping-type mechanism. This </span>type of <span lang="IN">mechanism consists of a pair of circular coils and latch-style jack with the armature. When the </span>electric <span lang="IN">current </span>is <span lang="IN">supplied to the coils sequentially, the control-rods</span>, which <span lang="IN">are held on the drive shaft</span>, can be driven<span lang="IN"> up</span>ward<span lang="IN"> or down</span>ward<span lang="IN"> in increments. </span>This <span lang="IN">sequential current </span>c<span lang="IN">ontrol</span> drive<span lang="IN"> system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or </span>known also as <span lang="IN">the Rod Control System (RCS). The p</span>urpose of this paper is to investigate the RCS reliability <span lang="IN">of APWR </span>using <span lang="IN">the Fault Tree Analysis (FTA)</span> method<span lang="IN"> since </span>the analysis of reliability which considers<span lang="IN"> the FTA</span> for common CRDM <span lang="IN">can </span>not <span lang="IN">be found</span> in <span lang="IN">any </span>public references. <span lang="IN">The FTA method is used to model the system reliability by developing the fault tree diagram of the system. </span>The<span lang="IN"> 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.</span></p>

scholarly journals Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Energies ◽  
2021 ◽  
Vol 14 (21) ◽  
pp. 7377
Author(s):  
Michał Górkiewicz ◽  
Jerzy Cetnar
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Reactor Performance ◽  
Fission Rate ◽  
Control Rod ◽  
Structured Design ◽  
The Core ◽  
Stable Core ◽  
Control Rods ◽  
Burnup Calculation

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

scholarly journals The effects of fuel type on control rod reactivity of pebble-bed reactor

Nukleonika ◽  
2019 ◽  
Vol 64 (4) ◽  
pp. 131-138
Author(s):  
◽  
Topan Setiadipura ◽  
Jim C. Kuijper ◽  
Keyword(s):  
Reactor Core ◽  
Fuel Type ◽  
Reactor Model ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Reactor Geometry ◽  
Calculation Results ◽  
Control Rods ◽  
Selection Of

Abstract As a crucial core physics parameter, the control rod reactivity has to be predicted for the control and safety of the reactor. This paper studies the control rod reactivity calculation of the pebble-bed reactor with three scenarios of UO2, (Th,U)O2, and PuO2 fuel type without any modifications in the configuration of the reactor core. The reactor geometry of HTR-10 was selected for the reactor model. The entire calculation of control rod reactivity was done using the MCNP6 code with ENDF/B-VII library. The calculation results show that the total reactivity worth of control rods in UO2-, (U,Th)O2-, and PuO2-fueled cores is 15.87, 15.25, and 14.33%Δk/k, respectively. These results prove that the effectiveness of total control rod in thorium and uranium cores is almost similar to but higher than that in plutonium cores. The highest reactivity worth of individual control rod in uranium, thorium and plutonium cores is 1.64, 1.44, and 1.53%Δk/k corresponding to CR8, CR1, and CR5, respectively. The other results demonstrate that the reactor can be safely shutdown with the control rods combination of CR3+CR5+CR8+CR10, CR2+CR3+CR7+CR8, and CR1+CR3+CR6+CR8 in UO2-, (U,Th)O2-, and PuO2-fueled cores, respectively. It can be concluded that, even though the calculation results are not so much different, however, the selection of control rods should be considered in the pebble-bed core design with different scenarios of fuel type.

scholarly journals Measurement and Comparison of Control Rod Worth of BTRR using Inhour Equation and Period reactivity conversion table

2017 ◽  
Vol 41 (1) ◽  
pp. 95-103
Author(s):  
Md Iqbal Hosan ◽  
MAM Soner ◽  
Md Fazlul Huq ◽  
Khorshed Ahmad Kabir
Keyword(s):  
Doubling Time ◽  
Nuclear Reactor ◽  
Reactor Control ◽  
Control Rod ◽  
Fuel Burn ◽  
Start Up ◽  
Conversion Table ◽  
Power Operation ◽  
Academy Of Sciences ◽  
Control Rods

In a nuclear reactor, control rod is a very essential part and plays the elementary role in the reactor control during reactor start up, normal power operation, experimental research and shutdown. To perform all these operations safely, knowledge of differential and integral worth of the control rod is mandatory. In this study, the differential and integral worth curve of all control rods of BAEC TRIGA Research Reactor (BTRR) have been determined by using the positive period method. Reactor period was measured from 1.5 folding time, doubling time, 5 folding time respectively; and in the above three cases reactivity has also been calculated from INHOUR equation and period reactivity conversion table. The total worth of all control rods of BTRR are measured as 14.888 $, 14.672 $, 14.348 $ from INHOUR equation and 13.978 $, 13.672 $, 13.357 $ from period reactivity conversion table for 1.5 folding time, doubling time and 5 folding time respectively. The measured reactivity has also been compared with the previously measured reactivity and due to fuel burn up of the reactor expected lower values were observed.Journal of Bangladesh Academy of Sciences, Vol. 41, No. 1, 95-103, 2017

Developments of the Hydraulic Driving System of the Control Rods for Nuclear Reactors

1999 ◽  
Author(s):  
Yuanqiang Wu
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Reactor Core ◽  
Hydraulic Pressure ◽  
Energy Agency ◽  
Control Rod ◽  
Driving System ◽  
History Of ◽  
The Cost ◽  
Control Rods

Abstract The developments of a new hydraulic driving system of the control rods for nuclear reactors are introduced in this paper. Compared with other driving systems of the control rods, this new hydraulic driving system can be set within the reactor pressure vessel. Under any serious condition, the control rods will not be ejected from the reactor core. Its structure is very simple and the mechanic chain is very short, and thus it is very reliable. It can reduce the height of the nuclear reactor by one-third, and thus dramatically reduce the cost of the reactor. It uses the dynamic hydraulic pressure to control the motion of the control rods. Under extreme conditions, such as the failure of control power supply, the control rods will drop into the reactor core because of their self-weight to shut down the nuclear reaction. Because of these features, International Atomic Energy Agency (IAEA) is very interested in this safe and economical new control rod driving system. A brief history of the developments of the hydraulic driving system is given. Three configurations, the orifice hydraulic step cylinder, the groove-orifice hydraulic step cylinder, and the piston-groove hydraulic step cylinder, are introduced and their working principles are explained. The reliability and safety of the new system are validated by two experimental works: hydraulic step cylinder (HSC) under seismic and rocking conditions. Results from these experiments are presented.

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

2020 ◽  
Vol 225 ◽  
pp. 04024
Author(s):  
Igor Lengar ◽  
Sebastjan Rupnik ◽  
Andrej Žohar ◽  
Vid Merljak ◽  
Marjan Kromar ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Nuclear Power ◽  
Physical Parameters ◽  
Ionization Chambers ◽  
Triga Reactor ◽  
Control Rod ◽  
Multiple Fission ◽  
Insertion Method ◽  
Flux Measurements ◽  
Control Rods

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.

Sign in / Sign up

Export Citation Format

Share Document