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.

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.

Preliminary Evaluation of the Fixed Bed Nuclear Reactor Concept Using IAEA-INPRO Methodology

2004 ◽  
Author(s):  
Farhang Sefidvash
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Reactor Core ◽  
Primary Circuit ◽  
Preliminary Evaluation ◽  
Spent Fuel ◽  
Energy Agency ◽  
Reactor Technology ◽  
Fuel Elements

The Atomic Energy Agency (IAEA) through its INPRO Project has developed a methodology to evaluate the innovative nuclear reactors. The main objectives of INPRO are to help to ensure that nuclear energy is available in the 21st century in a sustainable manner; and bring together both technology holders and technology users to achieve desired innovations in nuclear reactors and nuclear fuel cycles which are to be acceptable to the public because they are economic, safe, proliferation resistant, sustainable, and having reduced environmental impact. Here is a preliminary application of this methodology (IAEA-TECDOC-1362) to evaluate the Fixed Bed Nuclear Reactor Concept (FBNR). Some of the characteristics of the proposed reactor are: The FBNR is based on pressurized light water reactor technology. It is a small, modular, and integrated primary circuit reactor. The fuel elements of FBNR are 8 mm diameter spherical uranium dioxide pellets cladded by zircaloy or made of compacted TRISO type fuel particles. The reactor core is suspended by the flow of water coolant. The stop in flow causes the fuel elements leave the reactor core by the force of gravity and fall into a passively cooled fuel chamber or even leave the reactor completely and become deposited in the spent fuel pool. It is an inherently safe and passively cooled reactor concept. FBNR in its advanced versions can use supercritical steam or helium gas as coolant, and utilize MOX or thorium fuel.

Comparative Studies of Turbulence Models Application in Thermal Hydraulic Analysis of Nuclear Reactor Core

2010 ◽  
Author(s):  
Heyi Zeng ◽  
Yun Guo
Keyword(s):  
Fluid Dynamics ◽  
Turbulent Flow ◽  
Secondary Flow ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Reactor Core ◽  
Turbulence Models ◽  
Flow Phenomena ◽  
Nuclear Reactor Core ◽  
Complicated Geometry

Rod bundles are essential elements of pressurized water nuclear reactors. They consist of tightly packed arrays of rods, which contain the nuclear fuel and are surrounded by flowing liquid coolant. Flow phenomena in the subchannels bounded by adjacent rods are quite complex and exhibit patterns not present in pipe flows. Development of nuclear reactors and of fuel assemblies requires fluid dynamics analysis activities. The detailed prediction of velocity and temperature distributions inside a rod bundle is one of the main objectives of the current research in reactor thermal hydraulics. Computational fluid dynamics (CFD) simulation is of great interest for the design and safety analysis of nuclear reactors since it has recently achieved considerable advancements. In the present studies, numerical simulation were performed on developed turbulent flow through core subchannels with configurations of triangle and square lattice, and impact of different turbulence models built-in software package FLUENT upon simulation results of velocity distribution and hydraulic characteristics in channels with complicated geometry were compared and analyzed. Results show that simulation result greatly depends on turbulence models. Due to the complicated geometric construction, the complicated three-dimensional turbulent flow shows highly anisotropic characteristics. Turbulence models assuming isotropic turbulent viscosity failed to predict secondary flow phenomena during turbulent flow in fuel assembly channel. By solving Reynolds stresses transport equations, more elaborate Reynolds stress model (RSM) can catch secondary flow accurately. The present studies have provided valuable references and guidelines for further investigation on convective heat transfer simulation in complicated geometry and thermalhydrulic analysis of nuclear reactor core.

Release of 108Ag from Irradiated PWR Control Rod Absorbers under Deep Repository Conditions

2015 ◽  
Vol 1744 ◽  
pp. 217-222
Author(s):  
O. Roth ◽  
M. Granfors ◽  
A. Puranen ◽  
K. Spahiu
Keyword(s):  
Nuclear Fuel ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Nuclear Reactors ◽  
Reactor Irradiation ◽  
Spent Fuel ◽  
Control Rod ◽  
Potential Source ◽  
Absorber Material ◽  
Control Rods

ABSTRACTIn a future Swedish deep repository for spent nuclear fuel, irradiated control rods from PWR nuclear reactors are planned to be stored together with the spent fuel. The control rod absorber consists of an 80% Ag, 5% Cd, 15% In alloy with a steel cladding. Upon in-reactor irradiation 108Ag is produced by neutron capture. Release of 108Ag has been identified as a potential source term for release of radioactive substances from the deep repository.Under reducing deep repository conditions, the Ag corrosion rate is however expected to be low which would imply that the release rate of 108Ag should be low under these conditions. The aim of this study is to investigate the dissolution of PWR control rod absorber material under conditions relevant to a future deep repository for spent nuclear fuel. The experiments include tests using irradiated control rod absorber material from Ringhals 2, Sweden. Furthermore, un-irradiated control rod absorber alloy has been tested for comparison. The experiments indicate that the release of Ag from the alloy when exposed to water is strongly dependent on the redox conditions. Under aerated conditions Ag is released at a significant rate whereas no release could be measured after 133 days during leaching under H2.

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.

A Review on Specific Features of Small and Medium Sized Nuclear Power Plants

2010 ◽  
Author(s):  
Salah Ud-din Khan ◽  
Minjun Peng ◽  
Muhammad Zubair ◽  
Shaowu Wang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Nuclear Industry ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Energy Supply System

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.

scholarly journals Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

2017 ◽  
Vol 2017 ◽  
pp. 1-8 ◽  
Author(s):  
Tagor Malem Sembiring ◽  
Surian Pinem ◽  
Peng Hong Liem
Keyword(s):  
Nuclear Energy ◽  
Neutron Diffusion ◽  
Energy Agency ◽  
Control Rod ◽  
3 Dimensional ◽  
Nodal Method ◽  
Calculation Results ◽  
Control Rods ◽  
Neutron Diffusion Equation ◽  
Good Agreement

The in-house coupled neutronic and thermal-hydraulic (N/T-H) code of BATAN (National Nuclear Energy Agency of Indonesia), NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers), respectively.

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