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 A simulation of a pebble bed reactor core by the MCNP-4C computer code

2009 ◽  
Vol 24 (3) ◽  
pp. 177-182 ◽  
Author(s):  
Moshkbar Bakhshayesh ◽  
Naser Vosoughi
Keyword(s):  
Computing Time ◽  
Reactor Core ◽  
Computer Code ◽  
Critical Height ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Pebble Bed Reactors ◽  
New Generation ◽  
And Control ◽  
Control Rods

Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results), chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

Combined CFD and DEM Simulations of Fluid Flow and Heat Transfer in a Pebble Bed Reactor

2011 ◽  
Author(s):  
Xiang Zhao ◽  
Trent Montgomery ◽  
Sijun Zhang
Keyword(s):  
Heat Transfer ◽  
Fluid Flow ◽  
Discrete Element Method ◽  
Reactor Core ◽  
Turbulence Models ◽  
Discrete Element ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Packing Structure ◽  
Dem Simulations

This paper presents combined computational fluid dynamics (CFD) and discrete element method (DEM) simulations of fluid flow and relevant heat transfer in the pebble bed reactor core. In the pebble bed reactor core, the coolant passes highly complicated flow channels, which are formed by thousands of pebbles in a random way. The random packing structure of pebbles is crucial to CFD simulations results. The realistic packing structure in an entire pebble bed reactor (PBR) is generated by discrete element method (DEM). While in CFD calculations, selection of the turbulence models have great importance in accuracy and capturing the details of the flow features, in our numerical simulations both large eddy simulation (LES) and Reynolds-averaged Navier-Stokes (RANS) models are employed to investigate the effects of different turbulence models on gas flow field and relevant heat transfer. The calculations indicate the complex flow structure within the voids between the pebbles.

scholarly journals Study on the effect of various burnable poisons on pebble bed reactor with OTTO fuelling schemes using Serpent 2 Monte Carlo code

2021 ◽  
Vol 927 (1) ◽  
pp. 012018
Author(s):  
Nicholas Sidharta ◽  
Almanzo Arjuna
Keyword(s):  
Power Density ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Code ◽  
Volumetric Fraction ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Burnable Poison ◽  
Discrete Element Method Simulation ◽  
Upper Level

Abstract Pebble bed reactor with a once-through-then-out fuelling scheme has the advantage of simplifying the refueling system. However, the core upper-level power density is relatively higher than the bottom, producing an asymmetric core axial power distribution. Several burnable poison (BP) configurations are used to flatten the peak power density and improve power distribution while suppressing the excess core reactivity at the beginning of the burnup cycle. This study uses HTR-PM, China’s pebble bed reactor core, to simulate several burnable poison (BP) configurations. Serpent 2 coupled with Octave and a discrete element method simulation is used to model and simulate the pebble bed reactor core. It is found that erbium needs a large volumetric fraction in either QUADRISO or distributed BP to perform well. On the other hand, gadolinium and boron need a smaller volumetric fraction but perform worse in radial power distribution criteria in the fuel sphere. This study aims to verify the effect of BP added fuel pebbles on an OTTO refueling scheme HTR-PM core axial power distribution and excess reactivity.

The Re-Evaluation of the AVR Melt-Wire Experiment Using Modern Methods With Specific Focus on Bounding the Bypass Flow Effects

2008 ◽  
Author(s):  
Carel F. Viljoen ◽  
Sonat Sen ◽  
Frederik Reitsma ◽  
Onno Ubbink ◽  
Peter Pohl ◽  
...  
Keyword(s):  
Gas Flow ◽  
Coupled System ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Melting Temperatures ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Flow Effects ◽  
The Many

The AVR (Arbeitsgemeinschaft Versuchsreaktor) is a pebble bed type helium cooled graphite moderated high temperature reactor that operated in Germany for 21 years and was closed down in December 1988 [1]. The AVR melt-wire experiments [2], where graphite spheres with melt-wires of different melting temperatures were introduced into the core, indicate that measured pebble temperatures significantly exceeded temperatures calculated with the models used at the time [3]. These discrepancies are often attributed to the special design features of the AVR, in particular the control rod noses protruding into the core, and to inherent features of the pebble bed reactor. In order to reduce the uncertainty in design and safety calculations the PBMR Company is re-evaluating the AVR melt-wire experiments with updated models and tools. 3-D neutronics thermal-hydraulics analyses are performed utilizing a coupled VSOP99-STAR-CD calculation. In the coupled system VSOP99 [4] provides power profiles on a geometrical mesh to STAR-CD [5] while STAR-CD provides the fuel, moderator and solid structure temperatures to VSOP99. The different fuel histories and flow variations can be modelled with VSOP99 (although this is not yet included in the model) while the computational fluid dynamics (CFD) code, STAR-CD, adds higher-order thermal and gas flow modelling capabilities. This coupling therefore ensures that the correct thermal feedback to the neutronics is included. Of the many possible explanations for the higher-than-expected melt-wire temperatures, flow bypassing the pebble core was identified as potentially the largest contributor and was thus selected as the first topic to study. This paper reports the bounding effects of bypass flows on the gas temperatures in the top of the reactor. It also presents preliminary comparisons between measured temperatures above the core ceiling structure and calculated temperatures. Results to date confirm the importance of correctly modelling the bypass flows. Plans on future model improvements and other effects to be studied with the coupled VSOP99-STAR-CD tool are also included.

scholarly journals STUDY ON MCNP6 MODEL IN THE CALCULATION OF KINETIC PARAMETERS FOR PEBBLE BED REACTOR

2020 ◽  
Vol 60 (2) ◽  
pp. 175-184
Author(s):  
. Zuhair ◽  
. Suwoto ◽  
Topan Setiadipura ◽  
Zaki Su'ud
Keyword(s):  
Kinetic Parameters ◽  
Nuclear Reactor ◽  
Inherent Safety ◽  
Neutron Lifetime ◽  
Uranium Enrichment ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Monte Carlo Transport ◽  
Selection Of ◽  
Parameters Calculation

When conducting a nuclear reactor transient analysis, the most important parameter, called the kinetic parameter, is required. The calculation of kinetic parameters can be conducted using several methods. The deterministic method is one possible method that relies on the forward and adjoint neutron fluxes to provide the kinetic parameters calculation based on the perturbation theory. In this study, the Monte Carlo transport code MCNP6 was utilized to perform the exact prediction of the kinetic parameters of a pebble bed reactor. The core was modelled with a different fuel composition of uranium loading per pebble, <sup>235</sup>U enrichment and H/D ratio. It was found that <em>k</em>eff strongly depends on the uranium loading, uranium enrichment and H/D ratio while the <em>β</em>eff dependence is insignificant. The increase in the prompt neutron lifetime (ℓ) and mean generation time (Ʌ) as a function of H/D ratio are insignificant as compared to the decrease of those parameters in the case of uranium loading or uranium enrichment. These results conclude that the selection of uranium loading per pebble, <sup>235</sup>U enrichment and H/D ratio should be considered carefully for the control and inherent safety performances

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.

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.

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