Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
Zhao Chuanqi ◽  
Wang Kunpeng ◽  
Cao Liangzhi ◽  
Zheng Youqi
Keyword(s):  
Rare Earth ◽  
Fuel Assembly ◽  
Supercritical Water ◽  
Power Distribution ◽  
Rare Earth Oxide ◽  
Burnable Poison ◽  
The Core ◽  
Power Peak ◽  
Supercritical Water Cooled Reactor ◽  
Control Rods

Burnable poison (BP) is used to control excess reactivity in supercritical water cooled reactor (SCWR). It helps reduce the number of control rods. Over all BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for the annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP contents. Core performance with and without BP is compared. The results had shown that the core radial power peaking factor decreased after introducing BP. It was also shown that the core axial power peaking factor increased, and the power peak moved toward the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.

scholarly journals QUALIFICATION OF ANC9 GENERIC INSERT METHODOLOGY (GIM)

2021 ◽  
Vol 247 ◽  
pp. 02003
Author(s):  
Baocheng Zhang ◽  
Boyan D. Ivanov ◽  
Kevin Hoskins ◽  
Harish Huria ◽  
Andre Luiz Pereira Rebello
Keyword(s):  
Fuel Assembly ◽  
Temperature Coefficient ◽  
Power Distribution ◽  
Code System ◽  
Fuel Performance ◽  
Loading Cycle ◽  
Burnable Poison ◽  
The Core ◽  
Burnable Absorber

Discrete burnable poisons like the Wet Annular Burnable Absorber and Pyrex have been used in PWR core to improve core power distribution and provide more negative moderator temperature coefficient. The burnable absorber in the burnable poison rods burns fast and is completely gone after one cycle exposure in the core. Traditionally, the burnable poison rods are designed to be part of the fuel assembly. They are burned together with the fuel in the first loading cycle and discharged after one cycle. In recent years, different insertion scenarios of the burnable poison rods have been introduced in the PWR plants operation to improve the fuel performance, for instance, fresh burnable poison rods are inserted into a burned assembly; burned burnable poison rods stay in the original assembly or are replaced in a different assembly. A Generic Insert Methodology [1] was developed in Westinghouse and implemented in NEXUS/ANC9 code system. With this new methodology, ANC9 is able to follow the assembly history and model all types of absorber insert components for all kinds of insertion scenarios. An extensive methods validation and qualification effort has been completed by modeling different insertion cases. This paper provides details of the qualification cases along with the analysis of the results.

Main Steam Line Break Calculations Using a Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor

2008 ◽  
Author(s):  
Mathias Sta˚lek ◽  
Jo´zsef Ba´na´ti ◽  
Christophe Demazie`re
Keyword(s):  
Steady State ◽  
Fuel Assembly ◽  
Thermal Power ◽  
Steam Line ◽  
Pressurized Water ◽  
The Core ◽  
Different Types ◽  
Main Steam Line ◽  
Main Steam ◽  
Control Rods

A Main Steam Line Break (MSLB) is an important transient for Pressurized Water Reactors (PWR) due to the strong positive reactivity introduced by the over-cooling of the core. Since this effect is stronger when the Moderator Temperature Coefficient (MTC) has a large amplitude, a conservative result will be obtained for a high burnup of the fuel due to the more negative MTC late in the cycle. The calculations have been performed at a cycle burnup of 12.9742 GWd/tHM. The Swedish Ringhals-3 PWR is a three loop Westinghouse design, currently with a thermal power of 3000 MW. The PARCS model has 157 fuel assemblies of 8 different types. Four different types of reflector are used. The cross sections, and kinetic data were obtained from CASMO-4 calculations, using a cross section interface developed at the department. There are 24 axial nodes, and 2×2 radial nodes for each assembly. The transient option for calculating the effect of poisoning was used. The PARCS model has been validated against steady-state measurements from Ringhals-3 of the Relative Power Fraction (RPF) and of the core criticality. The RELAP5 model has 157 channels for the core which means that there is a one to one correspondence between the thermal hydraulics model and the neutronics model. There is eight axial nodes. Originally, the intention was to have 24 axial nodes but this proved not to work because of some limitation in RELAP5. There is currently no mixing between the different channels in the core. The feedwater, and turbines are modelled as boundary conditions. The stand-alone RELAP5 model has been validated against steady state measurements from Ringhals-3. A number of different cases were considered. In the first case, both the isolation of the feedwater for the broken loop, and all the control rods were assumed to work properly. For the second case one of the control rods was assumed to be stuck. The stuck rod was located in the fuel assembly with the highest power. This rod has also one of the highest rod worths. In the final case, the feedwater control valve for the broken loop was fully open. None of the cases led to any recriticality. The increase in power for each fuel assembly was also investigated. With the control rod located in the assembly with the highest power, the maximum power increase before scram turned out to be about 25% compared to the initial power.

Improving on Assembly of SCWR Using MOX Fuel

2011 ◽  
Vol 347-353 ◽  
pp. 1633-1636 ◽  
Author(s):  
Can Hui Sun ◽  
Tao Zhou ◽  
Zhou Sen Hou ◽  
Meng Ying Liu ◽  
Feng Luo
Keyword(s):  
Supercritical Water ◽  
Power Distribution ◽  
Security Issue ◽  
Original Design ◽  
Fuel Rod ◽  
Simple Power ◽  
Mox Fuel ◽  
Higher Temperature ◽  
The One ◽  
Supercritical Water Cooled Reactor

A calculation is made for certain Supercritical Water Cooled Reactor (SCWR) using UO2 fuel and MOX fuel respectively. The results indicate that MOX fuel has a simple power distribution with UO2 fuel, but there is a larger power uneven factor when using MOX fuel, and using MOX fuel including weapon grade Pu has larger power uneven factor than using MOX fuel including reactor grade Pu. However, in the case of same power distribution, the fuel rod using MOX fuel has a higher temperature than the one using UO2 fuel. Therefore with the more uneven power distribution, the fuel in SCWR using MOX fuel has a higher temperature. This will result in a big security issue when using MOX fuel in original design of SCWR. Through analyzing the result of power distribution, an improved assembly of SCWR is presented. It can reduce the power uneven factor and increase the security of fuel rod using the improved assembly of SCWR.

scholarly journals Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

2014 ◽  
Vol 2014 ◽  
pp. 1-8
Author(s):  
Po Hu ◽  
Paul P. H. Wilson
Keyword(s):  
Steady State ◽  
Supercritical Water ◽  
Power Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Core Model ◽  
Core Flow ◽  
The Core ◽  
Extended Code ◽  
Reference Design

This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

Fission Product Release Under Supercritical Water-Cooled Reactor Conditions

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
D. Guzonas ◽  
L. Qiu ◽  
S. Livingstone ◽  
S. Rousseau
Keyword(s):  
Supercritical Water ◽  
Fission Products ◽  
Fuel Pellet ◽  
Product Release ◽  
The Core ◽  
Fuel Pellets ◽  
Simulated Fuel ◽  
Core Components ◽  
High Concentrations ◽  
Supercritical Water Cooled Reactor

Most supercritical water-cooled reactor (SCWR) concepts being considered as part of the Generation IV initiative are direct cycle. In the event of a fuel defect, the coolant will contact the fuel pellet, potentially releasing fission products and actinides into the coolant and transporting them to the turbines. At the high pressure (25 MPa) in an SCWR, the coolant does not undergo a phase change as it passes through the critical temperature in the core, and nongaseous species may be transported out of the core and deposited on out-of-core components, leading to increased worker dose. It is therefore important to identify species with a high risk of release and develop models of their transport and deposition behavior. This paper presents the results of preliminary leaching tests in SCW of U-Th simulated fuel pellets prepared from natural U and Th containing representative concentrations of the (inactive) oxides of fission products corresponding to a fuel burnup of 60  GWd/ton. The results show that Sr and Ba are released at relatively high concentrations at 400°C and 500°C.

Preparation and Study of Rare Earth Oxide Nanoparticles with Core-Shell Structure

2013 ◽  
Vol 813 ◽  
pp. 332-335
Author(s):  
Mei Gui Ou ◽  
Chun Lin Yang ◽  
Shao Han Cai ◽  
Qi Wei Zhu
Keyword(s):  
Energy Transfer ◽  
Rare Earth ◽  
Precursor Solution ◽  
Rare Earth Oxide ◽  
Oxide Nanoparticles ◽  
Core Shell ◽  
Shell Nanoparticles ◽  
The Core ◽  
Core Shell Structure ◽  
Core Shell Nanoparticles

Core-shell nanoparticles Gd2O3:Tb3+/SiOx were obtained by encapsulating Gd2O3:Tb3+ in a polysiloxane shell. We studied the influence of two kinds of reagents (NaOH and Bu4NOH) reacting with precursor solution on size and luminescent property of nanoparticles. The result showed that the reaction involving NaOH was more favorable to the growth of nanoparticles, thus enhanced the energy transfer between the core and the shell of particles and improved their luminescent intensities.

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 Calculation of excore detector weighting functions for a sodium-cooled TRU burner mockup using MCNP5

2015 ◽  
Vol 5 (2) ◽  
pp. 15-25
Author(s):  
Viet Ha Pham Nhu ◽  
Min Jae Lee ◽  
Sunghwan Yun ◽  
Sang Ji Kim
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Detector Response ◽  
Weighting Functions ◽  
Control Rod ◽  
The Core ◽  
Power Regulation ◽  
Fuel Assemblies ◽  
Control Rods ◽  
Horizontal Layers

Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A.

scholarly journals ANTARES: COUPLING PARCS WITH CATHARE-3

2021 ◽  
Vol 247 ◽  
pp. 07005
Author(s):  
M. Forestier ◽  
G. Girault ◽  
F. Jacq ◽  
A. Sargeni
Keyword(s):  
Power Distribution ◽  
Safety Analysis ◽  
Conservative Approach ◽  
Reactor Safety ◽  
The Core ◽  
Moderator Effects ◽  
Current Technique ◽  
Power Peak ◽  
Improved Accuracy ◽  
Physical Quantities

In recent years, the IRSN has launched a new project to couple the first 3D version of the thermal hydraulic code CATHARE-3 (system) with the 3D, neutronic nodal code PARCS (core): ANTARES (Advanced Neutronics and Thermal-hydraulic for the Analysis of the Reactor Safety). The purpose of this project is to increase the IRSN capability to couple different codes, to calculate the core power distribution in CATHARE-3 and to improve the thermal hydraulic boundaries conditions in PARCS. In this way, the IRSN diversifies its available tools to perform safety analysis with improved accuracy. The current technique usually adopted in France for the safety demonstrations is the so-called ‘conservative' approach, which consists of reducing all the feedback (Doppler and moderator effects) and in modifying some physical quantities in such a way to increase a power peak in an accidental transient. For this reason, these facilities (‘penalties’) have been implemented in ANTARES. In this paper we will give two examples of accidental transients that can be simulated with ANTARES: a REA (Rod Ejection Accident) and an inadvertent boron dilution event.

Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Laurence K. H. Leung ◽  
Yanfei Rao ◽  
Krishna Podila
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Supercritical Water ◽  
Supercritical Pressure ◽  
Heat Transfer Correlation ◽  
Computational Tools ◽  
Experimental Heat ◽  
Experimental Heat Transfer ◽  
Transfer Behavior ◽  
Supercritical Water Cooled Reactor

Experimental data and correlations are not available for the fuel-assembly concept of the Canadian supercritical water-cooled reactor (SCWR). To facilitate the safety analyses, a strategy for developing a heat-transfer correlation has been established for the fuel-assembly concept at supercritical pressure conditions. It is based on an analytical approach using a computational fluid dynamics (CFD) tool and the ASSERT subchannel code to establish the heat transfer in supercritical pressure flow. Prior to the application, the CFD tool was assessed against experimental heat transfer data at the pseudocritical region obtained with bundle subassemblies to identify the appropriate turbulence model for use. Beyond the pseudocritical region, where the normal heat transfer behavior is anticipated, the ASSERT subchannel code also was assessed with appropriate closure relationships. Detailed information on the supporting experiments and the assessment results of the computational tools are presented.

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