scholarly journals Thermo-hydraulic Simulation of AP1000 Nuclear Reactor Fuel Assembly

2021 ◽  
Vol 31 (1) ◽  
pp. 60-71
Author(s):  
Leonardo Acosta Martínez ◽  
Carlos Rafael García Hernández ◽  
Jesus Rosales García ◽  
Annie Ortiz Puentes
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Capital Investment ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Geometric Model ◽  
Modular Construction ◽  
Construction Time ◽  
Hydraulic Simulation ◽  
In Series

One of the challenges of future nuclear power is the development of safer and more efficient nuclear reactor designs. The AP1000 reactor based on the PWR concept of generation III + has several advantages, which can be summarized as: a modular construction, which facilitates its manufacture in series reducing the total construction time, simplification of the different systems, reduction of the initial capital investment and improvement of safety through the implementation of passive emergency systems. Being a novel design it is important to study the thermohydraulic behavior of the core applying the most modern tools. To determine the thermohydraulic behavior of a typical AP1000 fuel assembly, a computational model based on CFD was developed. A coupled neutronic-thermohydraulic calculation was performed, allowing to obtain the axial power distribution in the typical fuel assembly. The geometric model built used the certified dimensions for this type of installation that appear in the corresponding manuals. The thermohydraulic study used the CFD-based program ANSYS-CFX, considering an eighth of the fuel assembly. The neutronic calculation was performed with the program MCNPX version 2.6e. The work shows the results that illustrate the behavior of the temperature and the heat transfer in different zones of the fuel assembly. The results obtained agree with the data reported in the literature, which allowed the verification of the consistency of the developed model.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

scholarly journals Fingerprinting commercial nuclear reactor forensics using the ORIGEN-S simulation code

2018 ◽  
Vol 48 ◽  
pp. 1860126
Author(s):  
Iyabo Usman ◽  
David Vermillion ◽  
Howard Hall ◽  
Steve Skutnik
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Reference Database ◽  
Cycle Period ◽  
Isotopic Ratios ◽  
Spent Fuel ◽  
Concentration Data ◽  
Simulation Code

The ability to determine the origin of a specific spent-fuel sample from a commercial nuclear reactor was studied using the Origen-S simulation code by calculating the plutonium and uranium isotopic concentration data for a range of nuclear power reactors. This range of reactors is based on a typical Westinghouse PWR fuel assembly with a fuel type of W17 X 17, having individual operating histories. Isotopic ratios of plutonium in nuclear reactors during the fuel-cycle period provide information on how the plutonium grows into the fuel as a function of burnup, as well as its attractiveness to proliferators. Using the results from the calculation of uranium and plutonium isotopic ratios, the origin of each spent-fuel assembly for a particular reactor can be predicted and documented for a future nuclear forensics reference database.

Power Distribution and Fuel Centerline Temperature in a Pressure-Tube Supercritical Water-Cooled Reactor (PT SCWR)

2012 ◽  
Author(s):  
W. Peiman ◽  
Eu. Saltanov ◽  
L. Grande ◽  
I. Pioro ◽  
B. Rouben ◽  
...  
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Pressure Tube ◽  
Thermochemical Processes ◽  
Flux Profile ◽  
Centerline Temperature

SuperCritical Water-cooled nuclear Reactor (SCWR) designs are one of six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic pressure-tube SCWR consists of distributed fuel channels with coolant inlet and outlet temperatures of 350 and 625°C at 25 MPa, respectively. Such reactor coolant outlet conditions allow for high thermal efficiencies of SCW Nuclear Power Plant (NPP) of about 45–50%. In addition to high thermal efficiencies, SCWR designs provide the means for co-generation of hydrogen through thermochemical processes such as the Cu–Cl cycle. The main objective of this paper is to determine the power distribution inside the core of an SCWR by using a lattice code - DRAGON and a diffusion code - DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermal-hydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature for UO2 and UC nuclear fuels. Results of an analysis showed that the fuel centerline temperature of UC was significantly lower than that of UO2. This paper also studies effects of energy groups on multi-group diffusion calculations and proposes nine energy groups for further neutronic studies related to SCWRs.

Sodium Cooled Fast Bread Reactor HCDA Codes Development

2017 ◽  
Author(s):  
Shi Tai ◽  
Zhang Dong-hui ◽  
Hu Wen-jun
Keyword(s):  
Liquid Metal ◽  
Fast Reactor ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Radioactive Material ◽  
Program Model ◽  
Primary Method ◽  
The World ◽  
Gen Iv

Liquid metal fast reactor is one of the Gen IV nuclear power, it is necessary to analyze hypothetical core disruptive accident (HCDA) of FBR to ensure that the system can prevent the radioactive material from leaking out. The modified Bethe-Tait model is the primary method to analyze hypothetical core disruptive accident in the world. In order to better analyze the nuclear reactor hypothetical core disruptive accident in China, an improved B-T model is used. At present, on the basis of the improved B-T model, power distribution of the CEFR add to the progress. The results of comparison between the program and SUREX program in France show that the program model can simulate the nuclear reactor hypothetical core disruptive accident in China.

Coolant Mixing in the VVER-440 Fuel Assembly Head

2014 ◽  
Author(s):  
Tomas Romsy ◽  
Pavel Zacha
Keyword(s):  
Temperature Field ◽  
Computational Model ◽  
Fuel Assembly ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Coolant Temperature ◽  
Ansys Fluent ◽  
Uniform Temperature Field ◽  
Mixing Character ◽  
Power Outputs

The issue of the temperature measurement in a nuclear reactor is an important element to ensure safe operation of the nuclear power plant. To prevent damages and radioactive releases the fuel in the reactor must be continuously cooled. The coolant temperature field of the VVER-440 reactor is measured with thermocouples installed at the outlet part of fuel assemblies. Since the power output of the fuel pins is not equal, a non-uniform temperature field at the inlet of the fuel assembly head is formed. Next, the temperature field is subsequently mixed by passing through the assembly head, which contains some constructional elements helping to mixing coolant flow. This mixing is not perfect and due to the effect described above, the signal on the thermocouple can be affected. This phenomenon was introduced in 2009 Atomic Energy Research (AER) Symposium in Bulgaria. 7 international institutions participated with the main goal to explain the mixing character of the coolant and to compare results. For further study of this phenomenon the new detailed computational model of the upper part of the fuel assembly was created and subsequently on the ANSYS FLUENT CFD code verified. The main output of these simulations is study of the coolant temperature distribution on the thermocouple. Computational model, based on the source geometry given by AER symposium, was created in preprocessor GAMBIT 2.4.6. Model contains over 13 million hexahedral cells. Thermohydraulic simulations where performed in ANSYS FLUENT v14.5 and results were compared with data from AER Symposium. There were considered two cases with different pins power outputs. With compare to AER symposium results the achieved resultant temperature on the thermocouple position for both cases indicate comparable accuracy. Furthermore, some flow fluctuations in the assembly head area where found.

scholarly journals Optimization of a Fuel Assembly for Supercritical Water-Cooled Reactor CSR1000

2021 ◽  
Vol 9 ◽  
Author(s):  
Fawen Zhu ◽  
Lele Zheng ◽  
Quan-Yao Ren ◽  
Ti Yue ◽  
Hua Pang ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Terminal Velocity ◽  
Structure Design ◽  
Control Rod ◽  
And Performance ◽  
Supercritical Water Cooled Reactor

As one of the Generation IV nuclear reactors, the SCWR (supercritical water-cooled reactor) has high economy and safety margin, good mechanical properties for its high thermal efficiency, and simplified structure design. As the key component of nuclear reactor, the fuel assembly has always been the main issue for the design of the SCWR. The design of the fuel assembly for CSR1000 proposed by the Nuclear Power Institute of China (NPIC) has been optimized and presented in this study, which is composed of four subassemblies welded by four filler strips and guide thimbles arranged close together in the cross-shaped passage. Aiming at improving the hydraulic buffer performance of the cruciform control rod, the scram time and terminal velocity of control rod assembly were calculated to assess the scram performance based on the computational fluid dynamics and dynamic mesh method, and the mechanical property and neutronic performance of assemblies were also investigated. It has been demonstrated that the optimized fuel assembly had good feasibility and performance, which was a promising design for CSR1000.

scholarly journals KYLIN-V2.0 CODE CALCULATION ABILITY VERIFICATION BASED ON VERA BENCHMAR

2021 ◽  
Vol 247 ◽  
pp. 10020
Author(s):  
Dongyong Wang ◽  
Yingrui Yu ◽  
Xingjie Peng ◽  
Chenlin Wang ◽  
Kun Liu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Nuclear Power ◽  
Neutron Transport ◽  
Monte Carlo Code ◽  
Neutron Transport Equation ◽  
Control Rod ◽  
Fuel Assemblies ◽  
Fuel Pin ◽  
Calculation Ability

Virtual Environmental for Reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance selfshielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference.

scholarly journals Off-site modular construction and design in nuclear power: A systematic literature review

2021 ◽  
Vol 134 ◽  
pp. 103664
Author(s):  
P.A. Wrigley ◽  
P. Wood ◽  
S. O'Neill ◽  
R. Hall ◽  
D. Robertson
Keyword(s):  
Literature Review ◽  
Systematic Literature Review ◽  
Nuclear Power ◽  
Modular Construction

scholarly journals Computational Fluid Dynamics Modeling of Steam Condensation on Nuclear Containment Wall Surfaces Based on Semiempirical Generalized Correlations

2012 ◽  
Vol 2012 ◽  
pp. 1-7 ◽  
Author(s):  
Pavan K. Sharma ◽  
B. Gera ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Pressure Rise ◽  
Dynamics Modeling ◽  
Steam Condensation ◽  
Noncondensable Gases ◽  
Nuclear Containment ◽  
Design Basis ◽  
Lumped Parameter ◽  
Transport Calculation

In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.

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