Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Vitali Kovaltchouk ◽  
Eleodor Nichita ◽  
Eugene Saltanov
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Pressure Tube ◽  
Coolant Temperature ◽  
Outlet Temperature ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Fuel Channel ◽  
Technical Specifications ◽  
Institute Of Technology

The axial power and coolant-temperature distributions in a fuel channel of the Generation IV pressure-tube super-critical water-cooled reactor (PT-SCWR) are found using coupled neutronics-thermal-hydraulics calculations. The simulations are performed for a channel loaded with a fresh, 78-element Th-Pu fuel assembly. Neutronics calculations are performed using the DONJON diffusion code using two-group homogenized cross sections produced using the lattice code DRAGON. The axial coolant temperature profile corresponding to a certain axial linear heat generation rate is found using a code developed in-house at University of Ontario Institute of Technology (UOIT). The effect of coolant density, coolant temperature, and fuel temperature variation along the channel is accounted for by generating macroscopic cross sections at several axial positions. Fixed-point iterations are performed between neutronics and thermal-hydraulics calculations. Neutronics calculations include the generation of two-group macroscopic cross sections at several axial positions, taking into account local parameters such as coolant temperature and density and average fuel temperature. The coolant flow rate is adjusted so that the outlet temperature of the coolant corresponds to the SCWR technical specifications. The converged axial power distribution is found to be asymmetric, resembling a cosine shape skewed toward the inlet (reactor top).

A New Method on Flux Mapping by Function Expansion

2010 ◽  
Author(s):  
Bin Zhong ◽  
Kan Wang ◽  
Ganglin Yu
Keyword(s):  
Power Distribution ◽  
Synthesis Method ◽  
Expansion Method ◽  
Mapping Method ◽  
Least Square ◽  
New Method ◽  
Coolant Temperature ◽  
Fuel Temperature ◽  
Function Expansion ◽  
Flux Mapping

The core flux (power) distribution is very important to safe and economical operation of nuclear reactor. It can be obtained by many methods depending on the desired accuracy and execution time. For on-line core surveillance and regulation, we need to get the real-time flux distribution. If the true local parameters such as fuel temperature, coolant temperature and material density were known, the solution of the diffusion equation with instantaneous parameters could, in principle, provide the necessary spatial details. However, in reality, it is impossible to obtain the operational “readings” of these parameters for each fuel cell. The detector results at certain locations can be applied to improve the results of the only diffusion calculations by Flux Mapping methods. Function expansion method is employed to express the approximate real distribution by the combination of several Flux Mapping method results as the expansion basis functions. The Harmonics Synthesis Method (HSM) and Least-Square method are combined to get a new Flux Mapping method in this paper. The simulation results show that the new method can be used for Flux Mapping and get better results.

Neutronics and Thermal Hydraulics Coupled 3-D Calculation of Pressure-Tube Deformation

2020 ◽  
Author(s):  
Eunhyun Ryu ◽  
Hangyu Joo ◽  
Seungyul Yoo ◽  
Jongyub Jung
Keyword(s):  
Dimensional Space ◽  
Three Dimensional ◽  
Safety Margin ◽  
Pressure Tube ◽  
Coolant Temperature ◽  
Main Concern ◽  
Thermal Hydraulics ◽  
Pressure Tubes ◽  
Fuel Pin ◽  
Heavy Water Reactor

Abstract Among the various parts in a pressurized heavy-water reactor (PHWR), pressure tubes are of tremendous importance. This is because they withstand extreme both pressure and temperature differences that exist between the Primary Heat Transport System (PHTS) and the moderator. The pressure tubes also contribute to prevention of fission product release from the PHTS to the PHWR plant (together with end fittings and nearby parts including plugs). When a PHWR is given a 1% derating, half is due to the aging of the pressure tubes. The main concern with pressure tubes is decrease of the safety margin. Most of the reduction comes from the effects caused by radial expansion and axial sagging, which are belong to four major phenomena including the thinning and the elongation. More specifically, the fuel-pin temperature distribution changes for the worse if deformation of the pressure tube occurs. Because there is extreme irradiation inside the core, the tube content is exposed to high temperature and high pressure. Thus, the shape of the pressure tube is deformed as times goes on. In this paper, using modeling of a deformed pressure tube in three-dimensional space, the effects on the fuel, coolant temperature, and coolant density, were studied quantitatively. This included a neutronics effect explored using coupled neutronics and thermal hydraulics (T/H) calculations. Among the results, only marginal changes of the neutronics effects were observed. The T/H results, which included temperature and density of the fuel and the coolant, were not critical. Through this study, we are now able to determine in new ways, conventional derating values from a pressure tube.

scholarly journals MULTI-PHYSICS TRANSIENT SIMULATIONS WITH TRIPOLI-4®

2021 ◽  
Vol 247 ◽  
pp. 07019 ◽  
Author(s):  
Margaux Faucher ◽  
Davide Mancusi ◽  
Andrea Zoia
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Coolant Temperature ◽  
Test Case ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Control Rod ◽  
Transient Simulations ◽  
Neutron Power

In this work, we present the first dynamic calculations performed with the Monte Carlo neutron transport code TRIPOLI-4R with thermal-hydraulics feedback. For this purpose, the Monte Carlo code was extended for multi-physics capabilities and coupled to the thermal-hydraulics subchannel code SUBCHANFLOW. As a test case for the verification of transient simulation capabilities, a 3x3-assembly mini-core benchmark based on the TMI-1 reactor is considered with a pin-by-pin description. Two reactivity excursion scenarios initiated by control-rod movement are simulated starting from a critical state and compared to analogous simulations performed using the Serpent 2 Monte-Carlo code. The time evolution of the neutron power, fuel temperature, coolant temperature and coolant density are analysed to assess the multi-physics capabilities of TRIPOLI-4. The stabilizing e_ects of thermal-hydraulics on the neutron power appear to be well taken into account. The computational requirements for massively parallel calculations are also discussed.

Thermal Evaluation for a Ultra High Temperature Reactor With Pebble Type Fuels

2009 ◽  
Author(s):  
Motoo Fumizawa ◽  
Yuta Kosuge ◽  
Hidenori Horiuchi
Keyword(s):  
High Temperature ◽  
Power Density ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Small Scale ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
System Pressure ◽  
High Temperature Reactor ◽  
Ultra High Temperature

This study presents a predictive thermal-hydraulic analysis with packed spheres in a nuclear gas-cooled reactor core. The predictive analysis considering the effects of high power density and the some porosity value were applied as a design condition for an Ultra High Temperature Reactor (UHTR). The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 °C was achieved in the case of an UHTREX at LASL, which was a small scale UHTR using hollow-rod as a fuel element. In the present study, the fuel was changed to a pebble type, a porous media. Several calculation based on HTGR-GT300 through GT600 were 4.8 w/cm3 through 9.6 w/cm3 respectively. As a result, the relation between the fuel temperature and the power density was obtained under the different system pressure and coolant outlet temperature. Finally, available design conditions are selected.

Materials and Stress Analysis of Fuel-Channel of a Generic 1200-MWel Pressure-Channel Reactor With Nuclear Steam Reheat

2014 ◽  
Author(s):  
Rand Abdullah ◽  
Matthew Baldock ◽  
Andrei Vincze ◽  
Khalil Sidawi ◽  
Igor Pioro
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Tube ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Fuel Channel ◽  
Suitable Material ◽  
Inconel 600 ◽  
Fuel Bundle ◽  
Ss 304

The objective of this paper is to modify the current existing CANFLEX® fuel-bundle design to examine its ability to withstand high-temperature conditions of a proposed generic reactor with nuclear steam reheat. One of this reactor’s characteristics is having Super-Heated Steam (SHS) channels in addition to Pressurized-Water (PW) channels in order to increase the thermal efficiency of the plant by about 7–12%. This increase may be attained by raising the outlet temperature of the SHS-channels coolant to about 550°C. Operating at the higher temperatures will definitely have an effect on the mechanical and neutronic properties of fuel-channel materials, specifically on fuel-sheath and pressure-tube materials. This paper compares Inconel-600 and SS-304 in order to determine the most suitable material for SHS-channel’s sheath and pressure tube. This is achieved by comparing strength of materials by performing stress- and displacement-analysis simulation using NX8.5 software (NX8.5, 2009). The analysis in this paper can also be applied to other Nuclear-Power Plants (NPPs) that require operating at higher temperatures such as Super-Critical Water-cooled Reactors (SCWRs).

A Fuel Channel Design for CANDU-SCWR

2006 ◽  
Author(s):  
C. K. Chow ◽  
S. J. Bushby ◽  
H. F. Khartabil
Keyword(s):  
Corrosion Resistance ◽  
Material Selection ◽  
Porous Ceramic ◽  
Pressure Tube ◽  
Thermodynamic Efficiency ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Fuel Channel ◽  
Pressure Tubes ◽  
Metal Liner

The CANDU®-Supercritical Water Reactor (CANDU-SCWR) is one of the six reactor concepts being considered by the Generation-IV International Forum (GIF) for international collaborative R&D. With SCW coolant, the thermodynamic efficiency is increased to over 40%. The CANDU-SCWR is moderated using heavy water, and it has fuel bundles residing inside horizontal pressure tubes, similar to the current CANDU design. The coolant, however, is light water at 25 MPa, with an inlet temperature of 350°C and an outlet temperature of 625°C. Because of the high temperature and high pressure of the coolant, the standard CANDU pressure tube design cannot be used. This paper presents one of the insulated pressure tube designs being considered for the CANDU-SCWR fuel channels. Unlike current CANDU reactors, the proposed CANDU-SCWR fuel channel does not use calandria tubes to separate the pressure tubes from the moderator. Each pressure tube is in direct contact with the moderator, which operates at an average temperature of about 80°C. The pressure tube is thermally insulated from the hot coolant by a porous ceramic insulator. A perforated metal liner protects the insulator from being damaged by the fuel bundles and erosion by the coolant. The coolant pressure is transmitted through the perforated metal liner and insulator and applied directly to the relatively cold pressure tube. The material selection for each fuel channel component depends on its function. The fuel sheaths and the perforated liner must have high corrosion resistance in SCW, although their resident times are significantly different. The insulator must have high thermal resistance and corrosion resistance in SCW, plus sufficient strength to bear the weight of the fuel bundles without significant thickness reduction during its design life. The pressure tube is the pressure boundary material, so it must have high strength to contain the coolant. One common requirement for all in-core fuel channel components is that they should be as neutron transparent as possible. The irradiation deformation of all these components must also be considered in their design. This paper presents the design of this fuel channel, reviews existing data for materials, indicates where more data are required, and summarizes our plans to obtain these data.

scholarly journals ANALYSIS OF UNCONTROLLED REACTIVITY INSERTION TRANSIENT OF TRIGA MARK 2000 BANDUNG USING MTR PLATE TYPE FUEL ELEMENT

2020 ◽  
Vol 23 (2) ◽  
pp. 69
Author(s):  
Surian Pinem ◽  
Tukiran Surbakti ◽  
Imam Kuntoro
Keyword(s):  
Fuel Element ◽  
Cross Sections ◽  
Feedback Mechanism ◽  
Heat Fluxes ◽  
Coolant Temperature ◽  
Fuel Temperature ◽  
Element Analysis ◽  
Reactivity Insertion ◽  
Type Fuel ◽  
Plate Type

ANALYSIS OF UNCONTROLLED REACTIVITY INSERTION TRANSIENT OF TRIGA MARK 2000 BANDUNG USING MTR PLATE TYPE FUEL ELEMENT. Analysis of uncontrolled reactivity insertion is very important for the safety of reactor operations. Determination of melting point limit, critical heat fluxes and melting temperatures of cladding are the main objectives for most of these studies to determine whether fuel temperature can withstand the transient insertion of reactivity. In this study, uncontrolled reactivity insertion transient was carried out due to the withdrawal of control rods in nominal power of 1 MW and 2 MW. Analysis of reactivity transient was carried out using the WIMSD/5B and MTRDYN codes. The WIMSD/5B code is used to generate cross sections and the MTRDYN program is used for analysis under transient conditions. Based on calculations on the initial power of 1 MW and 2 MW with an insertion of reactivity of greater than 0.5 $/s the reactor operation  is not safe because the fuel temperature exceeds the design limit. For reactivity insertion 0.5 $/s allows increased power can be stabilized by feedback reactivity. For 1 MW of nominal power, the maximum coolant temperature,  cladding and fuel are 86.39 oC, 164.86 oC and 165.33 oC, respectively. For 2 MW of nominal power,  the maximum coolant temperature,  cladding and fuel are 89.09 oC, 176.96 oC and 177.602 oC, respectively. Based on calculation,  It is concluded that the feedback mechanism can protect the fuel cladding from a local meltdown if reactivity insertion 0.5 $/s and the reactor is in nominal power of 1 MW and 2 MW.

scholarly journals The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Surian Pinem ◽  
Tagor Malem Sembiring ◽  
Peng Hong Liem
Keyword(s):  
Coolant Temperature ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Control Rod ◽  
3 Dimensional ◽  
Nodal Method ◽  
Benchmark Calculation ◽  
Power Peak ◽  
Calculation Results ◽  
Neutron Diffusion Equation

A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised). Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

scholarly journals Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor

2014 ◽  
Vol 4 (1) ◽  
pp. 10-25
Author(s):  
Ba Vien Luong ◽  
Vinh Vinh Le ◽  
Ton Nghiem Huynh ◽  
Kien Cuong Nguyen
Keyword(s):  
Steady State ◽  
Power Distribution ◽  
Research Reactor ◽  
Nuclear Research ◽  
Coolant Temperature ◽  
Thermal Hydraulics ◽  
Kinetics Parameters ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Full Core

The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum  hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.

Influence Analysis of Coupled Neutronics and Thermo-Hydraulics on Steady-State Characteristics of Supercritical Water-Cooled Reactor

2012 ◽  
Vol 472-475 ◽  
pp. 278-283
Author(s):  
Juan Chen ◽  
Tao Zhou ◽  
Zhou Sen Hou ◽  
Wan Xu Cheng ◽  
Can Hui Sun
Keyword(s):  
Steady State ◽  
Supercritical Water ◽  
Cross Sections ◽  
Power Distribution ◽  
Coupled Model ◽  
Calculation Model ◽  
Coolant Temperature ◽  
Coupling Condition ◽  
Coupling Calculation ◽  
Supercritical Water Cooled Reactor

the coulped neutronics and thermo-hydraulics model for supercritical water-cooled reactor (SCWR) is developed by internal coupling method. It is based on the two group neutron diffusion equations and the one-dimensional junction thermal analysis mode, in which the cross sections used for SCWR are generated by Dragon tool. Compared with the calculation results based on the non-coupling calculation model, the steady state characteristics under coupling calculation condition are detailed analyzed by considering parameters feedback at each axial node. The results show that, as coupled model is chosen its axial power distribution would give an obvious deviation from the cosine function that used for non-coupled model. Although the cladding temperature at most of the axial nodes rises with a shifted power peak, the maximum cladding temperature is finally decreased. For the above coupling condition, the maximum cladding temperature would appear at the external assemblies with lower coolant temperature but not at inner assemblies with higher coolant temperature. As the detailed description for coupling characteristics of supercritical water-cooled reactor is given, a certain theory reference for its system safety could be provided.

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