Coulped neutronics/thermal-hydraulics calculation of VVER-1000 fuel assembly

2021 ◽  
Vol 6 (2) ◽  
pp. 31-38
Author(s):  
Duy Long Ta ◽  
Huy Hiep Nguyen ◽  
Tuan Khai Nguyen ◽  
Vinh Thanh Tran ◽  
Huu Tiep Nguyen
Keyword(s):  
Numerical Calculation ◽  
Fuel Assembly ◽  
Computational Scheme ◽  
Convergence Criterion ◽  
Coupling Scheme ◽  
Thermal Hydraulics ◽  
Script Language ◽  
Master Program ◽  
Perl Script

This paper presents a computational scheme using MCNP5 and COBRA-EN for coupling neutronics/thermal hydraulics calculation of a VVER-1000 fuel assembly. A master program was written using the PERL script language to build the corresponding inputs for the MCNP5 and COBRA-EN calculations and to manage the coupling scheme. The hexagonal coolant channels have been used in the thermal hydraulics model using CORBRA-EN to simplify the coupling scheme. The results of two successive iterations were compared with an assigned convergence criterion and the loop calculation can be broken when the convergence criterion is satisfied. Numerical calculation has been performed based on a UO2fuel assembly of the VVER-1000 reactor.

AREVA’s Advanced Thermal-Hydraulic Codes and Methods for Levels III and IV Crud Risk Assessment

2012 ◽  
Author(s):  
John H. Jones ◽  
Mihnea S. Anghelescu ◽  
Michael S. Bradbury ◽  
Mathieu G. Martin ◽  
Azat Y. Galimov ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Fuel Assembly ◽  
Level Of Detail ◽  
Thermal Hydraulics ◽  
Plant Chemistry ◽  
Hydraulic Process ◽  
Operating Plant

From a crud deposition perspective the achievement of zero fuel failures requires the integration of core neutronics, core thermal-hydraulics, and plant chemistry disciplines. The various level of detail required in the assessment of crud risk is based on guidelines and a checklist developed and published by EPRI. In the guidelines EPRI defined two levels of assessment where core neutronics and thermal-hydraulics are coupled with plant chemistry. These two levels are called Levels III and IV in the EPRI guidelines. AREVA developed a process using the standard licensing tools and a number of specialized application tools and interfaces that allow Level III and IV calculations. The Level III calculations are performed using a typical reload licensing subchannel node scale (scale of several centimeters) whereas the Level IV calculations are performed on fractions of a millimeter scale. This paper provides an overview of the neutronics and thermal-hydraulic process used to perform Levels III and IV assessments and provides some results for a typical B&W-designed 177-fuel assembly operating plant. This same process could be applied to other plant types. The details of the plant chemistry process are not covered in this paper, as they are covered in other publications.

Review of fuel assembly and pool thermal hydraulics for fast reactors

2013 ◽  
Vol 265 ◽  
pp. 1205-1222 ◽  
Author(s):  
Ferry Roelofs ◽  
Vinay R. Gopala ◽  
Santhosh Jayaraju ◽  
Afaque Shams ◽  
Ed Komen
Keyword(s):  
Fuel Assembly ◽  
Thermal Hydraulics ◽  
Fast Reactors

scholarly journals MULTIPHYSICS MODELISATION OF AN UNPROTECTED LOSS OF FLOW TRANSIENT IN A SODIUM COOLED FAST REACTORS USING A NEUTRONIC-THERMAL-HYDRAULIC COUPLING SCHEME

2021 ◽  
Vol 247 ◽  
pp. 07001
Author(s):  
V. Pascal ◽  
Y. Gorsse ◽  
N. Alpy ◽  
K. Ammar ◽  
M. Anderhuber ◽  
...  
Keyword(s):  
Power Distribution ◽  
Spatial Competition ◽  
Spatial Scales ◽  
Coupling Scheme ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Cooling Flow ◽  
The Core ◽  
Feedback Model ◽  
Core Cooling

Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.

Systematic Qualification of the Coupled Neutron Transport and Thermal-Hydraulics Code DORT-TD/THERMIX

2008 ◽  
Author(s):  
Bismark Tyobeka ◽  
Andreas Pautz ◽  
Kostadin Ivanov
Keyword(s):  
Neutron Transport ◽  
Benchmark Problems ◽  
Coupling Scheme ◽  
Thermal Hydraulics ◽  
Comparison Of Results ◽  
Nuclear Systems ◽  
Analysis Computer ◽  
Computer Codes ◽  
Plant Data ◽  
Comparison Of The Results

In order to present credible results in nuclear design and safety analysis, computer codes must adhere to stringent qualification procedures imposed by nuclear licensing authorities. Such procedures form the basis for a quality assured verification and validation process. This is particularly true for advanced nuclear systems of Generation IV type, where little licensing experience exists as well as little or no plant data is available. Qualification of nuclear design and analysis codes can be achieved in various ways, namely: comparison of results from a code with results from another code i.e. code to code benchmarking; comparison of results from a given code with experimental results, i.e. code to experiment benchmarking; comparison of results from a given code with operational plant data; and finally, comparison of the results of a given code with known analytical solutions. In this paper, a systematic qualification of the coupled neutron transport and thermal hydraulics code DORT-TD/THERMIX is presented. As part of developing this coupled code to the level where it can be used as an independent tool by both designers of pebble-bed High-Temperature Gas-cooled Reactors (HTGRs) and regulators, an effort has been made to verify the coupling scheme as well as the validity of application for this code package. At these initial stages a code to code comparison has been adopted as the qualification method of choice. This is done for both steady-state and transient benchmark problems, ranging from simplified to detailed models. As shown in the results section, all benchmarks have been successfully recalculated and generally show good to very good agreement with the “reference” solutions.

scholarly journals WHOLE CORE COUPLING METHODOLOGIES WITHIN WIMS

2021 ◽  
Vol 247 ◽  
pp. 06006
Author(s):  
Brendan Tollit ◽  
Alan Charles ◽  
William Poole ◽  
Andrew Cox ◽  
Glynn Hosking ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Diffusion Flux ◽  
Thermal Hydraulics ◽  
Reactor Physics ◽  
Fuel Performance ◽  
The Core ◽  
Small Modular Reactors ◽  
Core Solution ◽  
Multiphysics Modelling

The ANSWERS® WIMS reactor physics code is being developed for whole core multiphysics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding and pin-by-pin differential depletion. An integrated thermal hydraulic solver permits differential temperature and density variations to feedback to the neutronics calculation. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver. Two coupling routes are presented and compared using a challenging PWR SMR benchmark. The first route, called GEOM, dynamically calculates the resonance shielding and homogenisation with the whole core flux solution. The second coupling route, called CAMELOT, separates the resonance shielding and pincell homogenisation from the whole core solution via generating tabulated cross sections. Both routes can use the MERLIN homogenised pin-by-pin whole core flux solver and couple to the same integrated thermal hydraulic solver, called ARTHUR. Heterogeneous differences between the neutronics and thermal hydraulics are mapped via thermal identifiers for neutronics materials and thermal regions. The ability for the integrated thermal hydraulic solver to call an external code via a Fortran-C-Python (FCP) interface is also summarised. This flexible external coupling permits one way coupling to an external fuel performance code or two way coupling to an external thermal hydraulic code.

scholarly journals SERPENT/SUBCHANFLOW COUPLED BURNUP CALCULATIONS FOR VVER FUEL ASSEMBLIES

2021 ◽  
Vol 247 ◽  
pp. 04005
Author(s):  
Diego Ferraro ◽  
Manuel García ◽  
Uwe Imke ◽  
Ville Valtavirta ◽  
Riku Tuominen ◽  
...  
Keyword(s):  
Neutron Transport ◽  
Nuclear Industry ◽  
Hydraulic Calculation ◽  
Coupling Scheme ◽  
Published Data ◽  
High Fidelity ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Safety Standards ◽  
Wide Range

The continuous improvement in nuclear industry safety standards and reactor designers’ and operators’ commercial goals represent a push for the development of highly accurate methodologies in reactor physics. This fact, combined with the availability of vast computational resources, allowed the development of a wide range of coupled state-of-the-art neutronic-thermal-hydraulic calculation tools worldwide during last decade. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes, suitable for high-fidelity calculations for PWR and VVER reactors. This work presents the results for a pin-by-pin coupled burnup calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel thermal-hydraulics code SUBCHANFLOW (SCF, developed by KIT, Germany) for two different VVER-type fuel assembly types. For such purpose, a recently refurbished master-slave coupling scheme is considered, which provides several new features such as burnup and transient calculations capabilities for square and hexagonal geometries. Main aspects of this coupling are presented for this burnup case, showing some of the capabilities now available. On top of that, the obtained global results are compared with available published data from a similar high-fidelity approach for the same FA design, showing a good agreement. Finally, a brief analysis of the main resources requirement and main bottlenecks identification are also included. The results presented here provide valuable insights and pave the way to tackle the final goals of the McSAFE project, which includes full-core pin-by-pin depletion calculation within a fully coupled MC-TH approach.

CFD Preliminary Assessment of the ALFRED FA Thermal-Hydraulics

2020 ◽  
Author(s):  
R. Marinari ◽  
I. Di Piazza ◽  
M. Tarantino ◽  
G. Grasso ◽  
M. Frignani
Keyword(s):  
Temperature Field ◽  
Fuel Assembly ◽  
Mass Flow ◽  
Flow Distribution ◽  
Test Case ◽  
Thermal Hydraulics ◽  
Safety Studies ◽  
Central Interest ◽  
Nominal Condition ◽  
Temperature Constraint

Abstract In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the coolability of the Fuel Assembly in nominal condition is of central interest. The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) is a 300 MWth pool-type reactor aimed at demonstrating the safe deployment of the Generation IV LFR technology. The ALFRED design, currently being developed by the Fostering ALFRED Construction international consortium, is based on prototypical solutions intended to be used in the next generation of lead-cooled Small Modular Reactors. Within the scope of FALCON and in the frame of investigating the thermal-hydraulics of the ALFRED core, a CFD computational model of the general Fuel Assembly (FA) is built looking for the assessment of its thermal field in nominal flow conditions both for the average FA and the hottest one. Starting from the experience in this kind of simulations and in experimental work, the whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal conditions is carried out. Results showed that the thermal hydraulic field predicted in the average FA by the code is in good agreement with analytical correlations and the temperature field on the pin clad is acceptable for clad material temperature constraint. About the results on the hot FA test case, the CFD results highlighted a peak temperature on the clad close to the clad temperature constraint. This result led to an upgrade of the mass flow distribution among the FA for achieving a 20% mass flow increase in the hottest one that guarantees higher temperature margin on the clad.

Research on the Local Neutronics Thermal-Hydraulic Coupled Response of a PWR During Hypothetical Blockage Scenarios

2014 ◽  
Author(s):  
Xiaorong Li ◽  
Shinian Peng
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Feedback Effect ◽  
Thermal Hydraulics ◽  
Coupled Response ◽  
Flow Blockage

The phenomenon of the partial flow blockage of a fuel assembly in a reactor core is investigated with a coupled 3D neutronics/thermal-hydraulics code in order to account for the space reactivity feedback effect which is of great importance during hypothetical blockage scenarios. This paper identifies the neutronics thermal-hydraulics coupled response in the blocked assembly during the transient and analyzes the details of the phenomenon.

Thermal-hydraulics Analysis of Fuel Assembly with Inner Duct Structure of a Sodium-cooled Fast Reactor

2017 ◽  
Vol 2017.25 (0) ◽  
pp. 305
Author(s):  
Norihiro KIKUCHI ◽  
Yasutomo IMAI ◽  
Ryuji YOSHIKAWA ◽  
Masaaki TANAKA ◽  
Hiroyuki OHSHIMA
Keyword(s):  
Fuel Assembly ◽  
Fast Reactor ◽  
Thermal Hydraulics
Sign in / Sign up

Export Citation Format

Share Document