scholarly journals COUPLED TRANSIENT ANALYSIS OF A CORE WITH FUEL ASSEMBLY BOWING WITH A HYBRID CTF/DYN3D MODEL

2021 ◽  
Vol 247 ◽  
pp. 06036
Author(s):  
Y Périn ◽  
A Travleev ◽  
M Zilly
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Extreme Case ◽  
Nucleate Boiling ◽  
Reconstruction Method ◽  
Thermal Hydraulics ◽  
Safety Criterion ◽  
Coolant Pump ◽  
Active Length ◽  
Nodal Diffusion

Fuel assembly bowing is a known phenomenon observed in many PWR reactors all over the world. The phenomenon is relevant to safety as it can lead to increased water gaps between assemblies which results in higher pin peaking factors. The goal of the present study is to assess the effect of assembly bowing not only for stead-state nominal conditions but also during a transient. The selected transient is the loss of one reactor coolant pump as it can be limiting especially regarding the Departure from Nucleate Boiling (DNB) safety criterion. This study focuses on an extreme case where the bowing is simulated in the core hot assembly by keeping the water gap constant over the whole core active length. The resulting cross-sections and form functions obtained from a 2d infinite lattice model are used in the nodal diffusion code DYN3D applying its pin-by-pin reconstruction method. For the transient simulation, DYN3D is coupled with the thermal-hydraulics subchannel code CTF on the SALOME platform. Several modelling options are compared: nominal geometry for neutronics and thermal-hydraulics (TH); mixed: neutronics with increased water gap, TH with nominal geometry; and increased water gap for both neutronics and TH. The results confirm that the increased water gap should be considered in both models in order to reduce the conservatism.

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.

Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

2018 ◽  
Author(s):  
Yuchuan Guo ◽  
Guanbo Wang ◽  
Dazhi Qian ◽  
Heng Yu ◽  
Bo Hu
Keyword(s):  
Fuel Assembly ◽  
System Analysis ◽  
Research Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Regulation System ◽  
Lower Power ◽  
Emergency Shutdown ◽  
Flow Blockage

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

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.

The Analysis of Influence of Fuel Pellet and Coolant Temperature Distributions on the VVER-1000 Active Core Characteristics

2013 ◽  
Author(s):  
A. A. Mishin ◽  
V. V. Galchenko
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fuel Pellet ◽  
Coolant Temperature ◽  
Radial Temperature ◽  
Group Constant ◽  
Temperature Distributions ◽  
Active Core ◽  
Core Characteristics

The accuracy and quality of neutron-physical calculations of the active core characteristics depend heavily on the few-group constant preparation procedure. The method, based on using average in the fuel assembly fuel and coolant parameters is currently used for preparing macroscopic cross-sections. The question is what impact would considering the uneven distribution of those parameters, made on the few-group constant preparation stage exert on further analysis of the reactor facility behavior during steady-state and transients operation. The study carries out comparative analysis of the neutron-physical characteristics of the VVER-1000 core using the standard approach and using distributed in the fuel assembly fuel and coolant parameters while preparing few-group constants. It’s revealed that the fuel pellet and coolant radial temperature distributions affect the multiplication factor and temperature reactivity effect values.

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 NEUTRON NOISE PATTERNS FROM COUPLED FUEL-ASSEMBLY VIBRATIONS

2021 ◽  
Vol 247 ◽  
pp. 02015
Author(s):  
M. Viebach ◽  
C. Lange ◽  
M. Seidl ◽  
Y. Bilodid ◽  
A. Hurtado
Keyword(s):  
Neutron Flux ◽  
Fuel Assembly ◽  
Cross Sections ◽  
Three Dimensional ◽  
Time Dependent ◽  
Radial Dependence ◽  
Feedback Parameter ◽  
Fuel Assemblies ◽  
Neutron Noise ◽  
The Impact

The neutron flux fluctuation magnitude of KWU-built PWRs shows a hitherto unexplained correlation with the types of loaded fuel assemblies. Also, certain measured long-range neutron flux fluctuation patterns in neighboring core quadrants still lack a closed understanding of their origin. The explanation of these phenomena has recently revived a new interest in neutron noise research. The contribution at hand investigates the idea that a synchronized coolant-driven vibration of major parts of the fuel-assembly ensemble leads to these phenomena. Starting with an assumed mode of such collective vibration, the resulting effects on the time-dependent neutron-flux distribution are analyzed via a DYN3D simulation. A three-dimensional representation of the time-dependent bow of all fuel assemblies is taken into account as a nodal DYN3D feedback parameter by time-dependent variations of the fuel-assembly pitch. The impact of its variation on the cross sections is quantified using a cross-section library that is generated from the output of corresponding CASMO5 calculations. The DYN3D simulation qualitatively reproduces the measured neutron-flux fluctuation patterns. The magnitude of the fluctuations and its radial dependence are comparable to the measured details. The results imply that collective fuel-assembly vibrations are a promising candidate for being the key to understand long-known fluctuation patterns in KWU built PWRs. Further research should elaborate on possible excitation mechanisms of the assumed vibration modes.

scholarly journals Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review

Inventions ◽  
2020 ◽  
Vol 5 (3) ◽  
pp. 47
Author(s):  
Giovanni Giustini
Keyword(s):  
Heat Transfer ◽  
Liquid Water ◽  
Cfd Simulation ◽  
High Surface ◽  
Computational Modelling ◽  
Nucleate Boiling ◽  
Heat Fluxes ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Water Reactors

The boiling process is utterly fundamental to the design and safety of water-cooled fission reactors. Both boiling water reactors and pressurised water reactors use boiling under high-pressure subcooled liquid flow conditions to achieve high surface heat fluxes required for their operation. Liquid water is an excellent coolant, which is why water-cooled reactors can have such small sizes and high-power densities, yet also have relatively low component temperatures. Steam is in contrast a very poor coolant. A good understanding of how liquid water coolant turns into steam is correspondingly vital. This need is particularly pressing because heat transfer by water when it is only partially steam (‘nucleate boiling’ regime) is particularly effective, providing a great incentive to operate a plant in this regime. Computational modelling of boiling, using computational fluid dynamics (CFD) simulation at the ‘component scale’ typical of nuclear subchannel analysis and at the scale of the single bubbles, is a core activity of current nuclear thermal hydraulics research. This paper gives an overview of recent literature on computational modelling of boiling. The knowledge and capabilities embodied in the surveyed literature entail theoretical, experimental and modelling work, and enabled the scientific community to improve its current understanding of the fundamental heat transfer phenomena in boiling fluids and to develop more accurate tools for the prediction of two-phase cooling in nuclear systems. Data and insights gathered on the fundamental heat transfer processes associated with the behaviour of single bubbles enabled us to develop and apply more capable modelling tools for engineering simulation and to obtain reliable estimates of the heat transfer rates associated with the growth and departure of steam bubbles from heated surfaces. While results so far are promising, much work is still needed in terms of development of fundamental understanding of the physical processes and application of improved modelling capabilities to industrially relevant flows.

scholarly journals Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 531-536 ◽  
Author(s):  
Igor P. Królikowski ◽  
Jerzy Cetnar
Keyword(s):  
Cross Sections ◽  
Nuclear Reactions ◽  
Nuclear Reactors ◽  
Input Parameter ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled Analysis ◽  
Thermal Hydraulics ◽  
Power Profile ◽  
The Impact

Abstract Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

scholarly journals Analysing river network dynamics and active length - discharge relationship using water presence sensors

2021 ◽  
Author(s):  
Francesca Zanetti ◽  
Nicola Durighetto ◽  
Filippo Vingiani ◽  
Gianluca Botter
Keyword(s):  
Power Law ◽  
Cross Sections ◽  
Temporal Dynamics ◽  
River Network ◽  
Low Flow ◽  
Stream Network ◽  
Entire Study ◽  
Temporary Streams ◽  
Active Length ◽  
Spatio Temporal

Abstract. Despite the importance of temporary streams for the provision of key ecosystem services, their experimental monitoring remains challenging because of the practical difficulties in performing accurate high-frequency surveys of the flowing portion of river networks. In this study, about 30 electrical resistance (ER) sensors were deployed in a high relief 2.6 km2 catchment of the Italian Alps to monitor the spatio-temporal dynamics of the active river network during the fall of 2019. The set-up of the ER sensors was personalized to make them more flexible for the deployment in the field and more accurate under low flow conditions. Available ER data were analyzed, compared to field based estimates of the nodes' persistency and then used to generate a sequence of maps representing the active reaches of the stream network with a sub-daily temporal resolution. This allowed a proper estimate of the joint variations of active river network length (L) and catchment discharge (Q) during the entire study period. Our analysis revealed a high cross-correlation between the statistics of individual ER signals and the flow persistencies of the cross sections where the sensors were placed. The observed spatial and temporal dynamics of the actively flowing channels also revealed the diversity of the hydrological behaviour of distinct zones of the study catchment, which was attributed to differences in the catchment geology and stream-bed composition. The more pronounced responsiveness of the total active length to small precipitation events as compared to the catchment discharge led to important hysteresis in the L vs. Q relationship, thereby impairing the performances of a power-law model frequently used in the literature to relate these two quantities. Consequently, in our study site the adoption of a unique power-law L-Q relationship to infer flowing length variability from observed discharges would underestimate the actual variations of L by 40%. Our work emphasizes the potential of ER sensors for analysing spatio-temporal dynamics of active channels in temporary streams, discussing the major limitations of this type of technology emerging from the specific application presented herein.

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.

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