TRACE simulation of sodium boiling in pin bundle experiments under loss-of-flow conditions

Chenu, Aurélia; Mikityuk, Konstantin; Chawla, R.

doi:10.1016/j.nucengdes.2009.07.015

Cited by 24 publications

(2 citation statements)

References 17 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…But since sodium experiments are expensive and involve safety concerns, large volumes of numerical simulations were also reported on sodium boiling. The reported numerical simulations on sodium boiling are generally based on lumped approaches such as porous body [6] or subchannel analysis [7]. Recently, Vyskocil and Macek [8] performed 2D CFD simulation of R12 boiling in pipe flow using RPI model using FLUENT and NEPTUNE codes.…”

Section: Introductionmentioning

confidence: 99%

CFD Simulation of Sodium Boiling in Heated Pipe using RPI Model

Raj¹,

Velusamy²

2017

World Congress on Momentum, Heat and Mass Transfer

View full text Add to dashboard Cite

CFD prediction of sodium flow boiling in a 2-D asymmetric heated pipe is carried out using RPI wall heat flux partitioning model available in the commercial CFD code FLUENT. The two phase models are validated with published pressurized water and R12 flow boiling experimental data. The influence of radial mesh density on validation parameter is investigated. Void fraction, wall temperature and liquid temperature at different axial section are taken as the validation parameter for water experiment. Among the various parameters predicted, the void fraction of water at the outlet gave the largest deviation of 35% with experimental data. But its value is in close comparison near the location of boiling initiation. Further the void fraction calculated based on heat balance compares well with the CFD results. Radial distributions of void fraction, vapour velocity, bubble diameter at the end of heated section are the parameters compared with R12 experimental data. Except bubble diameter, all other parameters gave reasonably good comparison. But when invoking breakup and coalescence model the predicted value of bubble diameter value approaches experimental data. Finally, sodium boiling simulations were carried out with the chosen models. It was found that, unlike water and R12, in the case of sodium, stable boiling was not achieved and boiling is highly localized to the end of heated section. Periodic growth and re-condensation of sodium vapour is found. Further, temporal oscillation in the heat transfer coefficient up to 8% was observed in the boiling zone. During sodium bubble expansion the evaporative heat flux dominates and once the bubble detaches, the quenching heat flux dominates. Unlike pressurised water evaporative and quenching heat flux have almost equal participation in boiling region.

show abstract

Section: Introductionmentioning

confidence: 99%

CFD Simulation of Sodium Boiling in Heated Pipe using RPI Model

Raj¹,

Velusamy²

2017

World Congress on Momentum, Heat and Mass Transfer

View full text Add to dashboard Cite

show abstract

“…The number of such codes is large, even if attention is confined only to liquid-metal fast reactors. The best known foreign grid codes are DYANA-P [1] (India), NETFLOW++ [2] (Japan), MARS-LMR [3] (Korea) as well as codes with wider applications but constructed according to the grid principle and initially based on one-dimensional models (USA) [4,5], RELAP-5 (USA), ATHLET [6] (Germany), CATHARE (France) [7,8], and SAS4A/SASSYS-1 (USA) [9]. Many of these codes, with the exception of DYANA-P and SAS4A/SASSYS-1, were developed for reactors cooled by water and then modified and additionally developed so as to include reactors cooled by liquid metal.…”

mentioning

confidence: 99%

Russian Codes for Safety Analysis of Sodium-Cooled Fast Reactors

Rachkov¹,

Khomyakov²,

Shvetsov³

2014

At Energy

View full text Add to dashboard Cite

An approach to validating the safety of sodium-cooled fast reactors which was devised in Russia as these reactors were developed is presented. The collection of computational codes used for validating safety is described. Attention is focused on a group of codes which are intended for systematic analysis of all phases of a serious beyond design-basis accident resulting in the destruction of the core. The domestic approach to the analysis of the consequences of serious accidents is compared to the methodology developed abroad. Up to now, the code system under development has made it possible to perform the main tasks of design validation, but the chosen path of development, where a separate computational code was developed for each individual problem, is not exhausted. On the one hand, the complexity of the mathematical models drawn into the analysis has increased significantly, while on the other hand significant advances have been made in the development of computational technologies. The answer to the present situation is the tendency observed in the world nuclear community toward developing integrated codes and multipurpose computational systems. In Russia, this tendency is realized in the program for the development of a new generation of code systems within the framework of a special federal program.The approach to the safety analysis of sodium-cooled fast reactors was formed on the basis of domestic and foreign experience accumulated in the process of designing and operating these reactors. One of the main tools for solving this general problem is a set of computational codes used for validating the characteristics of fast reactors and analyzing their changes under different initial events, including accidents. As numerical modeling methods develop and the computational hardware improves, computer codes play an increasingly important role in the validation process. In practice, both the code developers and code users rejected, first and foremost, the list of applied problems whose solution required complete and comprehensive validation of the reactor parameters and safety. Such lists are, on the one hand, universal and applicable for definite types of reactors, for example, sodium-cooled fast reactors, while on the other hand the characteristics of each particular design must be taken into account in them. The lists are compiled by the joint efforts of design organizations and oversight agencies in application to each design being developed and include a list of disruptions of normal operation, design-basis and beyond design-basis accidents.The task of safety analysis is to show that the operational limits and safety criteria formulated in the regulatory documents are not violated in all initial failures and accidents included on the list.Since the lists of initial disruptions and accidents cover a large number of phenomena on different levels, a set of different codes differing by the spectrum of the phenomena being modeled, the degree of detail of the modeling and, in consequence, the computational accura...

show abstract