Inter-subassembly heat transfer of sodium cooled fast reactors: Validation of the NETFLOW code

Mochizuki, Hiroyasu

doi:10.1016/j.nucengdes.2007.03.023

Cited by 26 publications

(2 citation statements)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…A part of the cold primary sodium exiting from DHX flows to the bottom of the subassemblies (SA) along core periphery and then flows radially inwards and upwards through inter-wrapper space between sub-assemblies. This flow is termed as 'inter-wrapper flow' (Nishimura et al 2000;Mochizuki 2007). A part of the total decay heat generated within the core is transferred to hot pool by sodium passing through the sub-assembly, while the rest of the decay heat is transferred to hot pool by the inter-wrapper flow.…”

Section: Inter-wrapper Flowmentioning

confidence: 99%

Overview of pool hydraulic design of Indian prototype fast breeder reactor

Velusamy

Chellapandi

Chetal

et al. 2010

Sadhana

View full text Add to dashboard Cite

Abstract. Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

show abstract

Section: Inter-wrapper Flowmentioning

confidence: 99%

Overview of pool hydraulic design of Indian prototype fast breeder reactor

Velusamy

Chellapandi

Chetal

et al. 2010

Sadhana

View full text Add to dashboard Cite

show abstract

“…The flow behavior in loop B exhibited a decreasing trend about 100 s later because of the degradation of natural circulation due to the loss of the heat removal function of the loop B IHX when the PHTS pump within the loop became stuck, as described below. The results of this flow rate analysis were cross-checked with the results derived using the NETFLOW code 14) whose validation was based on natural circulation flow analysis of the domestic experimental fast reactor JOYO. As shown in Fig.…”

Section: Loss Of 3-loop Shts Flow Plant Transient Response Analysis Rmentioning

confidence: 99%

Assessment of FBR MONJU Accident Management Reliability in Causing Reactor Trips

Sotsu

Kurisaka

2010

J. Nucl. Sci. Technol.

View full text Add to dashboard Cite

This paper describes a method and application of quantitatively evaluating Accident Management (AM) reliability upon a reactor trip failure for the MONJU fast breeder reactor using a PSA technique. The present method comprises an allowable time estimation that is based on plant transient response analysis using the Super-COPD code that was developed for use in best estimates of the plant dynamics of MONJU and in estimating failure probability of operator's actions in AMs within the allowable time based on time records obtained from simulator training. Application of this method to MONJU resulted in the estimation that the allowable time for an unprotected loss-of-heat sink event would be more than the longest observed time of 326 s. The corresponding operation failure probability would be less than 0.1 even after taking the uncertainty into consideration. Combining this with a level 1 PSA revealed that the total frequency of core damage accompanying a reactor trip failure at MONJU could be decreased by at least 50 percent due to the reactor trip AM.

show abstract

Applications of computational fluid dynamics in design of sodium-cooled fast reactors

Velusamy

2019

Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment

View full text Add to dashboard Cite

Inter-subassembly heat transfer of sodium cooled fast reactors: Validation of the NETFLOW code

Cited by 26 publications

References 2 publications

Overview of pool hydraulic design of Indian prototype fast breeder reactor

Overview of pool hydraulic design of Indian prototype fast breeder reactor

Assessment of FBR MONJU Accident Management Reliability in Causing Reactor Trips

Applications of computational fluid dynamics in design of sodium-cooled fast reactors

Contact Info

Product

Resources

About