Several decades have been spent on the study of flow instabilities 
                  in boiling two-phase natural circulation systems. It is felt to have a review 
                  and summarize the state-of-the-art research carried out in this area, which 
                  would be quite useful to the design and safety of current and future light water 
                  reactors with natural circulation core cooling. With that purpose, a review of flow 
                  instabilities in boiling natural circulation systems has been carried out. An attempt 
                  has been made to classify the instabilities occurring in natural circulation systems 
                  similar to that in forced convection boiling systems. The mechanism of instabilities 
                  occurring in two-phase natural circulation systems have been explained based on 
                  these classifications. The characteristics of different instabilities as well as the effects 
                  of different operating and geometric parameters on them have been reviewed.

A. K. Nayak

P. K. Vijayan

Flow Instabilities in Boiling Two-Phase Natural Circulation Systems: A Review

Dilip Saha

John Cleveland

Natural Circulation in Nuclear Reactor Systems

In natural circulation loops, the driving force is usually low as it depends on the riser height which is generally of the order of a few meters. The heat transport capability of natural circulation loops (NCLs) is directly proportional to the flow rate it can generate. With low driving force, the straightforward way to enhance the flow is to reduce the frictional losses. A simple way to do this is to increase the loop diameter which can be easily adopted in pressure tube designs such as the AHWR and the natural circulation boilers employed in fossil-fuelled power plants. Further, the loop diameter also plays an important role on the stability behavior. An extensive experimental and theoretical investigation of the effect of loop diameter on the steady state and stability behavior of single- and two-phase natural circulation loops have been carried out and the results of this study are presented in this paper.

D. Saha

M. R. Gartia

Effect of Loop Diameter on the Steady State and  Stability 
                        Behaviour of Single-Phase and Two-Phase  Natural Circulation Loops

This paper presents 
                  a model of a simplified boiling water reactor 
                  (SBWR) to analyze the steady-state and 
                  transient behavior. The SBWR model is based on 
                  approximations of lumped and distributed 
                  parameters to consider neutronics and natural 
                  circulation processes. The main components of 
                  the model are vessel dome, downcomer, lower 
                  plenum, core (channel and fuel), upper plenum, 
                  pressure, and level controls. Further 
                  consideration of the model is the natural 
                  circulation path in the internal circuit of the 
                  reactor, which governs the safety performance of 
                  the SBWR. To demonstrate the applicability of 
                  the model, the predictions were compared with 
                  plant data, manufacturer_s predictions, and 
                  RELAP5 under steady-state and transient 
                  conditions of a typical BWR. In steady-state 
                  conditions, the profiles of the main variables 
                  of the SBWR core such as superficial velocity, 
                  void fraction, temperatures, and convective heat 
                  transfer coefficient are presented and analyzed. 
                  The transient behavior of SBWR was analyzed 
                  during the closure of all main steam line 
                  isolation valves (MSIVs). Our results in this 
                  transient show that the cooling 
                  system due to natural circulation in the SBWR is 
                  around 70% of the rated core flow. According 
                  to the results shown here, one 
                  of the main conclusions of this work is that the 
                  simplified model could be very helpful in the 
                  licensing process.

Gilberto Espinosa-Paredes

Alejandro Nu&#241;ez-Carrera

SBWR Model for Steady-State 
and Transient Analysis

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II &amp; III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

Avinash J. Gaikwad

Sharad Bhartya

Kannan Iyer

Rajesh Kumar

A. D. Contractor

H. G. Lele

S. F. Vhora

A. K. Maurya

A. K. Ghosh

H. S. Kushwaha

Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

Natural circulation experiments were conducted at Oregon State University using the advanced plant experiment (APEX-CE) integral system test facility as configured to simulate a typical <svg style="vertical-align:-0.0pt;width:33.974998px;" id="M1" height="10.9125" version="1.1" viewBox="0 0 33.974998 10.9125" width="33.974998" xmlns:xlink="http://www.w3.org/1999/xlink" xmlns="http://www.w3.org/2000/svg">
	
		
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</svg> Combustion Engineering nuclear steam supply system. This paper describes the mechanisms by which natural circulation flow was interrupted under single-phase and two-phase natural circulation conditions in APEX-CE.

José N. Reyes

Flow Stagnation under Single and Two-Phase Natural Circulation Conditions in the APEX-CE Test Facility

Natural circulation characteristics at low pressure/low power have been studied by performing experimental investigations and numerical simulations. The PANDA large-scale facility was used to provide valuable, high quality data on natural circulation characteristics as a function of several parameters and for a wide range of operating conditions. The new experimental data allow for testing and improving the capabilities of the thermal-hydraulic computer codes to be used for treating natural circulation loops in a range with increased attention.  This paper presents a synthesis of a part of the results obtained within the EU-Project NACUSP “natural circulation and stability performance of boiling water reactors.” It does so by using the experimental results produced in PANDA and by showing some examples of numerical simulations performed with the thermal-hydraulic code ATHLET.

Domenico Paladino

Max Huggenberger

Frank Schäfer

Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations

The aim of this paper is to address fluid flow behavior of natural circulation in a 2D-annular loop filled with water. A two-dimensional, numerical analysis of natural convection in a 2D-annular closed-loop thermosyphon has been performed for various radius ratios from 1.2 to 2.0, the loop being heated at a constant flux over the bottom half and cooled at a constant temperature over the top half. It has been numerically shown that natural convection in a 2D-annular closed-loop thermosyphon is capable of showing pseudoconductive regime at pitchfork bifurcation, stationary convective regimes without and with recirculating regions occurring at the entrance of the exchangers, oscillatory convection at Hopf bifurcation and Lorenz-like chaotic flow. The complexity of the dynamic properties experimentally encountered in toroidal or rectangular loops is thus also found here.

Gilles Desrayaud

Alberto Fichera

Numerical Analysis of General Trends in Single-Phase Natural Circulation in a 2D-Annular Loop

The aim of this paper is to collect and resume the work done to build and develop, at the University of Pisa, an engineering tool related to the natural circulation. After a brief description of the different loop flow regimes in single phase and two phase, the derivation of a suitable tool to judge the NC performance in a generic system is presented. Finally, an extensive comparison among the NC performance of various nuclear power plants having different design is done to show a practical application of the NC flow map.

M. Cherubini

W. Giannotti

D. Araneo

F. D&#39;Auria

Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation

As the 2400&#x2009;MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR) reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named &#x201c;probabilistic engineering assessment&#x201d; (PEA), constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD) of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs) is based on uncertainties propagation into thermal-hydraulic (T-H) calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a &#x201c;passive&#x201d; mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary) for the core damage frequency (CDF), the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

C. Bassi

M. Marques

Reliability Assessment of 2400&#x2009;MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

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Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.

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Science and Technology of Nuclear Installations

Natural Circulation in Nuclear Reactor Systems

Natural Circulation in Nuclear Reactor Systems

Articles

Flow Instabilities in Boiling Two-Phase Natural Circulation Systems: A Review

Natural Circulation in Nuclear Reactor Systems

Effect of Loop Diameter on the Steady State and Stability Behaviour of Single-Phase and Two-Phase Natural Circulation Loops

SBWR Model for Steady-State and Transient Analysis

Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

Flow Stagnation under Single and Two-Phase Natural Circulation Conditions in the APEX-CE Test Facility

Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations

Numerical Analysis of General Trends in Single-Phase Natural Circulation in a 2D-Annular Loop

Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation

Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations