scholarly journals Preliminary Assessment of Decay Heat Removal Systems in the ESFR-SMART Design: The Role of Natural Air Convection Around Steam Generators Outer Shells in Accidental Conditions

2020 ◽  
Author(s):  
Jeremy Bittan ◽  
Clement Bore ◽  
Joel Guidez
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Steam Generators ◽  
Preliminary Assessment ◽  
Safety Criterion ◽  
Safety Level ◽  
Decay Heat ◽  
Air Convection ◽  
The Impact

Abstract In the frame of the ESFR-SMART European project, aiming at improving the safety level of the European Sodium cooled Fast Reactor (ESFR), this paper presents the preliminary assessment of decay heat removal systems in the ESFR-SMART design: the role of natural air convection around Steam Generators outer shells in accidental conditions. Both theoretical and CATHARE code (Thermal Hydraulics reference code) calculations are presented. The impact of an additional chimney at the top of each casing as well as running primary and secondary pumps on the heat removal capacity are equally evaluated. This paper shows that the evacuation of decay heat thanks to completely passive air natural circulation alone, in case of Fukushima like accident, should lead to temperatures of sodium in the reactor vessel temporarily exceeding the safety criterion of 650°C. The addition of chimneys increase the capacities but is not sufficient to evacuate the decay heat safely. If the primary and secondary side pumps are running, the safety criterion should be met.

Simulation Model of a Passive Decay Heat Removal System for Lead-Cooled Fast Reactors

2014 ◽  
Vol 137 (3) ◽  
Author(s):  
Lorenzo Damiani ◽  
Alessandro Pini Prato
Keyword(s):  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Coolant Temperature ◽  
Feed Water ◽  
Steam Generators ◽  
Fast Reactors ◽  
Primary Coolant ◽  
Reaction Heat ◽  
Decay Heat

The generation IV lead cooled fast reactors are of particular interest for the Italian research: several influential companies (Ansaldo Nucleare, ENEA) are involved in these important European R&D projects. At present, one significant European project in progress is lead cooled European advanced demonstrator reactor (LEADER) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, advanced lead fast reactor European demonstrator (ALFRED). The demonstrator has to include technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. The authors have addressed the issue of bayonet-tube steam generators proposing the external boiling bayonet steam generator (EBBSG) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. This is possible thanks to an external feed-water boiling, based on the known Loeffler scheme, coupled to the bayonet tube concept. In the present paper, the authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the proposed DHR system is investigated through a Matlab-Simulink model. The results are satisfactory since, according to the simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

scholarly journals Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

2008 ◽  
Vol 2008 ◽  
pp. 1-8 ◽  
Author(s):  
Giacomino Bandini ◽  
Paride Meloni ◽  
Massimiliano Polidori ◽  
Maddalena Casamirra ◽  
Francesco Castiglia ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Thermal Power ◽  
Heat Removal ◽  
Normal Operation ◽  
Steam Generators ◽  
Flow Paths ◽  
Decay Heat ◽  
Driven System ◽  
Accelerator Driven System ◽  
Relap5 Code

The development of a conceptual design of an industrial-scale transmutation facility (EFIT) of several 100 MW thermal power based on accelerator-driven system (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

scholarly journals 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

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

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 & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

scholarly journals CATHARE Assessment of Natural Circulation in the PKL Test Facility during Asymmetric Cooldown Transients

2012 ◽  
Vol 2012 ◽  
pp. 1-10 ◽  
Author(s):  
Anis Bousbia Salah ◽  
Jacques Vlassenbroeck
Keyword(s):  
Natural Circulation ◽  
Test Facility ◽  
Steam Generators ◽  
Calculation Results ◽  
The Impact

Results of the CATHARE code calculations related to asymmetric cooldown tests in the PKL facility are presented. The test under consideration is the G2.1 experiment performed within the OECD/NEA PKL-2 project. It consists of carrying out a cooldown under natural circulation conditions in presence of two (out of four) emptied Steam Generators (SGs) and isolated on their secondary sides. The main goal of the current study is to assess the impact of a chosen cooldown strategy upon the occurrence of a Natural Circulation Interruption (NCI) in the inactive (i.e., noncooling) loops. For this purpose, three G2.1 test runs were investigated. The calculation results emphasize, mainly, the effect of the cooldown strategy, and the conditions that could lead to the occurrence of the NCI phenomenon.

Performance of Sodium Natural Circulation Loop for Passvie Decay Heat Removal

2008 ◽  
Author(s):  
Hae-Yong Jeong ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Yong-Bum Lee ◽  
Dohee Hahn ◽  
...  
Keyword(s):  
Fast Reactor ◽  
System Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Circulation Loop ◽  
Ambient Atmosphere ◽  
Steady State Condition ◽  
Decay Heat ◽  
Natural Circulation Loop ◽  
Test Loop

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Analysis of a Convection Loop for GFR Post-LOCA Decay Heat Removal

2004 ◽  
Author(s):  
Wesley C. Williams ◽  
Pavel Hejzlar ◽  
Pradip Saha
Keyword(s):  
Heat Transfer ◽  
Steady State ◽  
Flow Regime ◽  
Natural Circulation ◽  
Heat Removal ◽  
Computer Code ◽  
Steady State Analysis ◽  
Decay Heat ◽  
Cooling Loops ◽  
State Analysis

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA GFR. The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO2 outperforms helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops.

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