scholarly journals INVESTIGASI KARAKTERISTIK TERMOHIDROLIKA TERAS REAKTOR DAYA KECIL DENGAN PENDINGINAN SIRKULASI ALAM MENGGUNAKAN RELAP5

2016 ◽  
Vol 18 (1) ◽  
pp. 1
Author(s):  
Susyadi Susyadi ◽  
Hendro Tjahjono ◽  
Sukmanto Dibyo ◽  
Jupiter Sitorus Pane
Keyword(s):  
Flow Rate ◽  
Cooling System ◽  
Natural Circulation ◽  
Safety Margin ◽  
Inherent Safety ◽  
Hydraulic Characteristics ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Core ◽  
Small Modular Reactor

ABSTRAK INVESTIGASI KARAKTERISTIK TERMOHIDROLIKA TERAS REAKTOR DAYA KECIL DENGAN PENDINGINAN SIRKULASI ALAM MENGGUNAKAN RELAP5. Reaktor modular daya-kecil (small modular reactor, SMR) memiliki prospek tinggi untuk dibangun di Indonesia. Keluaran dayanya yang relatif kecil dan disainnya yang kompak serta dapat dikonstruksi secara modular memberikan keunggulan fleksibilitas pembangunan yang lebih baik dibanding reaktor konvensional berdaya besar. Disain sistem reaktor kategori ini sangat bervariasi, salah satu diantaranya adalah jenis reaktor air tekan (pressurized water reactor, PWR) yang menerapkan sirkulasi alamiah pada sistem pendingin primernya. Selain itu reaktor ini juga memiliki teras (core) lebih pendek dibanding PWR konvensional. Dari kedua perbedaan tersebut maka terdapat kemungkinan perbedaan pola perpindahan panas yang dapat berimplikasi terhadap keselamatan secara keseluruhan. Oleh karena itu, pada penelitian ini dilakukan investigasi terhadap karakteristik termohidrolika teras reaktor tersebut khususnya karakteristik temperatur fluida dan bahan bakar serta laju alir fluidanya. Tujuannya adalah untuk mengetahui perbedaan marjin keselamatan temperatur teras reaktor bila dibanding dengan PWR konvensional. Investigasi dilakukan dengan menggunakan program RELAP5, dimana secara parsial teras reaktor dimodelkan menggunakan model-model generik yang ada pada program dan dilakukan beberapa perhitungan kondisi tunak. Hasil perhitungan menunjukkan bahwa saat beroperasi pada daya nominalnya, reaktor modular ini memiliki margin temperatur pendidihan sebesar 2K lebih baik dibanding reaktor konvensional. Selain itu, keunggulan marjin keselamatan reaktor modular daya-kecil ini juga ditunjukkan dari naiknya laju alir mengikuti kenaikan dayanya yang berarti memiliki sifat keselamatan yang melekat (inherent safety). Kata kunci: reaktor modular daya-kecil, PWR, sirkulasi alam, RELAP5, termohidrolika   ABSTRACT INVESTIGATION ON CORE THERMAL HYDRAULIC CHARACTERISTICS OF SMALL MODULAR REACTOR WITH NATURAL CIRCULATION COOLING USING RELAP5. Small modular reactor (SMR ) is very prospective to be deployed in Indonesia. Its low output power, compact design and capability to be constructed modularly provide better deployment flexibility compared to a large conventional reactor. There are various designs of SMRs, one of them implements natural circulation for its primary cooling system or in other words the reactor uses no primary pumps. Besides, the dimension of fuel element is shorter than the one used by large reactor. These two aspects may produce different heat transfer behavior, which could lead to a safety implication.  For that reason, this research investigates thermal hydraulic characteristics of the core of SMR with naturally circulating coolant, especially on the fuel and coolant temperatures and mass flow rate. The purpose is to identify the thermal safety margin difference of the reactor compared with conventional PWR.  The investigation was performed using RELAP5 in which the core was partially represented by means of generic models of the program and continued with steady state calculations. The result shows that during nominal power operation, the reactor has better of 2K  degree for boiling temperature margin than the large conventional PWR. In addition, the excellence of SMR safety margin was shown by the increase of primary coolant flow rate following the increase of power, which means that the reactor has a distinctive inherent safety. Keywords: small modular reaktor, PWR, natural circulation, RELAP5, thermal-hydraulic

PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale

2010 ◽  
Author(s):  
Klaus Umminger ◽  
Simon Philipp Schollenberger ◽  
Se´bastien Cornille ◽  
Claire Agnoux ◽  
Delphine Quintin ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Limited Range ◽  
Test Facility ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Physical Phenomena ◽  
Boron Dilution

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

scholarly journals Core Design of a Small Pressurized Water Reactor with AP1000 Fuel Assembly Using SRAC and COBRA-EN Codes

2020 ◽  
Vol 2020 ◽  
pp. 1-7
Author(s):  
Van Khanh Hoang
Keyword(s):  
Safety Margin ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Control Rod ◽  
The Core ◽  
Small Modular Reactor ◽  
Active Core ◽  
And Performance

This paper presents the core design and performance characteristics of a 300 MWt small modular reactor (SMR) with fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system with the JENDL-4.0 data library and the CORBRA-EN code. The calculated temperature coefficients including fuel temperature, coolant temperature, and isothermal temperature coefficient provide adequate negative reactivity feedbacks. The thermal-hydraulic analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and clad surface temperature. A safety analysis using the CORBRA-EN code shows that the core will remain covered during the entire transient procedure of the fast transient of remarkably increasing power that would be caused by the ejection of control rod. The analysis results indicate that the core with a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

Heterogeneous Inherent Boron Dilution Transient: A CFD Analysis of a Hot Boron Depleted Slug Interacting With Colder Borated Water in a PWR

2010 ◽  
Author(s):  
Philippe Freydier ◽  
Bruno Gaudron ◽  
Se´bastien Cornille ◽  
Virginie Lombard ◽  
Pauline Bertrand
Keyword(s):  
Boron Concentration ◽  
Cooling System ◽  
Initial Conditions ◽  
Natural Circulation ◽  
Saturation Temperature ◽  
Pressurized Water ◽  
The Core ◽  
Borated Water ◽  
Cfd Analyses ◽  
Boron Dilution

For the study of the Heterogeneous Inherent Boron Dilution transient in a Pressurized Water Reactor, a Small Break Loss Of Coolant Accident (SB-LOCA) is postulated. Natural Circulation (NC) may be interrupted and, under Reflux-Condenser (RC) conditions, the steam formed in the core condensates in the Steam Generator (SG) U-tubes: a boron-depleted slug may accumulate in the crossover leg and in the SG outlet chamber. If NC restarts as the Reactor Cooling System (RCS) is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. The mixing of the boron depleted slug with the borated water in the Cold Legs (CLs), downcomer and lower plenum after Restart of Natural Circulation (RNC) is quantified by means of Computational Fluid Dynamics (CFD) analyses. The CFD code STAR-CD is used to perform this analysis. Boundary conditions for this calculation — especially the boron-depleted slug size and the NC restart mass flow rate — are extrapolated from PKL experimental findings. The initial conditions are derived from an overall plant analysis performed with the CATHARE system code. Buoyancy effects, both in the cold leg and in the downcomer, are very significant phenomena for the evaluation of the slug transport and mixing: the hot (saturation temperature) boron-depleted water slug tends to accumulate in the upper parts of the cold legs and in the upper part of the downcomer (above the cold legs), before being pushed and dragged down. The boron concentration distribution at the core inlet during the transient, evaluated with STAR-CD, is compared with a critical value in order to check that boron concentration at the core inlet is always above the threshold necessary for the core to remain subcritical.

Feasibility Assessment of Air-Cooling System As an Ultimate Heat Sink of the ATOM System

2018 ◽  
Author(s):  
Doyoung Shin ◽  
Gwang Hyeok Seo ◽  
Min Wook Na ◽  
Sung Joong Kim ◽  
Yonghee Kim ◽  
...  
Keyword(s):  
Flow Rate ◽  
Heat Sink ◽  
Cooling System ◽  
Air Cooling ◽  
Primary Coolant ◽  
Final Heat ◽  
Feasibility Assessment ◽  
Environment Temperature ◽  
Air Cooling System ◽  
Atom System

Nowadays Small Modular Reactors (SMRs) have been receiving considerable attentions worldwide for potential advantages of an excellent flexibility for siting, low capital investment, and advanced safety. In Korea, a new research project has launched for the development of a conceptual design of a further advanced SMR which aims for a naturally-safe and autonomous operation, so called Autonomous Transportable On-demand reactor Module (ATOM). Major design objectives of the ATOM system are focused on the soluble boron-free (SBF) primary coolant system which enables the SMR to operate automatically in a load following mode. For the secondary system, the SCO2 power conversion cycle with air-cooling system as a final heat sink is being considered. The air-cooling system is expected to show flexible response even to extreme environmental conditions, such as a desert where utilization of cooling water is limited. The objective of this study is a feasibility assessment for applying the air-cooling system as a final heat sink of the ATOM by means of experimental work. As a 1st phase of the ATOM development, we first conducted the experiments using a typically considered primary coolant, water-steam, to verify that air flow has enough cooling capability to remove developed heat which the coolant carries. An Integrated Condensation Loop with Air-cooling System (ICLASS) experimental facility with three pressure boundaries (Steam, coolant, and air) was established. The cooling capability of the air-cooling system was evaluated by varying steam mass flow rate, coolant flow rate, and air environment temperature as experiment variables. Overall heat transfer rate by condensation was compared with numerical simulations of a 1D thermal-hydraulics analysis code, using the MARS model of the ICLASS facility.

Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

The simulation research on the natural circulation operation characteristic of FNPP in rolling and inclined condition

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 4-16
Author(s):  
R. Li ◽  
M. Peng ◽  
G. Xia ◽  
H. Li
Keyword(s):  
Nuclear Power ◽  
Natural Circulation ◽  
Rapid Development ◽  
Rolling Motion ◽  
Stable Operation ◽  
Circulation Flow ◽  
Operation Condition ◽  
Pressurized Water ◽  
The Core ◽  
Secondary Side

Abstract Recently, the FNPP (Floating Nuclear Power Plant) has got more and more attention and rapid development due to very wide prospect application on remote areas or islands. In general, the IPWR (Integral Pressurized Water Reactor) is adopted to meet the requirements of the limited space, the nuclear safety and the maneuverability in marine. The IPWR could depend on natural circulation operation to remove the residual heat of core under accident or low load operation condition. Because the driving head is low, the natural circulation flow is likely to be influenced by rolling and inclined condition. To clarify the natural circulation flow characteristics of the core in FNPP rolling motion and inclined condition, based on the modified THEATRe code by adding the ocean motion module and spatial coordinate convert module, the main thermal-hydraulic parameters variation in rolling and inclined condition were obtained. The effect of inclined angle, rolling amplitude and period on the natural circulation flow were discussed. The natural circulation flow in the core fluctuates periodically with rolling motion. And the inclination and rolling will also cause the degree of steam superheat of OTSG secondary side fluctuate, which could impact on the stable operation of secondary side system.

Influence of Steam Generator Tube Bundle Vibrations on the Operating Diagram of a Nuclear Plant During Stretch Out

2006 ◽  
Author(s):  
Andre´ Adobes ◽  
Joe¨l Pillet ◽  
Franck David ◽  
Michae¨l Gaudin
Keyword(s):  
Steam Generator ◽  
Cooling System ◽  
Tube Bundle ◽  
Thermal Power ◽  
Nuclear Plant ◽  
Coolant Temperature ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Normal Cycle ◽  
System Pressure

During the normal cycle of a pressurized water reactor, boron concentration is reduced in the core until fuel burns up. A stretch out of the normal cycle is however possible afterwards, provided primary coolant temperature is reduced. In those stretch out periods, nuclear operators want to keep constant thermal power exchanged in the steam generator, in order to preserve its performances. Under that constraint, the required reduction in primary coolant temperature involves both a decrease of secondary cooling system pressure and an increase of tube bundle vibrations. Since neither pressure nor vibrations should exceed some given thresholds in order to preserve component integrity, the reduction of primary coolant temperature has to be limited. Nuclear plant operators thereafter need an operating diagram, i.e. a diagram that provides minimum allowed primary coolant temperature versus power rate. In that context, we propose a method to derive such a diagram, by combining, on the one hand a code for simulating primary and secondary fluid flows in steam generators and, on the other hand, a software that allows one to predict fluid elastic tube bundle instabilities. That method allows one to take into account both tube fouling and plugging. It is now used by French utility “Electricite´ De France”, in order to check or supplement the analysis that are provided by steam generator manufacturers.

Two-Phase Natural Circulation Cooling Performance Assessment and Safety Analysis for APR1400 Core Catcher System

2014 ◽  
Author(s):  
Ki Won Song ◽  
Shripad T. Revankar ◽  
Hyun Sun Park ◽  
Bo Rhee ◽  
Kwang Soon Ha ◽  
...  
Keyword(s):  
Void Fraction ◽  
Flow Rate ◽  
Natural Circulation ◽  
Scaling Analysis ◽  
Circulation Flow ◽  
Two Phase ◽  
Cooling Performance ◽  
The Core ◽  
Core Catcher ◽  
Circulation Flow Rate

The two-phase natural circulation cooling performance of the APR1400 core catcher system is studied utilizing a drift flux flow model developed via scaling analysis and with an air-water experimental facility. Scaling analysis was carried out to identify key parameters, so that model facility could simulates two-phase natural circulation. In the experimental apparatus, instead of steam, air is injected into the top wall of the test channel to simulate bubble formation and void distribution due to boiling water in the core catcher channel. Measurement of void fraction critical to the heat transfer between the wall and coolant is carried out at certain key position using double-sensor conductivity probes. Results from the model provide expected natural circulation flow rate in the cooling channel of the core catcher system. The observed flow regimes and the data on void fraction are presented. For a given design of the down comer piping entrance condition bubble entrainment was observed that significantly reduced the natural circulation flow rate.

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