scholarly journals Experimental evaluation of natural convection in the IPR-R1 Triga research reactor at 264 kW and 105 kW

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Amir Zacarias Mesquita
Keyword(s):  
Natural Convection ◽  
Technology Development ◽  
Cooling System ◽  
Natural Circulation ◽  
Fuel Temperature ◽  
Operational Parameters ◽  
Reactor Operating ◽  
Core Cooling ◽  
Reactor Pool ◽  
Forced Cooling

In order to study the safety aspects connected with the permanent increase of the maximum steady state power of the IPR-R1 Triga Reactor of the Nuclear Technology Development Center (CDTN), experimental measurements were done with the reactor operating at power levels of 265 kW and 105 kW, with the pool forced cooling system turned off. A number of parameters were measured in real-time such as fuel and water temperatures, radiation levels, reactivity, and influence of cooling system. Information on all aspects of reactor operation was displayed on the Data Acquisition System (DAS) shown the IPR-R1 online performance. The DAS was developed to monitor and record all operational parameters. Information displayed on the monitor was recorded on hard disk in a historical database. This paper summarizes the behavior of some operational parameters, and in particular, the evolution of the temperature in the fuel element centerline positioned in the core hottest location. The natural circulation test was performed to confirm the cooling capability of the natural convection in the IPR-R1 reactor. It was confirmed that the IPR-R1 has capability of long-term core cooling by natural circulation operating at 265 kW. The measured maximum fuel temperature of about 300 oC was lower than the operating limit of 550 oC. It has been proven that without cooling in the primary the gamma dose rate above reactor pool at high power levels was rather high.

scholarly journals PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE

2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono
Keyword(s):  
Cooling System ◽  
Fuel Temperature ◽  
Blow Down ◽  
Beam Tube ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Dry Out

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

Advanced Passive Thermal Management for LED Bulb Systems

2013 ◽  
Vol 2013 (DPC) ◽  
pp. 001277-001293
Author(s):  
James Petroski
Keyword(s):  
Natural Convection ◽  
Heat Sink ◽  
Energy Savings ◽  
Cooling System ◽  
Point Of View ◽  
Hazardous Substances ◽  
Size And Shape ◽  
Lighting Systems ◽  
Liquid Cooling System ◽  
Forced Cooling

The movement to LED lighting systems worldwide is accelerating quickly as energy savings and reduction of hazardous substances (RoHS) increase in importance. Furthering this trend are government regulations, rebate programs and declining prices. The market drive today is to replace light bulbs of common outputs (60W, 75W and 100W) without resorting to Compact Fluorescent (CFL) bulbs containing mercury while maintaining the standard industry bulb size and shape referred to as A19 for fixture retrofitting. This A19 size and shape restriction causes a small heat sink which is only capable of dissipating heat for 60W equivalent LED bulbs with natural convection. 75W and 100W equivalent bulbs require larger sizes, some method of forced cooling, or some unusual liquid cooling system; generally none of these approaches are desirable for light bulbs from a consumer point of view. Thus, there is interest in developing natural convection cooled A19 light bulb designs for LEDs that cool far more effectively than today's current designs. Current A19 size heat sink designs typically have thermal resistances of 5–7 °C/W. A more efficient method of cooling can be created using a chimney-based design to lower system thermal resistances below 4 °C/W while meeting all other requirements for bulb system design. Numerical studies and test data are in good agreement for various orientations including methods for keeping the chimney partially active in horizontal orientations. Such chimney-based designs are capable of cooling 75W and 100W equivalent LED light bulbs in the limited volume constraints of A19-size devices.

The Plant Feature and Performance of DMS (Double MS: Modular Simplified and Medium Small Reactor)

2008 ◽  
Author(s):  
Yukiko Kawabata ◽  
Masayoshi Matsuura ◽  
Shizuka Hirako ◽  
Takashi Hoshi
Keyword(s):  
Economic Efficiency ◽  
Cooling System ◽  
Natural Circulation ◽  
Circulation System ◽  
Surface Separation ◽  
Power Company ◽  
Loss Of Coolant Accident ◽  
Core Cover ◽  
And Performance ◽  
Core Cooling

The Japan Atomic Power Company has initiative in developing the DMS concept as a 400MWe-class light water reactor. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts and effective use of proven technology. The decrease in primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about two meters compared with 3.7 meters in the conventional BWR. The short active fuel length reduces the drop in core pressure, and overcomes the natural circulation system. And by using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to improve safety efficiency, developing an Emergency Core Cooling System (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following. 1: Eliminating high-pressure injection systems. 2: Adopting passive safety-related systems. 3: Optimizing distribution for the systems and power source for the ECCS. In this way the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of Loss of Coolant Accident (LOCA) analysis, core cover can be achieved by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.

PBMR Technology Development Projects at Stellenbosch University, South Africa

2008 ◽  
Author(s):  
R. T. Dobson
Keyword(s):  
Technology Development ◽  
Cooling System ◽  
Heat Engine ◽  
Organic Rankine Cycle ◽  
Development Program ◽  
Heat Pipes ◽  
Development Projects ◽  
Fuel Temperature ◽  
Two Phase ◽  
Research Activities

PBMR has initiated a research and development program wherein a network of expertise relating to PBMR-specific technology is to be established. As a result of this initiative four specific PBMR sponsored technology development projects have been initiated at Stellenbosch University. The work done and still to be done towards these projects will be presented. The first project relates to the characterization of the flow dynamics of particles (ions, atoms and clusters) in a high pressure and velocity (9 MPa and 120 m/s) stream of helium due to various body-force fields (magnetic, electric and centrifugal); the ultimate objective of this project is to develop a graphite dust and particle scrubbing system. The second project relates to an entirely passive reactor cooling system (RCCS) using thermosyphon-type heat pipes with no pumps and active controls. The third project relates to the fuel temperature measurement under normal and loss of coolant pressure conditions using a fibre-optic Bragg-grating method. A fourth project relates to energy efficiency improvement by the conversion of waste, decay, after and residual heat into electrical power. This project makes use of two-phase closed loop thermosyphon-type heat pipes to transport the heat to an external heat engine, such as free piston type Stirling engine or organic Rankine cycle system. The research activities needed to meet the objectives of the above projects will be presented and discussed in this paper.

Advanced Natural Convection Cooling Designs for LED Bulb Systems

2013 ◽  
Author(s):  
James Petroski
Keyword(s):  
Natural Convection ◽  
Heat Sink ◽  
Energy Savings ◽  
Cooling System ◽  
Point Of View ◽  
Light Bulb ◽  
Size And Shape ◽  
Lighting Systems ◽  
Liquid Cooling System ◽  
Forced Cooling

The movement to LED lighting systems worldwide is accelerating quickly as energy savings and reduction in hazardous materials increase in importance. Government regulations and rapidly lowering prices help to further this trend. Today’s strong drive is to replace light bulbs of common outputs (60W, 75W and 100W) without resorting to Compact Fluorescent (CFL) bulbs containing mercury while maintaining the standard industry bulb size and shape referred to as A19. For many bulb designs, this A19 size and shape restriction forces a small heat sink which is barely capable of dissipating heat for 60W equivalent LED bulbs with natural convection for today’s LED efficacies. 75W and 100W equivalent bulbs require larger sizes, some method of forced cooling, or some unusual liquid cooling system; generally none of these approaches are desirable for light bulbs from a consumer point of view. Thus, there is interest in developing natural convection cooled A19 light bulb designs for LEDs that cool far more effectively than today’s current designs. Current A19 size heat sink designs typically have thermal resistances of 5–7°C/W. This paper presents designs utilizing the effects of chimney cooling, well developed for other fields that reduce heat sink resistances by significant amounts while meeting all other requirements for bulb system design. Numerical studies and test data show performance of 3–4°C/W for various orientations including methods for keeping the chimney partially active in horizontal orientations. Significant parameters are also studied with effects upon performance. The simulations are in good agreement with the experimental data. Such chimney-based designs are shown to enable 75W and 100W equivalent LED light bulb designs critical for faster penetration of LED systems into general lighting applications.

Review and Status of the Gen-IV CANDU-SCWR Passive Moderator Core Cooling System

2008 ◽  
Author(s):  
H. F. Khartabil
Keyword(s):  
Heat Sink ◽  
Cooling System ◽  
Natural Circulation ◽  
Power Level ◽  
Heat Removal ◽  
Circulation System ◽  
The Past ◽  
Test Loop ◽  
Flow Oscillations ◽  
Core Cooling

Enhanced safety is an important priority in the development of Generation IV reactors, which can be accomplished through the use of improved passive heat removal systems. In CANDU® reactors, the separation between the low-pressure moderator and high-pressure coolant provides a unique passive heat sink for decay heat removal during accident scenarios. Methods for enhancing this passive heat sink for the GenIV CANDU-SCWR (supercritical water cooled reactor) have been under investigation for the past several years to support a “no core melt” reactor design concept (1, 2). Initially, to test feasibility, tests and analysis at AECL studied a full-height passive cooling loop and showed that a flashing-driven natural circulation system was possible in principle. However, flow oscillations were observed at low powers and could not be readily explained through analysis. While these oscillations were not considered to be detrimental to the heat removal capability, additional separate-effects experiments were conducted and causal mechanisms proposed for the oscillations. In addition, these separate effects tests suggested that oscillations could be avoided at any power level by suitable design. A new test loop with a more representative geometry was recently constructed and commissioned. Preliminary commissioning tests confirmed conclusions from the separate effects tests. In this paper, the new tests are compared to the past tests to explain the improved and more stable loop operation. This comparison suggests that a complete system coupled to an ultimate heat sink has the potential to improve loop operation even more by eliminating or significantly reducing flow oscillations at low powers. Plans for validating this conclusion will be provided.

Study of the Behaviors of an Open-Loop Heat-Removal System

2018 ◽  
Author(s):  
Xianmao Wang ◽  
Yonggang Shen ◽  
Jiang Yang ◽  
Yong Ouyang ◽  
Min Rui ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Calculation Model ◽  
Open Loop ◽  
Cooling Systems ◽  
Core Cooling ◽  
Passive Cooling Systems ◽  
Heat Removal System ◽  
Removal System

In the third generation of nuclear reactors, passive systems have been widely used such as passive core cooling system and passive containment cooling system, which usually relay on natural circulation induced by buoyancy force to remove heat. Most of these passive cooling systems are closed-loop natural circulations. In recent years, some open-loop heat-removal systems have also been put forward. Open-loop heat-removal systems have its own advantages such as its simplification and low costs. However, the thermal-hydraulic behaviors of open-loop heat-removal systems are still not totally clear and need further study. In this study, a simplified open-loop passive containment cooling system is studied. A calculation model is built based on RELAP SCDAPSIM code. The thermal-hydraulic behaviors of the system are studied. By changing some key parameters of the system, the influences of these parameters on the system are evaluated.

Experimental Study of the Effect of Hydrogen Inflow on Passive Core Cooling System With Natural Circulation Flow

2020 ◽  
Author(s):  
Yasunori Yamamoto ◽  
Masayoshi Mori ◽  
Kosuke Ono ◽  
Tetsuya Takada
Keyword(s):  
Steady State ◽  
Cooling System ◽  
Natural Circulation ◽  
Operating Conditions ◽  
Quasi Steady State ◽  
Circulation Flow ◽  
Term Operation ◽  
Accident Conditions ◽  
Core Cooling

Abstract Isolation Condenser (IC) is one of the passive core cooling systems with natural circulation flow, which is effective for safety measures against station black out. Once core uncover occurs, hydrogen generated in the core affects operating condition of ICs. To use ICs as an important safety measure not only for transient conditions but also for accident conditions, robustness of ICs against hydrogen inflow must be understood well. In this study, experiments with high pressure steam were conducted using experimental setup simulating IC, where helium was injected to simulate hydrogen effects. When the pressure in an accumulator increased high enough, natural circulation flow generated in the experimental loop. After the long-term operation, the pressure and the natural circulation flow rate achieved nearly constant. The pressure at quasi-steady state increased with increasing the helium injection amount. The pressure difference in a section including outlet side of a vertical pipe was slightly increased when helium was injected which may have indicated that the helium accumulated in the section and caused increment of the pressure loss. The startup pressure of the IC simulator also increased when helium was injected, where the driving force by the water head difference also decreased. Though long-term operations were performed after helium injection, the effect of injected helium on operating conditions of the IC remained for quasi-steady state conditions.

Investigation of Countermeasure Against Local Temperature Rise in Vessel Cooling System in Loss of Core Cooling Test Without Nuclear Heating

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Masato Ono ◽  
Atsushi Shimizu ◽  
Makoto Kondo ◽  
Yosuke Shimazaki ◽  
Masanori Shinohara ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Natural Circulation ◽  
Reactor Core ◽  
Severe Accident ◽  
Local Temperature ◽  
Water Cooling ◽  
The Core ◽  
Cooling Tube ◽  
Core Cooling

In the loss of core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of the reactor core is stopped without inserting control rods into the core and, furthermore, without cooling by the vessel cooling system (VCS) to verify safety evaluation codes to investigate the inherent safety of high-temperature gas-cooled reactor (HTGR) be secured by natural phenomena to make it possible to design a severe accident-free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel (RPV) by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water-cooling tube without thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from helium gas circulators (HGCs) by stopping water flow in the VCS, the local higher temperature position was specified in the uncovered water-cooling tube of the VCS, although the temperature was sufficiently lower than the maximum allowable working temperature, and the natural circulation of water had an insufficient cooling effect on the temperature of the water-cooling tube below 1°C. Then, a new safe and secured procedure for the loss of core cooling test was established, which will be carried out soon after the restart of HTTR.

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