Loss of Core Cooling Test With One Cooling Line Inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Yusuke Fujiwara ◽  
Takahiro Nemoto ◽  
Daisuke Tochio ◽  
Masanori Shinohara ◽  
Masato Ono ◽  
...  
Keyword(s):  
Control System ◽  
High Temperature ◽  
Cooling System ◽  
Thermal Power ◽  
Reactor Power ◽  
Coolant Temperature ◽  
Water Cooling ◽  
Test Reactor ◽  
Core Cooling ◽  
Cooling Function

In the high-temperature engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system (PCS) were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the temperature changes of permanent reflector block, RPV, and the biological shielding concrete were quite slow during the test. The temperature decrease of RPV was several degrees during the test. The numerical result showed a good agreement with the test result of temperature rise of biological shielding concrete around 1 °C by the numerical method that uses a calibrated thermal resistance by using the measured temperatures of RPV and the air outside of biological shielding concrete. The temperature increase of water cooling tube panel of VCS was calculated to be about 15 °C which is sufficiently small in the view point of property protection. It was confirmed that the sufficient cooling capacity of VCS can be maintained even in case that one of two water cooling lines of VCS loses its function.

Loss of Core Cooling Test Without One Cooling Line in Vessel Cooling System (VCS) of High Temperature Engineering Test Reactor (HTTR)

2016 ◽  
Author(s):  
Yusuke Fujiwara ◽  
Takahiro Nemoto ◽  
Daisuke Tochio ◽  
Masanori Shinohara ◽  
Masato Ono ◽  
...  
Keyword(s):  
Control System ◽  
Pressure Vessel ◽  
Cooling System ◽  
Reactor Power ◽  
Reactor Pressure Vessel ◽  
Measured Temperature ◽  
Helium Gas ◽  
Test Reactor ◽  
Core Cooling ◽  
Reactor Pressure

In the High Temperature engineering Test Reactor (HTTR), the Vessel Cooling System (VCS) which is arranged around the reactor pressure vessel, removes residual heat and decay heat from the reactor core passively when the forced core cooling is lost. The test was carried out at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system was isolated, and three helium gas circulators in the primary cooling system were stopped to simulate the loss of forced cooling flow in the core. Moreover, one cooling line of the VCS was stopped to simulate the partial loss of cooling function from the surface of the reactor pressure vessel. The test results showed that the reactor power immediately decreased to almost zero and was stable as soon as the helium gas circulators were stopped. The power decrease is caused by negative feedback effect of reactivity. On the other hand, temperature changes of core internal structures, the reactor pressure vessel and the biological shielding concrete were slowly during the test. The measured temperature of the reactor pressure vessel decreased for several degrees during the test. The measured temperature increase of biological shielding made of concrete was small within 1 °C. The numerical analysis showed that the temperature increase of VCS cooling tube was about 15°C which is sufficiently small, which did not significantly affect the temperature of biological shielding. As the results, it was confirmed that the cooling ability of VCS can be kept sufficiently even in case that one of two water cooling lines of VCS is lost.

Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive

2013 ◽  
Author(s):  
Shoji Takada ◽  
Shunki Yanagi ◽  
Kazuhiko Iigaki ◽  
Masanori Shinohara ◽  
Daisuke Tochio ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Thermal Radiation ◽  
Cooling System ◽  
Evaluation Model ◽  
Reactor Core ◽  
Reactor Power ◽  
Coolant Temperature ◽  
Water Cooling ◽  
Reactor Outlet

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.

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.

Design of cooling system for inspection manipulator and analysis based on experiment

2016 ◽  
Vol 43 (2) ◽  
pp. 231-240 ◽  
Author(s):  
Tan Chen ◽  
Wei-jun Zhang ◽  
Jian-jun Yuan ◽  
Liang Du ◽  
Ze-yu Zhou
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Insulation Material ◽  
Water Cooling ◽  
Content Type ◽  
Cooling Method ◽  
Temperature Experiment ◽  
Gas Cooling ◽  
Loop Feedback ◽  
Electrical Components

Purpose – This paper aims to present a different cooling method (water cooling) to protect all the mechanical/electrical components for Tokamak in-vessel inspection manipulator. The method is demonstrated effective through high temperature experiment, which provides an economical and robust approach for manipulators to work normally under high temperature. Design/methodology/approach – The design of cooling system uses spiral copper tube structure, which is versatile for all types of key components of manipulator, including motors, encoders, drives and vision systems. Besides, temperature sensors are set at different positions of the manipulator to display temperature data to construct a close-loop feedback control system with cooling components. Findings – The cooling system for the whole inspection manipulator working under high temperature is effective. Using insulation material such as rubber foam as component coating can significantly reduce the environmental heat transferred to cooling system. Originality/value – Compared with nitrogen gas cooling applied in robotic protection design, although it is of less interest in prior research, water cooling method proves to be effective and economical through our high temperature experiment. This paper also presents an energetic analysis method to probe into the global process of water cooling and to evaluate the cooling system.

Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs

2012 ◽  
Author(s):  
Kuniyoshi Takamatsu ◽  
Kazuhiro Sawa
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Control ◽  
Reactor Power ◽  
Reactor Kinetics ◽  
The Core ◽  
Power Control System ◽  
Stop Test ◽  
High Temperature Reactor ◽  
Forced Cooling

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results. We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.

scholarly journals THE RESEARCH OF CIRCULATION WATER SUPPLY SYSTEM OF POWER UNIT OF THERMAL POWER PLANT WITH HELLER COOLING TOWER

2020 ◽  
Vol 2020 (2) ◽  
pp. 1-9
Author(s):  
Mykola Bosak ◽  
◽  
Oleksandr Hvozdetskyi ◽  
Bohdan Pitsyshyn ◽  
Serhii Vdovychuk ◽  
...  
Keyword(s):  
Power Plant ◽  
Water Supply ◽  
Power Unit ◽  
Cooling Tower ◽  
Cooling System ◽  
Thermal Power ◽  
Water Cooling ◽  
Pressure Losses ◽  
Circulating Water ◽  
Water Cooling System

Analytical hydraulic researches of the circulating water cooling system of the power unit of a thermal power plant with Heller cooling tower have been performed. Analytical studies were performed on the basis of experimental data obtained during the start-up tests of the circulating water cooling system of the “Hrazdan-5” power unit with a capacity of 300 MW. Studies of the circulating water cooling system were carried out at an electric power of the power unit of 200 - 299 MW, with a thermal load of 320 - 396 Gcal/hr. By circulating pumps (CP), water mixed with condensate is fed to the cooling tower, from where it is returned through the turbine for spraying by nozzles in the turbine steam condenser. An attempt to increase the water supply to the condenser by increasing the size of the nozzles did not give the expected results. The amount of the water supply to the circulating pumping station depends on the pressure loss in the circulating water cooling system. The highest pressure losses are in hydro turbines (HT), which are part of the circulating pumping station. Therefore, by adjusting the load of the hydro turbine, with a decrease in water pressure losses, you can increase the water supply by circulating pumps to the condenser. Experimental data and theoretical dependences were used to calculate the changed hydraulic characteristics of the circulating water cooling system. As a result of reducing the pressure losses in the section of the hydro turbine from 1.04 to 0.15 kgf/cm2, the dictating point for the pressure of circulating pumping station will be the turbine steam condenser. The thermal power plant cooling tower is designed to service two power units. Activation of the peak cooler sectors of the cooling tower gives a reduction of the cooled water temperature by 2-4 °С only with the spraying system.

Effect of Coolant Water Temperature of ECCS on Failure Probability of RPV

2019 ◽  
Author(s):  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Kai Lu ◽  
Tadashi Watanabe ◽  
Yinsheng Li
Keyword(s):  
Water Temperature ◽  
Thermal Shock ◽  
Failure Probability ◽  
Structural Integrity ◽  
Cooling System ◽  
Pressure Vessels ◽  
Coolant Temperature ◽  
Pressurized Water ◽  
Coolant Water ◽  
Core Cooling

Abstract For reactor pressure vessels (RPVs) of pressurized water reactors, temperature of the coolant water in the emergency core cooling system (ECCS) may influence the structural integrity of the RPV during pressurized thermal shock (PTS) events. By focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. The analysis results show that the failure probability of RPV decreased dramatically when the coolant temperature in accumulator as well as in the high- and low-pressure injection systems (HPI/LPI) was increased, although the increase in coolant temperature in the HPI/LPI only did not lead to a decrease in the failure probability.

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.

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a Fluoride-Salt-Cooled High-Temperature Test Reactor

2015 ◽  
Vol 1 (1) ◽  
Author(s):  
Yao Xiao ◽  
Lin-Wen Hu ◽  
Suizheng Qiu ◽  
Dalin Zhang ◽  
Su Guanghui ◽  
...  
Keyword(s):  
High Temperature ◽  
Computer Code ◽  
The United States ◽  
Coolant Temperature ◽  
Transfer Model ◽  
Fluoride Salt ◽  
Pebble Bed ◽  
Reference Case ◽  
Test Reactor ◽  
Temperature Test

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses high-temperature tristructural isotropic (TRISO) fuel with a low-pressure liquid salt coolant. Design of the fluoride-salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble-bed core design with a coolant temperature of 600–700°C is being planned for construction by the Chinese Academy of Sciences’ (CAS) Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal-hydraulic transient analyses of an FHTR using SINAP’s pebble-bed core design as a reference case. A point kinetic model is implemented using computer code by coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating several transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that SINAP’s pebble-bed core is a very safe reactor design.

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