Development of Technique for Controlling Natural Circulation Flow by Using Helium Gas

2014 ◽  
Author(s):  
Hirofumi Hatori ◽  
Naoto Yanagawa ◽  
Tetsuaki Takeda ◽  
Shumpei Funatani
Keyword(s):  
Steady State ◽  
Pressure Vessel ◽  
Control Method ◽  
Natural Circulation ◽  
Vertical Channel ◽  
Vertical Tube ◽  
Reactor Pressure Vessel ◽  
Circulation Flow ◽  
Helium Gas ◽  
Reactor Pressure

The purpose of this study is to investigate a control method of natural circulation flow of air by injection of helium gas. A depressurization accident by the primary pipe rupture is one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the double coaxial duct connecting between a reactor core and an intermediate heat exchanger (IHX) module breaks, air is expected to enter the reactor pressure vessel from the breach and oxidize in-core graphite structures. Then, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In such condition, injection of helium gas into the channel by a passive method can prevent occurrence of the natural circulation flow of air in the reactor pressure vessel. Therefore, it is thought that oxidation of in-core graphite structures by air ingress can be prevented by establishing this method. The experiment has been carried out regarding the natural circulation flow using a circular tube consisting of a reverse U-shaped type. The vertical channel consists of one side heated tube and the other side cooled tube. The experimental procedure is as follows. Firstly, the apparatus is filled with air and one vertical tube is heated. Then, natural circulation of air will be produced in the channel. After the steady state is established, a small amount of helium gas is injected from the top of the channel. The velocity, mole fraction, temperature of gas, and temperature of the tube wall are measured during the experiment. The results were obtained as follows. When the temperature difference between the both vertical tubes was kept at about 60K, the velocity of the natural circulation flow of air was measured about 0.17m/s. During a steady state, a small amount of helium gas was injected into the channel. When the volume of injected helium gas is about 5.7% of the total volume of the channel, the velocity of the natural circulation flow of air became around zero. After 810 seconds elapsed, the natural circulation flow suddenly produced again. The natural circulation flow of air can be controlled by injecting of helium gas.

Study on Control Method of Natural Circulation by Injection of Helium Gas

2012 ◽  
Author(s):  
Masashi Nomura ◽  
Tetsuaki Takeda ◽  
Shumpei Funatani ◽  
Takuya Shimura
Keyword(s):  
Pressure Vessel ◽  
Control Method ◽  
Natural Circulation ◽  
Vertical Channel ◽  
Reactor Pressure Vessel ◽  
Experimental Results ◽  
Circulation Flow ◽  
Helium Gas ◽  
Air Ingress ◽  
Reactor Pressure

This study is to investigate a control method of natural circulation of air by injection of helium gas. A depressurization accident is one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomena during the depressurization accident of the VHTR, therefore, it is important to develop the method for prevention of air ingress during the accident. The experiment has been carried out regarding natural circulation using a circular tube consisting of the loop type or the reverse U-shaped type. The vertical channel consists of the one side heated tube and the other side cooled tube. The experimental results were obtained as follows. When the temperature difference between the vertical tubes was kept at 52K, the velocity of natural circulation flow became about 12cm/s. During this steady state, a small amount of helium injected to the channel. Then, the flow velocity of natural circulation suddenly decreased. The volume of injected helium is about 3% of the total volume of the channel. The velocity became around zero. After 1500 seconds elapsed, the natural circulation suddenly produced again. The experimental results show that the natural circulation flow of air can be controlled by the method of helium gas injection. This paper also discusses an overview of the method for the prevention of air ingress during the primary pipes rupture accident.

Experiment and Numerical Analysis of Control Method of Natural Circulation by Injection of Helium Gas

2013 ◽  
Author(s):  
Naoto Yanagawa ◽  
Masashi Nomura ◽  
Tetsuaki Takeda ◽  
Shumpei Funatani
Keyword(s):  
Numerical Analysis ◽  
Mole Fraction ◽  
Pressure Vessel ◽  
Control Method ◽  
Natural Circulation ◽  
Circulation Flow ◽  
Helium Gas ◽  
Air Ingress ◽  
The One ◽  
Reactor Pressure

This study is to investigate a control method of the natural circulation of the air by the injection of helium gas. A depressurization is the one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, the air is predicted to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of the air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomenon during the depressurization accident of the VHTR, it is important to develop the method for prevention of air ingress during the accident. In this study, the air ingress process is discussed by comparing the experimental and analytical results of the reverse U-shaped channel which has parallel channels. The experiment of the natural circulation using a circular tube consisted of the reverse U-shaped type has been carried out. The vertical channel is consisted of the one side heated and the other side cooled pipe. The experimental apparatus is filled with the air and one side vertical tube is heated. A very small amount of helium gas is injected from the top of the channel. The velocity and the mole fraction of each gas are also calculated by using heat and mass transfer numerical analysis of multi-component gas. The result shows that the numerical analysis is considered to be well simulated the experiment. The natural circulation of the air has very weak velocity after the injection of helium gas. About 780 seconds later, the natural circulation suddenly produces. The natural circulation flow of the air can be controlled by the method of helium gas injection. The mechanism of the phenomenon is found that mole fraction is changed by the molecular diffusion and the very weak circulation.

Experimental Investigation on Boiling Flow Characteristics Under Passive IVR-ERVC Conditions

2018 ◽  
Author(s):  
Fan Wang ◽  
Bo Kuang ◽  
Pengfei Liu ◽  
Longkun He
Keyword(s):  
Pressure Vessel ◽  
Natural Circulation ◽  
Flow Characteristics ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Test Facility ◽  
Circulation Flow ◽  
Boiling Flow ◽  
Reactor Pressure

In vessel retention (IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. During postulated severe accidents, the heat generated due to the molten debris relocation to the lower reactor pressure vessel head needs to be removed continuously to prevent vessel failure. Besides the local critical heat flux (CHF) of outer wall surface which is the first importance, a stable feature of natural circulation flow and an effective natural circulation capability within the external reactor vessel cooling (ERVC) channel tend to be rather crucial for the success of IVR. Under this circumstance, a full-height ERVC test infrastructure for large advanced pressurized water reactor (PWR) IVR strategy engineering validation, namely reactor pressure vessel external cooling II test facility (REPEC-II), has been designed and constructed in Shanghai Jiao Tong University (SJTU). And therefore, a brief introduction to the SJTU REPEC II facility as well as the experimental progress to date, is hereby given in the paper. During test campaign on the REPEC II facility, the one-dimensional natural circulation boiling flow characteristics during IVR-ERVC severe accident mitigation are investigated, with the experimental observation and measurement on natural circulation flow characteristics within the REPEC II test facility. Based on the abundant results acquired in the test campaign, it is attempted, in this paper, to summarize and evaluate the ERVC performances and trends under various practical engineered conditions. The main evaluation results includes: influence on ERVC flow characteristics of various non-uniform heat load distribution cooling limits, the observed sinusoidal oscillation is suggested to be flashing-induced density wave oscillations and the oscillation period correlated well with the passing time of single-phase liquid in the riser. It is expected that these conclusions may help designers to have a reliable estimate of the impact of some related engineered factors on real IVR-ERVC performance.

Nodalization of NEK in APROS and Steady State Simulation

2014 ◽  
Author(s):  
Tadeja Polach ◽  
Klemen Debelak ◽  
Ivica Bašić ◽  
Luka Štrubelj
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Building Blocks ◽  
Reactor Pressure Vessel ◽  
Primary Circuit ◽  
Physical Parameters ◽  
Detailed Model ◽  
Reactor Pressure

A model of the primary circuit and part of the secondary circuit of the Slovenian Krško NPP – NEK was built using APROS - Advanced PROcess Simulation environment. The data used to describe the properties of the system modelled in APROS, were the data describing Krško NPP and its operational properties after the uprating and the introduction of the 18-month cycle. Basis for data collections, nodalization, structure and simplifications was NEK RELAP5\MOD3.3 Engineering Handbook and the 23rd cycle. In order to build a model describing all the important parameters, the available elements in APROS environment were used as building blocks for each system. The goal was to create a detailed model nodalization, which would give accurate results and would run on reasonable processing power. Each submodel was checked to verify that the partial results are within the allowable limits and that the description of the physical parameters is consistent with the real components. The model includes reactor pressure vessel, reactor coolant pumps and primary piping, steam generator, part of main steam, part of feedwater, pressurizer and reactor core kinetics. The regulation of pressurizer level and pressure, steam generator level and control rod is also modelled. The model consists of more than 400 thermal hydraulic volumes. The aim of building this model was a through thermal hydraulic analysis of the PWR systems present in the NPP Krško. Several simulations of the steady states at different power levels were performed. The resulting data describing the flow rates in steam generator feedwater, reactor pressure vessel, including bypass flows, heat transfer in reactor core and steam generator, thermal losses to containment, liquid level in pressurizer and steam generator, pressure drops in primary circuit and other parameters were then compared to the results of different types of calculation and to the testing data obtained from Krško NPP. The next step was to identify variations in results and determine whether they are consequence of wrong parameters, measurement deviation or numerical error. In that manner the model was verified and validated (in the sense of comparison with available system surveillance plant test results) to ensure the correct setup, initial and boundary conditions were applied in order to get reliable steady state results.

Buoyancy Driven Coolant Mixing Studies of Natural Circulation Flows at the ROCOM Test Facility Using ANSYS CFX

2006 ◽  
Author(s):  
Thomas Ho¨hne ◽  
So¨ren Kliem ◽  
Ulrich Rohde ◽  
Frank-Peter Weiß
Keyword(s):  
Pressure Vessel ◽  
Natural Circulation ◽  
Reactor Pressure Vessel ◽  
Density Difference ◽  
Test Facility ◽  
Mean Flow ◽  
Pressurized Water ◽  
The Core ◽  
Ansys Cfx ◽  
Reactor Pressure

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Thermal hydraulics analyses showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show a stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

scholarly journals Process of Air Ingress during a Depressurization Accident of GTHTR300

2018 ◽  
Vol 2018 ◽  
pp. 1-15 ◽  
Author(s):  
Tomoya Shiga ◽  
Yudai Tanaka ◽  
Tetsuaki Takada
Keyword(s):  
High Temperature ◽  
Natural Convection ◽  
Pressure Vessel ◽  
Molecular Diffusion ◽  
Natural Circulation ◽  
Onset Time ◽  
Side Wall ◽  
Reactor Pressure Vessel ◽  
Experimental Apparatus ◽  
Reactor Pressure

A depressurization accident is the design-basis accidents of a gas turbine high temperature reactor, GTHTR300, which is JAEA’s design and one of the Very-High-Temperature Reactors (VHTR). When a primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. Therefore, it is important to know a mixing process of different kinds of gases in the stable and unstable density stratified fluid layer. In order to predict or analyze the air ingress phenomena during the depressurization accident, we have conducted an experiment to obtain the mixing process of two component gases and the characteristics of natural circulation. The experimental apparatus consists of a storage tank and a reverse U-shaped vertical rectangular passage. One side wall of the high temperature side vertical passage is heated and the other side wall is cooled. The other experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The outside of the double coaxial vessel is cooled and the inside is heated. The results obtained in this study are as follows. When the primary pipe is connected at the bottom of the reactor pressure vessel, onset time of natural circulation of air is affected by not only molecular diffusion but also localized natural convection. When the wall temperature difference is large, onset time of natural circulation of air is strongly affected by natural convection rather than molecular diffusion. When the primary pipe is connected at the side of the reactor pressure vessel, air will enter the bottom space in the reactor pressure vessel by counter-current flow at the coaxial double pipe break part immediately. Afterward, air will enter the reactor core by localized natural convection and molecular diffusion.

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.

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

2014 ◽  
Vol 10 (1) ◽  
pp. 123-127 ◽  
Author(s):  
Gyeong-Geun Lee ◽  
Yong-Bok Lee ◽  
Min-Chul Kim ◽  
Junhyun Kwon
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Reactor Pressure
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