In-vessel coolability and retention of a core melt. Volume 1

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In that case, the decay heat will be removed by the reactor cavity flooding, to prevent the escape of radioactive material from the reactor vessel. But there is controversy that currently proposed strategy without additional measures could provide sufficient heat removal in higher power reactors. China is now developing higher power passive PWR with an operating power of MW and MW, in order to obtain the independent intellectual property rights which is very important for nuclear exports.

Non-peer reviewed

In the design, the feasibility of applying the IVR-ERVC strategy to keep the integrity of the pressure vessel is one of the key problems. So it is necessary to investigate it carefully. The researchers believe that the thermal load to the lower head is maximized when the debris pool has reached a steady thermal state. Heat transfer correlations available for steady molten pool are provided to calculate the thermal energy on the lower head. Critical heat flux as function of position on the lower head is also obtained based on the steady state, which will not change with inlet water temperature, mass velocity or heat flux distribution.

Although at steady state the total thermal load to lower head is maximized, it is not guaranteed everywhere because the heat flux is not distributed uniformly. So in the early stage, the transient behavior of the molten corium may impose a greater threat to the integrity of the reactor pressure vessel. Considering that the boundary condition at the outer surface of RPV lower head varies in the transient melting process, the existing critical heat flux expression is not enough.

In order to evaluate the feasibility of IVR-ERVC strategy in a transient severe accident, we need to know the critical heat flux in different conditions. Experimental work is effective, but as we know, this kind of experimental work is very expensive and takes a long time. So it's useful to develop a theoretical model to predict CHF under this situation, in which the characteristics in ERVC condition such as inclined heating wall and non-uniform heat flux must be considered.

Among them, boundary layer separation model Tong, , bubble crowding model Weisman and Pei, , sublayer dryout model Lee and Mudawwar, and interfacial lift-off model Galloway and Mudawar, are receiving attention in the flow conditions. Guo et al. The model was verified in uniform heat flux condition, but it is unknown whether it is effective in non-uniform heat flux condition.

In this paper, a theoretical CHF model is developed suitable for the outer surface of the lower head, and the effect of various parameters on CHF are investigated. They thought that under low quality condition the flow region could be divided into bubbly layer and bulk flow layer, and the limited turbulent interchange between them leads to the onset of CHF.

1. Introduction

The max value of void fraction in bubbly layer was postulated to be 0. The red line showed the approximate location of heater surface. Bubble crowding and vapor blanketing appeared near the wall, and small bubbles were dispersed in the bulk flow region. From the above equation, we know that the turbulent interchange at the interface and the vapor qualities of the bubbly layer and the bulk flow are precondition to calculate the critical heat flux.


In-vessel coolability and retention of a core melt. Volume 2 - Digital Library

It is assumed that only those velocity fluctuations which are larger than the vapor generation velocity could penetrate to the interface, the turbulent interchange was determined by:. In the model of Weisman and Pei, vapor quality was calculated using homogeneous flow model. The vapor quality in bubbly layer is written as:. The slip ratio S is defined as the ratio ofvapor velocity to liquid velocity, and can be expressed as:. It is assumed to be one half of the velocity at the interface, which can be calculated by the Karman velocity profile. From Eq. Then the vapor quality of the bulk flow is given as:.

The overall power approach assumes that the critical power for non-uniform heated tube is the same as that for uniform heated tube at the same cross-sectional geometry and heated length at given inlet conditions. The local conditions approach assumes that the CHF is independent of upstream history.

In-vessel coolability and retention of a core melt. Volume 2

The F-factor approach derived by Tong assumes that there is a superheated liquid layer between the bubbly layer and the heated wall. The enthalpy of the liquid layer at CHF is the same under uniform and non-uniform power shape. In the subcooled region or at low qualities, the upstream memory effect is small, and the local heat flux determines the CHF.

At high qualities, however, the memory effect becomes strong, and the average heat flux determines the CHF. In the IVR-ERVC conditions, given the water at the upper tank being saturated at atmospheric pressure, water subcooling at the heater inlet is about 10 K. Even at the CHF point, subcooled boiling is the dominant regime. So the local heat flux determines the CHF here. The present method is to calculate CHF values at different angular degrees first. Then we get the ratio of CHF value to the local heat flux at different locations.

If we gradually increase the local heat flux, boiling crisis will take place at the point with minimal ratio value as shown in Fig. ULPU experiments Theofanous et al. The test facility was an effective full-scale simulation of the reactor axisymmetric geometry. Configuration I was focused on the bottom of the lower head in a saturated pool boiling condition. The configurations IV and V studied the effect of the streamlined flow path. The experimental facility was constructed as shown in Fig.

The height of the facility was about 6 m.

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The radius of the heater blocks was 1. The heater blocks were made of 7. Power shaping was used to simulate the axisymmetric geometry in the. The comparison of predictions with experimental data is as Fig. It can be seen that with the increase of orientation, the critical heat flux on the heated wall first increases then decreases, which can be explained by our developed model. Bubble rise velocity increases with the orientation, so the slip ratio and steam quality in the bubbly layer tend to become bigger. Meanwhile, the thickness of the bubbly layer also increases. These factors result in increase of CHF.

But at high orientation, the upstream heat length is longer, and the upstream overall power is bigger, which makes the vapor quality of the CHF point bigger and easier to reach boiling crisis. This results in decrease of CHF.

chapter and author info

Sensitivity and uncertainty calculations are being performed in order to demonstrate the reliability and consistency of the ASTEC calculations. It consists of a network with several local databases. The protection of confidential data is an important feature that is taken into account as the information security of the database. The public web site www. For the communication between all network members, the e-collaborative Internet Advanced Communication Tool is used.

About papers related to SARNET work in the last 5 years have been presented in conferences or published in scientific journals. The dissemination of information is also done through periodic newsletters or participation to public events.

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  • In-vessel coolability and retention of a core melt. Volume 2|INIS.
  • 1. Introduction.
  • They are becoming the major event in the world on this topic. The Education and Training programme is focusing on raising the competence level of the university students Master and PhD and researchers engaged in SA research. Towards this purpose, education courses are elaborated on the phenomenology of the SA various areas. The teaching is not a survey but an in-depth treatment in order to allow the students and researchers to understand the methodology in the topics further and use analysis computer codes, mainly ASTEC, more effectively for any type of NPP. The description of the scenarios with event trees and fault trees is performed, with indication of the probabilities of the various events occurring.

    Best-estimate analyses are provided with uncertainty analyses. Four one-week educational courses were organised during the last five years, gathering from 40 to persons: the latter was organised in the University of Pisa in January , with a special focus on Gen. Another training course will be proposed in the future for staff of plant operators or regulatory authorities, with emphasis on identifying what the SAM procedures are based on, and why they are effective.

    It covers historical aspects of water-cooled reactors safety principles and phenomena concerning in-vessel accident progression, early and late containment failure, fission product release and transport. It contains also a description of analysis tools or codes, of management and termination of SA, as well as environmental management.

    It gives elements also on Gen. III reactors. The final review was performed in , and the publication is planned in the second part of Finally, a programme enables university students and researchers to go into different laboratories for education and training in the SA area. One of the high-priority issues concerns the core and debris coolability and thermal-hydraulics within particulate debris during core reflooding.

    PSA results do not give a unanimous answer for the ranking of the issue.