Diameter Increases in Steam Generating Heavy Water Reactor Zircaloy Cans under Loss-of-Coolant Accident Conditions

2009 ◽  
pp. 24-24-12 ◽  
Author(s):  
KM Rose ◽  
ED Hindle
Keyword(s):  
Heavy Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Accident Conditions

Analysis of Pressure Suppression Pool Swell Using Microcomputers

1992 ◽  
Author(s):  
P. Saha ◽  
B. K. Rakshit ◽  
P. Mukhopadhyay
Keyword(s):  
Operating System ◽  
Heat Transport ◽  
Transport System ◽  
Heavy Water ◽  
Computer Software ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Suppression Pool

Abstract The present paper discusses the development of a computer software or code for a best-estimate analysis of Pressure Suppression Pool Hydrodynamics in a Pressurized Heavy Water Reactor (PHWR) system during a Loss-of-Coolant Accident (LOCA) at the primary heat transport system. The software has been developed on Microcomputers, namely, PC-XT or AT (286) under MS-DOS operating system.

Best-Estimate Analysis of Loss-of-Coolant Accident in Pressurized Heavy Water Reactors Using Microcomputers

1992 ◽  
Author(s):  
P. Saha ◽  
T. K. Das ◽  
A. Chanda ◽  
S. Ray
Keyword(s):  
Heavy Water ◽  
Computer Software ◽  
The Other ◽  
Water Reactor ◽  
Canadian French ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Water Reactors

Abstract The present paper discusses the development of a computer software or code for best-estimate analysis of Loss-of-Coolant Accident (LOCA) in Pressurized Heavy Water Reactor (PHWR) systems. The formulation is comparable to U.S., Canadian, French LOCA codes, namely, TRAC, RELAPS, ATHENA, CATHARE, etc. However, the present software has been developed on Microcomputers, namely, PC-XT and AT, whereas the other softwares were developed and are being used primarily on Mainframes such as CDC-7600, CYBER-176, CRAY, etc.

Simulation of Containment Thermal-Hydraulic Phenomena in the Marviken Blowdown Experiment With the CONTAIN and ASTEC CPA Codes

2009 ◽  
Author(s):  
Ivo Kljenak ◽  
Borut Mavko
Keyword(s):  
Heavy Water ◽  
Experimental Measurements ◽  
Experimental Facility ◽  
Water Reactor ◽  
Hydraulic Behavior ◽  
Heavy Water Reactor ◽  
Accident Conditions

The Boiling Heavy Water Reactor Blowdown 16 Experiment, which was performed in the Marviken experimental facility, was simulated with the CONTAIN and ASTEC CPA codes. The main purpose of the work was the assessment of the codes for simulating thermal-hydraulic phenomena in a BWR containment at accident conditions. Simulated pressures, atmosphere temperatures and wetwell pool characteristics are compared to experimental measurements. The results show that both codes satisfactorily reproduced the containment thermal-hydraulic behavior.

Transient Analysis for the Case of Changes in Coolant Inventory for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
H. G. Lele ◽  
A. Srivastava ◽  
B. Chatterjee ◽  
A. J. Gaikwad ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Heavy Water ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Two Phase ◽  
Decay Heat ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Core Cooling

Safety of nuclear reactor needs to be assessed against different categories of Postulated initiating events. Advanced Heavy Water Reactor is natural circulation light water cooled and heavy water moderated pressure tube type of reactor. Inventory of the system is important parameter in determination of flow characteristics of this natural circulation reactor. In view of this, various events that cause changes in PHT system inventory are analysed in this paper. One of the reason for decrease in coolant inventory is hypothetical Loss of coolant accident (LOCA) This event is of very low probability but important from designing engineered safeguard system of a reactor. Loss of coolant accident in a nuclear reactor can cause voiding of the reactor core due to expulsion of primary coolant from break. In such, a situation the reactor core experiences very low heat removal rate from the nuclear fuel though the decay heat generation continues even after tripping of the reactor. Heat generation in the reactor core is due to various sources such as decay heat, stored heat etc, can lead to heating of fuel elements. However, Emergency core cooling systems of the reactor are actuated and prevent undesirable temperature rise. These events are called design basis events and focus is on adequacy of Emergency Core Cooling System (ECCS) and fuel integrity. The scenarios, phenomena encountered and consequences depend upon size and location of break, system characteristics, and actuation and capability of different protection and engineered safeguard systems of the reactor system. Moreover, this reactor has several passive features to ensure safety of this reactor. which are considered in analyzing these events. Events under category of decrease in coolant inventory includes loss of coolant accidents due to break at different locations of different sizes. Various locations considered in this paper are steam line, inlet header, inlet feeder, ECCS header, downcomer, pressure tube, Isolation condenser inlet header, instrument line break at inlet header and steam drum. The paper also considers scenario emerging due to malfunctions like relief valve stuck open. Causes for events under category of increase in coolant inventory are Increase in Drum level controller set point, Inadvertent valving in of Accumulators and Inadvertent valving in of Gravity driven water pool (GDWP). Last two events are not analysed as they are not possible. The analysis for the above events is complex due to various complex and wide ranges of phenomena involved during different pies under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, two-phase natural circulation under depleted inventory conditions. Coupled neutronics and thermal hydraulics behaviour, Phenomena under LOCA, phenomena during ECCS injection, direct injection into fuel rod, advanced accumulator injection., vapour pull through and coupled controller and thermal hydraulics. Modelling of these phenomena for each event is discussed in this paper. In this paper summary of analyses for representive event is presented.

Ballooning of Pressure Tube Under LOCA in an Indian Pressurised Heavy Water Reactor

2010 ◽  
Author(s):  
Ravi Kumar ◽  
Gopal Nandan ◽  
P. K. Sahoo ◽  
B. Chatterjee ◽  
D. Mukhopadhyay ◽  
...  
Keyword(s):  
Heavy Water ◽  
Temperature Rise ◽  
Test Section ◽  
High Heat ◽  
Pressure Tube ◽  
Experimental Simulation ◽  
Electrical Heating ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Heavy Water Reactor

A study has been carried out by experimental simulation of the loss of coolant accident (LOCA) in an Indian pressurized heavy water reactor (IPHWR). The experiment has been carried out taking a completely voided fuel channel of Indian PHWR at 40 bar inside pressure as test-section. In order to simulate the rate of heat generation during LOCA, the pressure tube (PT) was electrically heated with a 12VDC/3500A rectifier. Initially the set-up was maintained at 300 °C temperature by resistance heating of PT. After attaining nearly steady state a step input of 21 kW electrical heating was given to the test-section which resulted in the temperature rise of PT with a gross rate of 2.8 °C/s. The ballooning deformation of test-section tube i.e. PT initiated at 575 °C temperature. With the progress of ballooning the rate of temperature rise was reduced due to high heat transmission to CT and subsequently to water in the tank surrounding CT. The pressure tube (PT) and calendria tube (CT) contact established at the average PT temperature of 680 °C. The contact was also confirmed from the average temperature profile of CT.

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