scholarly journals Multiphysical Simulations for the IAEA/ISCP Benchmark Model on the Contact of Pressure Tube and Calandria Tube in the Moderator System of CANDU-6 PHWR

2018 ◽  
Vol 2018 ◽  
pp. 1-8
Author(s):  
Hyoung Tae Kim ◽  
Se-Myong Chang ◽  
Young Woo Son ◽  
Taegee Min

The LOCA (loss of coolant accident) is a kind of severe accident in the operation of PHWR (pressurized heavy water reactor) as well as other nuclear facilities, and possible cause of LOCA can be counted on the ballooning of pressure tube (PT) contacted to the outer calandria tube (CT) in the moderator system of CANDU-6 reactors. In the paper, we simulated the 150-kW experimental facility proposed by IAEA/ISCP, modeling the transient creeping behavior of pressurized tube heated with thermal radiation between the gaps of two concentric pipes. The outer boundary is simplified with a switched model that depends on the local temperature. With a multiphysical model supported by a commercial code, COMSOL multiphysics, the unsteady phenomena are simulated with models concerning various kinds of mechanics such as thermodynamics, nonlinear structural dynamics, and two-phase boiling heat transfer models.

Author(s):  
Chang Hwan Park ◽  
Doo Yong Lee ◽  
Ik Jeong ◽  
Un Chul Lee ◽  
Kune Y. Suh ◽  
...  

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.


Author(s):  
H. G. Lele ◽  
A. Srivastava ◽  
B. Chatterjee ◽  
A. J. Gaikwad ◽  
Rajesh Kumar ◽  
...  

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.


2013 ◽  
Vol 135 (4) ◽  
Author(s):  
Ashwini K. Yadav ◽  
Ravi Kumar ◽  
Akhilesh Gupta ◽  
B. Chatterjee ◽  
P. Majumdar ◽  
...  

Some postulated events for pressurized heavy water reactor (PHWR) small break loss of coolant accident (SBLOCA) may lead to flow stratification in the reactor channels. Such stratified flow causes a circumferential temperature gradient in the fuel bundle as well as in the surrounding pressure tube (PT). The present investigation has been performed to study the thermomechanical behavior of a PT under an asymmetric heat-up condition arising from flow stratification in a 19 pin fuel element simulator. A series of experiments has been carried out at various stratification levels and PT internal pressures. The asymmetrical heat-up creates a temperature difference of 400 °C across the diameter of the PT. At high temperature the internal pressure causes ballooning of the PT. With the stratification, ballooning is found to get initiated at top hot side of PT and further propagates unevenly over its periphery. Axially ballooning is found to get initiated from center and then propagates toward both the ends of the PT. This results in an axial temperature gradient on the PT in addition of circumferential gradient. For a pressure higher than 4.0 MPa, the integrity of PT is found to be lost due to the combined effect of circumferential and axial temperature gradient generated under uneven strain distribution.


2011 ◽  
Vol 133 (1) ◽  
Author(s):  
P. Majumdar ◽  
B. Chatterjee ◽  
G. Nandan ◽  
D. Mukhopadhyay ◽  
H. G. Lele

In Indian pressurized heavy water reactor (PHWR), loss of coolant accident with simultaneous failure of emergency core cooling system can lead to significant temperature rise in the pressure tube (PT) with the system internal pressure varying from 9 MPa to 0.1 MPa during the event. This high temperature can cause metallurgical and geometrical changes in the PT. PT would deform plastically due to internal pressure and fuel weight. A computer code “PTCREEP” based on physical models was developed to simulate the ballooning deformation expected during the channel heatup condition under internal pressure. This paper presents the assessment of the code PTCREEP against the set of experiments conducted with PT material used in Indian PHWRs.


Author(s):  
P. Saha ◽  
B. K. Rakshit ◽  
P. Mukhopadhyay

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.


Author(s):  
E. W. Coryell ◽  
E. A. Harvego ◽  
L. J. Siefken

The SCDAP-3D© computer code (Coryell 2001) has been developed at the Idaho National Engineering & Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D© relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of RELAP5-3D©. Enhancements to the severe accident models include the ability to simulate high burnup and alternative fuel, as well as modifications to support advanced reactor analyses, such as those described by the Department of Energy’s Generation IV (GenIV) initiative. Initial development of SCDAP-3D© is complete and two widely varying but successful applications of the code are summarized. The first application is to large break loss of coolant accident analysis performed for a reactor with alternative fuel, and the second is a calculation of International Standard Problem 45 (ISP-45) or the QUENCH 6 experiment.


2018 ◽  
Vol 2018 ◽  
pp. 1-24
Author(s):  
J. C. de la Rosa Blul ◽  
S. Brumm ◽  
F. Mascari ◽  
S. J. Lee ◽  
L. Carenini

A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.


2015 ◽  
Vol 2015 ◽  
pp. 1-9
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci

As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters.


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