Experimental Investigations on the Isolation Condenser Performance in Integral Test Loop

2008 ◽  
Author(s):  
Manish Sharma ◽  
D. S. Pilkhwal ◽  
P. K. Vijayan ◽  
D. Saha
Keyword(s):  
Heavy Water ◽  
Cooling System ◽  
Water System ◽  
Heat Removal ◽  
Experimental Investigations ◽  
Feed Water ◽  
Integral Test ◽  
Water Reactor ◽  
Test Loop ◽  
Heavy Water Reactor

The proposed Advanced Heavy Water Reactor (AHWR) is a light water cooled and heavy water moderated pressure tube type boiling water reactor based on natural circulation. AHWR adopts several passive concepts with a view to simplify the design and to enhance safety and public acceptability. One such feature is passive decay heat removal using isolation condenser (IC) system during a station blackout. A scaled Integral Test Loop (ITL) was set up in BARC to simulate the overall system behavior studies for Advanced Heavy Water Reactor (AHWR). This facility simulates the Main Heat Transport System (MHTS), Emergency Core Cooling System (ECCS) and Isolation Condenser system (ICS) system, Feed Water System (FWS) and the associated controls. Power to volume scaling philosophy has been adopted for the design of the ITL systems. To evaluate the performance of IC, experiments have been carried out in ITL. The test results have been simulated using RELAP5/ MOD3.2. This paper deals with the experiments conducted, nodalization scheme adopted for ITL in RELAP5/MOD 3.2 simulation, transient predictions made and the results obtained in detail.

Thermal Hydraulic Analysis Due to the Changes in Heat Removal for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
B. Chatterjee ◽  
A. Srivastava ◽  
D. Mukhopadhyay ◽  
P. Majumdar ◽  
H. G. Lele ◽  
...  
Keyword(s):  
Heavy Water ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Power ◽  
Circulation System ◽  
Feed Water ◽  
Set Point ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
Pressure Controller

Advanced Heavy Water Reactor is natural circulation light water cooled and heavy water moderated pressure tube type of reactor. Changes in heat removal by primary heat transport system of a reactor have significant impact on various important system parameters like pressures, qualities, reactor power and flows. Increase in heat removal leads to Cooldown of the system subsequently reducing pressure, void increase and changes in power and flows of the system. Decrease in heat removal leads to warm-up of the system subsequently raising pressure, void collapse, and changes in power and flows of the system. The behaviour is complex as system under consideration is natural circulation system. Causes for events under category of increase in heat removal are mainly malfunctioning of feed water heaters, Isolation Condensers (IC) inlet valves and controllers. These events lead to cooldown of system and addition of positive reactivity addition due to void collapse. Various events considered are Feed Water System malfunctions that result in decrease in feed water temperature, inadvertent opening of IC valve, Failure of PHT Pressure Control System and Decrease in pressure controller set point to 67 bars. Causes for events under category of decrease in heat removal are mainly malfunctioning of controllers, feedwater valves and operating events like turbine trip. Functioning of passive cooling system and different valves play important role for these events. These events lead to increase in system pressure. Various events considered are Loss of normal feed water flow (multiple trains), Turbine trip without bypass without IC, Turbine trip without bypass with IC, Turbine trip with bypass without IC, Increase in PHT pressure controller set point, Decrease in level controller set point, Turbine Trip with setback, Decrease in steam flow and Class IV power failure. Changes in the system voids and pressures as a result of change in the heat removal leads to complex reactivity feedback due to coolant temperatures, void fraction and fuel temperatures. These changes in the reactor power together with void distribution change affect two-phase natural circulation flow. This paper brings out these aspects. It discusses descretisation of the system and brings out various design aspects. In this paper summary of analysis for each event is presented, various modeling complexities are brought out, evaluation of acceptance criteria is made and design implications of each event is discussed.

Experimental Demonstration of Safety During Extended Station Blackout in an Integral Test Loop of a Natural Circulation Boiling Water Reactor

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
A. K. Nayak ◽  
Mukesh Kumar ◽  
A. K. Vishnoi ◽  
Vikas Jain ◽  
D. K. Chandraker
Keyword(s):  
Safety Concern ◽  
Stress Test ◽  
Heat Removal ◽  
Test Facility ◽  
Integral Test ◽  
Water Reactor ◽  
Decay Heat ◽  
Safety Margins ◽  
Test Loop ◽  
Station Blackout

Decay heat removal for prolonged period of station blackout (SBO) is a safety concern of the nuclear reactors. Aftermath of Fukushima, safety evaluation (performance under severe conditions: stress test) of the reactors was carried out worldwide. It includes establishment of grace period of the reactors. Similar exercises for advanced heavy water reactor (AHWR) were also performed and the design of AHWR was established for its robustness against such events. Decay heat removal during extended SBO is such a condition to be qualified. In this regard, experiments in the integral test loop (ITL), a full scale test facility of AHWR, were conducted for continuous 7 days of extended SBO. Experiment was started with 6.8 MPa as the initial reactor pressure and decay heat removal was demonstrated for 7 days of SBO by passive means. It is observed that the pressure falls down to 1 MPa in 3 h. The design of AHWR was evaluated from safety critical aspects during such an event experimentally. During this event, the clad surface temperature was found to be well within safe limits of operations. As a result of this experiment, it can be concluded that the design of AHWR is capable to remove decay heat for 7 days of SBO with sufficient safety margins.

Advanced heavy water reactor control with the aid of adaptive second-order sliding mode controller

2019 ◽  
Vol 37 (4) ◽  
pp. 1237-1259 ◽  
Author(s):  
Rupam Gupta Roy ◽  
Dibyendu Ghoshal
Keyword(s):  
Heavy Water ◽  
Sliding Mode ◽  
Second Order ◽  
The State ◽  
Feed Water ◽  
Water Recovery ◽  
Content Type ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
Second Order Sliding Mode

Purpose Advanced heavy water reactor (AHWR) is a pressure tube type of heavy water reactor. It eliminates high-pressure heavy water coolant resulting in a reduction of heavy water leakage losses and eliminating heavy water recovery system. It recovers the heat generated in the moderator for feed water heating. However, it requires a satisfactory technological response to develop an effective controller that attains the challenges of the very high-level safety system. Hence, they require application-specific improvement for better controlling performance. Design/methodology/approach The purpose of this study intends to propose a system for controlling state vectors v1 and v2and in AHWR using Grey Wolf second-order sliding mode control (GW-SoSMC) technique. The main aim of the paper is to minimize the errors between the predicted and desired azimuthal angles of the system. With this proposed method, it is possible to mitigate both the chattering phenomenon and controlling performance of AHWR system. It implements a SoSMC controller based on GWO algorithm for the purpose of controlling the state vectors in the AHWR system. It aims to accomplish a controller for improving the performance of the AHWR system. Findings Through the performance analysis, the efficiency of the proposed GW-SoSMC technique was verified by comparing it with various conventional algorithms, such as GW-SMC, FF-SoSMC, ABC-SoSMC, GS-SoSMC and GA-SoSMC. From the analysis, it was obtained that the implemented GW-SoSMC technique was 65.3 per cent superior to GW-SMC, 65.32 per cent superior to both FF-SoSMC and 65 per cent superior to ABC-SoSMC, 65.8 per cent superior to the GS-SoSMC and 58 per cent superior to the GA-SoSMC methods. Thus, the effectiveness of the proposed method in controlling the state vectors in AHWR was obtained. Originality/value This paper presents a technique for controlling the state vectors in the AHWR system using GWO algorithm. This is the first work that uses GWO-based optimization for controlling state vectors in the AHWR system.

Analysis for the Case of Channel Flow and Reactivity Changes for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
A. Srivastava ◽  
P. Majumdar ◽  
D. Mukhopadhyay ◽  
H. G. Lele ◽  
S. K. Gupta
Keyword(s):  
Heavy Water ◽  
Natural Circulation ◽  
Heat Removal ◽  
Coolant Flow ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Control Rod ◽  
Water Reactor ◽  
Wide Range ◽  
Heavy Water Reactor

The proposed Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. Decrease in coolant flow or control rod malfunction can lead to undesirable rise in clad surface temperature depending upon severity and characteristics and response of the reactor and associated systems. In this paper safety assessment of the AHWR is made due to above events of different severity. Cause for events under category of decrease in coolant flow is mainly channel blockage of different severity at different locations. There is no other reason as it is natural circulation based reactor. Effect of flow decrease can be different in different channels and at different axial locations. In this paper channel blockages of different sizes are analysed at core inlet and using slave channel approach. Changes in reactivities can occur due to inadvertent withdrawal of one or more control rods from reactor core. In this analysis one control rod assembly is assumed to be removed from core. The event is simulated by addition of 5 mk reactivity in 120 seconds depending on the speed of withdrawal of assembly. The analysis for the above events are complex due to various complex and wide range of phenomena involved during different PIEs under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, coupled neutronics and thermal hydraulics behaviour, and coupled controller and thermal hydraulics. In this paper summary of analysis for each event is presented. In this paper, various modeling complexities are brought out; evaluation of acceptance criteria is made and design implications of each event are discussed.

Effect of Inlet Configuration on Moderator Velocity and Temperature Distribution Inside the Calandria of a Heavy Water Reactor

2008 ◽  
Author(s):  
Abhijeet Mohan Vaidya ◽  
Naresh Kumar Maheshwari ◽  
Pallippattu Krishnan Vijayan ◽  
Dilip Saha ◽  
Ratan Kumar Sinha
Keyword(s):  
Temperature Distribution ◽  
Impact Velocity ◽  
Heavy Water ◽  
Computational Study ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
The Impact

Computational study of the moderator flow in calandria vessel of a heavy water reactor is carried out for three different inlet nozzle configurations. For the computations, PHOENICS CFD code is used. The flow and temperature distribution for all the configurations are determined. The impact of moderator inlet jets on adjacent calandria tubes is studied. Based on these studies, it is found that the inlet nozzles can be designed in such a way that it can keep the impact velocity on calandria tubes within limit while keeping maximum moderator temperature well below its boiling limit.

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