Investigations to Arrive at a Stable Start-Up Procedure for AHWR

Author(s):  
R. K. Bagul ◽  
D. S. Pilkhwal ◽  
P. K. Vijayan ◽  
D. Saha

Natural circulation is being adopted as a mode of core heat removal in several nuclear reactors that are under development. This is due to the passive nature of natural circulation that enhances the system safety and reliability. However, major concern in the design of natural circulation based reactor systems is to avoid the flow instabilities that may occur under certain operating conditions, i.e. unstable operational regime. Therefore various reactor operational transients such as start-up, power raising, setback and also the steady state operating points must fall within the stable regime. The choice of operating procedures needs to be made judiciously and which also needs to be validated and supported by experiments. Advanced Heavy Water Reactor (AHWR) being developed in India is a pressure tube type natural circulation boiling water-cooled reactor, wherein major part of the power is generated by thorium. Experiments and analytical studies have been performed to arrive at a rational start-up procedure for AHWR. Experimental results obtained in simple rectangular natural circulation loops as well as in a scaled down facility have revealed the importance of external pressurization to avoid the flashing and Type-I instabilities that occur at low pressure during start-up.

2018 ◽  
Vol 68 (1) ◽  
pp. 1-10
Author(s):  
František Dzianik ◽  
Štefan Gužela ◽  
Eva Puškášová

Abstract The paper deals with the process properties in terms of the heat transfer, i.e. the thermal performance of the thermal-process units within a helium loop intended for the testing of the decay heat removal (DHR) from the model of the gas-cooled fast reactor (GFR). The system is characterised by a natural circulation of helium, as a coolant, and assume the steady operating conditions of the circulation. The helium loop consists of four main components: the model of the gas-cooled fast reactor, the model of the heat exchanger for the decay heat removal, hot piping branch and cold piping branch. Using the thermal calculations, the thermal performance of the heat exchanger model and the thermal performance of the gas-cooled fast reactor model are determined. The calculations have been done for several defined operating conditions which correspond to the different helium flow rates within the system.


Author(s):  
Kannan N. Iyer ◽  
Aboobacker Kadengal

This paper lays out the procedure for arriving at the dimensions of a model facility to simulate a pressure tube type reactor. The Advanced Heavy Water Reactor, whose design is being evolved in the Indian scenario, is used as a basis for the evolution of the model facility. The non-dimensional groups that need to be preserved are identified and the design is evolved by satisfying these non-dimensional groups. The inevitable distortions that get introduced are discussed and a suitable compensation procedure evolved. Finally, the evolved model is shown to satisfy both steady state and characteristic equation similarity.


Author(s):  
Frances Viereckl ◽  
R. Manthey ◽  
C. Schuster ◽  
A. Hurtado

Passive safety systems represent one field of research concerning the safety-related enhancement of nuclear power plants. Passive safety systems can ensure the safe removal of decay heat without an input of electrical or mechanical energy for commissioning or operation. The heat removal chain is guaranteed on the basis of the physical principles condensation, heat conduction, boiling and natural circulation. The thermal hydraulic processes in passive safety systems disagree with the plant-specific thermal hydraulics because of different operating conditions. Since the established system codes are validated for the plant-specific conditions, the operational behavior of passive safety systems is currently not sufficiently predictable. On this account, the German Federal Ministry of Education and Research initiated the joint project PANAS to investigate the decay heat removal by passive safety systems on the basis of experimental analyses, modelling and validation. Object is the heat removal chain in advanced boiling water reactors consisting of emergency condensers (EC; heat transfer from reactor core to core flooding pools) and containment cooling condensers (CCC; heat transfer from the containment to the shielding/storage pool). At Technische Universität Dresden, the test facility GENEVA was constructed for the experimental investigation of the operational behavior of the CCC. GENEVA models the CCC concerning the original thermal hydraulic conditions of the heat source and heat sink as well as the tube geometry for the heat transfer. In this way, the comparability of the thermal hydraulic phenomena is given. Previous experiments focused on the stability analysis of the natural circulation in the test facility. The focus of PANAS is on the condensation process of saturated steam at the outside of the slightly inclined tubes and the convection respectively boiling of both a stable and an unstable two-phase flow inside these tubes. For a detailed analysis, condensation rates at the outside as well as the flow structure inside have to be investigated experimentally. Therefore, the instrumentation in the heat transfer section of GENEVA is considerably enhanced. This enhancement comprises an optical measuring system for the film thickness or droplet size of the condensate, a tipping scale for the condensate mass flow, void probes for the steam void fraction and more than 100 thermocouples outside and inside the tubes for temperature profiles in axial, radial and azimuthal direction. By reference to these parameters, it is possible to examine the thermal hydraulic models for the heat transfer. The paper outlines the available models in system codes regarding condensation and boiling concerning the operating conditions of the CCC. Since dropwise condensation could be observed in previous experiments and the condensation models in system codes focus on film condensation, the review is extended beyond native models. A sensitivity analysis of the reviewed models regarding condensation shows huge differences concerning the value of the heat transfer coefficient. Furthermore, the courses of the condensation models present different dependencies regarding the heat transfer coefficient and the wall temperature. Due to this, the necessity of the experimental investigation and later the revision of the condensation models in system codes is confirmed. The comparison of the reviewed models with first experimental results outlines the tendency for the numerical description of the condensation process. Based on the investigation and validation of models concerning the heat transfer processes in the CCC, the operational behavior will be accurately predictable by established system codes, which enhances the safety investigation and the licensing. Although the conception of this investigation is founded on the CCC, the adapted models will be able to characterize the heat transfer processes boiling and condensation for saturation conditions at a relatively low pressure (maximum 4 bar) and for natural convection in general.


Author(s):  
B. Chatterjee ◽  
A. Srivastava ◽  
D. Mukhopadhyay ◽  
P. Majumdar ◽  
H. G. Lele ◽  
...  

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.


2017 ◽  
Vol 67 (1) ◽  
pp. 29-36
Author(s):  
František Dzianik ◽  
Štefan Gužela

Abstract The paper deals with the hydrodynamic properties, i.e. the consumption of mechanical energy expressed by pressure drops within a helium loop intended for the testing of decay heat removal (DHR) from the model of a gas-cooled fast reactor (GFR). The system is characterised by the natural circulation of helium, as a coolant, and assume steady operating conditions of circulation. The helium loop consists of four main components: model of gas-cooled fast reactor, model of the heat exchanger for decay heat removal, hot piping branch and cold piping branch. Using the process hydrodynamic calculations, the pressure drops of circulating helium within the main components of the helium loop were determined. The calculations have been done for several defined operating conditions which correspond to the different helium flow rates within the system.


2019 ◽  
Vol 69 (1) ◽  
pp. 39-50
Author(s):  
František Dzianik ◽  
Štefan Gužela ◽  
Eva Puškášová

AbstractThe paper presents a comparison of the process properties of two types of the heat exchangers designed for the heat removal from a high temperature helium cooling loop with steady natural circulation of helium. The first considered heat exchanger is a shell and tube heat exchanger with U-tubes and the other one is a helical coil heat exchanger. Using the thermal and hydrodynamic process calculations, the thermal performance of the two alternative heat exchangers are determined, as well as the pressure drops of flowing fluids in their workspaces. The calculations have been done for several defined operating conditions of two considered types of heat exchangers. The operating conditions of heat exchangers correspond to the certain helium flow rates.


Author(s):  
A. Srivastava ◽  
P. Majumdar ◽  
D. Mukhopadhyay ◽  
H. G. Lele ◽  
S. K. Gupta

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.


2020 ◽  
pp. 10-21
Author(s):  
V. G. Babashov ◽  
◽  
N. M. Varrik ◽  

The emergence of new types of space and aviation technology necessitates the development of new types of thermal protection systems capable of operating at high temperature and long operating times. There are several types of thermal protection systems for different operating conditions: active thermal protection systems using forced supply of coolant to the protected surface, passive thermal protection systems using materials with low thermal conductivity without additional heat removal, high-temperature systems, which are simultaneously elements of the bearing structure and provide thermal protection, ablation materials. Heat protection systems in the form of rigid tiles and flexible panels, felt and mats are most common kind of heat protecting systems. This article examines the trends of development of flexible reusable heat protection systems intended for passive protection of aircraft structural structures from overheating.


Kerntechnik ◽  
2011 ◽  
Vol 76 (4) ◽  
pp. 237-243
Author(s):  
R. Kumar ◽  
A. J. Gaikwad ◽  
A. D. Contractor ◽  
A. Srivastava ◽  
H. G. Lele ◽  
...  

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