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.

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.

Thermohydraulic characteristics of the condenser in the natural-circulation loop of an MTR modular heavy-water reactor

1992 ◽  
Vol 72 (2) ◽  
pp. 131-137
Author(s):  
B. F. Balunov ◽  
D. G. Govyadko ◽  
T. S. Zhivitskaya ◽  
V. I. Kiselev ◽  
B. R. Bergel'son ◽  
...  
Keyword(s):  
Heavy Water ◽  
Natural Circulation ◽  
Circulation Loop ◽  
Water Reactor ◽  
Natural Circulation Loop ◽  
Heavy Water Reactor

Effect of Ocean Conditions on Neutronic/Thermal-Hydraulic Coupling of IPWR

2018 ◽  
Author(s):  
Genglei Xia
Keyword(s):  
Natural Circulation ◽  
Oscillation Amplitude ◽  
Three Dimensional ◽  
Simulation Models ◽  
Reactor Power ◽  
Coolant Flow ◽  
Uneven Distribution ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Ocean Conditions

In this paper, a thermal-hydraulics analysis code under non-inertial system was developed by establishing dynamic simulation models of typical ocean conditions and modifying RELAP5 code. Based on the modified code, the effects of the inclination, fluctuation and rolling conditions on the natural circulation characteristics of an integrated PWR were studied. The dynamic input and output interface model of RELAP5 was further established, and RELAP5 was coupled with a two-group three-dimensional neutron kinetics code to realize the improvement of nuclear feedback module in thermal hydraulic code. And the distribution of core flow and power under different motions was analyzed. The results indicate that, the uneven distribution of coolant flow increases with the increasing inclination angle, this trend leads to an uneven distribution of primary coolant flow. The flow fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. The code developed in this paper has the functions of modelling ocean conditions and three-dimensional coupled neutronics/thermal-hydraulics, and can be used as a simulation tool for Floating PWR.

Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4

2010 ◽  
Author(s):  
Xuhua Ye ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Parameter Study ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Pressure Peak ◽  
Residual Heat

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.

Scaling of Natural Circulation Boiling Systems

2003 ◽  
Author(s):  
Kannan N. Iyer ◽  
Aboobacker Kadengal
Keyword(s):  
Steady State ◽  
Characteristic Equation ◽  
Heavy Water ◽  
Natural Circulation ◽  
Pressure Tube ◽  
Water Reactor ◽  
Type Reactor ◽  
Heavy Water Reactor ◽  
Tube Type ◽  
Indian Scenario

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.

Determination of radioactivity concentration in the primary coolant channel of a Pressurized Heavy Water Reactor

2010 ◽  
Vol 286 (1) ◽  
pp. 47-54 ◽  
Author(s):  
Sanhita Chaudhury ◽  
Chhavi Agarwal ◽  
A. Goswami ◽  
Amol Mhatre ◽  
Manohar Gathibandhe ◽  
...  
Keyword(s):  
Heavy Water ◽  
Radioactivity Concentration ◽  
Primary Coolant ◽  
Water Reactor ◽  
Heavy Water Reactor

On the MEGAPIE Target Thermal Hydraulics: A RELAP5 Analysis

2004 ◽  
Author(s):  
W. H. Leung
Keyword(s):  
Proton Beam ◽  
Natural Circulation ◽  
Heat Removal ◽  
Target Material ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Removal Capacity ◽  
Circulation Mode ◽  
Long Run ◽  
Cooling Loops

A new type of liquid-metal target is designed for the Spallation Neutron Source in PSI. LBE is selected to be the target material and the primary coolant as well. RELAP5/MOD 3.2 is used to analyze the system thermal hydraulics. The nominal conditions are chosen based on temperature constraints from the design assessments. The steady state results are in the proximity of the design specifications and the heat removal capacity is adequately deployed. The normal thermal hydraulic transients, namely the proton beam and beam interrupt, are studied. A basic PID (Proportional, Integral, Derivative) control is implemented in the RELAP5 for regulating the target temperature. It is found that the control chain works very well for the beam trip in limiting the temperature fluctuations. In a beam interrupt, the proton beam is completely turned off without recovering. The transition from full power to hot-standby is quite smooth, but it becomes oscillatory in the long run due to the timelags in the cooling loops’ responses. An off normal case of target main coolant trip has also been studied. Without the main pump, the target can still be operated in the natural circulation mode, and the control can cope with the normal beam transients and restarting the target from hot standby.

Sign in / Sign up

Export Citation Format

Share Document