Control Rod Channels Research Articles

Thermohydraulic characteristics of coolant gas flow through a control rod channel in a very high temperature gas cooled reactor

A calculation code was developed to evaluate the thermohydraulic performance of a coolant flow through a control rod channel in a very high temperature gas cooled reactor (VHTR) and a high temperature engineering test reactor (HTTR). A one-dimensional flow network model was employed in the present calculation code. The calculated results agreed well with the experimental ones on the flow rate distribution and the total pressure loss in an isothermal coolant flow. The thermohydraulic characteristics of the HTTR control rod channel were evaluated by the code under various conditions, including the normal operating conditions of a HTTR.

Hydraulics of coolant flow through control rod channel in very high temperature gas cooled reactor.

Flow rate distribution and total pressure loss of a coolant flow through a control rod channel in the Very High Temperature Gas Cooled Reactor (VHTR) were analytically and experimentally examined. Helium gas of atmospheric temperature was used in the experiment; and the total mass flow rates ranged 0.005∼0.05 kg/s and the gas pressures ranged 0.14∼0.42 MPa. Pressure losses and flow rates in the control rod channel were measured. An analysis was made by using a one-dimensional flow network model for the inner and outer channels and the gap. The analytical results agreed fairly well with the experimental results on the flow rate distribution and the total pressure loss in the control rod channel.

Hydraulics of Coolant Flow through Control Rod Channel in Very High Temperature Gas Cooled Reactor

Hydraulics of Coolant Flow through Control Rod Channel in Very High Temperature Gas Cooled Reactor

Discharge of14C by nuclear power stations with RBMK-1000 reactors

This paper presents the results of a project in the course of which the rate of formation and the intensity of the discharge of C-14 by n..mu..clear power stations with RBMK-1000 reactors were determined by a computational and experimental route. The systems of a reactor in which the formation of C-14 takes place are presented and include: the coolant, the cooling system of the control and safety rod channels, the nitrogen-helium purging circuit of the reactor space, the nitrogen purging circuit of the metal structure of the circumreactor space, the fuel elements, and graphite brickwork. The results of a numerical estimate of the normalized values of the rate of formation and intensity of the discharge of C-14 for the RBMK-1000 when operating at a nominal power are shown.

An effective homogenization method for heterogeneous assemblies

An effective homogenization method has been developed for heterogeneous assemblies such as fuel assemblies with and without control blades in BWR and control-rod channels in FBR. Effective homogenized cross sections are calculated so as to preserve the integrated reaction rates in a heterogeneous assembly in each group by iteratively changing the cross section used in homogeneous super-cell calculations in a model composed of the heterogeneous assembly and a fuel region. The method has been applied to the rod-worth calculation for pin rods in the fast critical assembly ZPPR-10 and to the power-density calculation of a test BWR core.

Unified Diffusion Coefficient for Analysis of Sodium-Void Worth in Fast Critical Assembly with Control-Rod Channels

Unified Diffusion Coefficient for Analysis of Sodium-Void Worth in Fast Critical Assembly with Control-Rod Channels

Unified diffusion coefficient for analysis of sodium-void worth in fast critical assembly with control-rod channels.

A unified diffusion coefficient which is applicable not only to fuel drawers but also to sodium followers or control-rod drawers has been derived by extending the Benoist method to analyze the sodium-void worths in fast critical assemblies including singularities such as sodium followers and control-rod drawers. The unified diffusion coefficient has been obtained by calculating average neutron leakage rates in perpendicular and parallel directions in a super-cell consisting of different drawers such as fuel drawers and sodium followers by using transport theory. Then the interference effect of the neutron streaming between different drawers on diffusion coefficients is taken into account. Diffusion coefficients of fuel drawers, sodium followers and control-rod drawers contained in the ZPPR-3 modified phase 3 core have been calculated on the basis of the present method, Benoist's method, Seki & Sasaki's method and the conventional method, and the results have been compared. The applicability of the unified diffusion coefficient to sodium-void worth calculations has been checked in one- and two-dimensional test geometries, and it has been found that the use of the unified diffusion coefficient leads to reasonable results without adopting the transport corrections in the models including sodium followers. Further the sodium-void worths in the ZPPR-3 modified phase 3 core have been analyzed using the unified diffusion coefficient.

Fission Power Distribution near a Cruciform Water Gap

An experiment was conducted in a critical assembly to determine water channel power peaking in a heterogeneous, highly enriched, water-moderated reactor subassembly. In addition to horizontal profiles of power density in a plane below a bank of control rods, data were obtained on the interacting effects of the control rod channel and bottom reflector power peaks. Experimental techniques are discussed. A computation for a comparable multiregion cell was made using a few-group diffusion theory digital computer code. Comparison of the results showed that variations in fast neutron spectrum and in manner of homogenization of materials within cell regions had no appreciable effects upon the power density peak. The choice of thermal neutron spectrum is important. Agreement between all experimental values and calculations with a Maxwellian spectrum was excellent. The comparison between experiment and calculations with a hardened thermal neutron spectrum was relatively poor. The results of this investigation indicate that great care must be exercised in interpreting experimental data on power distribution, and that two-dimensional diffusion theory calculations of power density are substantially verified. The use of U/ sup 235/ monitor foils is recommended in order to obtain a true fission power distribution. lt was found that variations in themore » water gap width of the order of lO% had negligible effect upon the horizontal peak-toaverage power ratio in the fuel. The use of metallic control rod extensions was found to decrease the horizontal peak-to-average ratio substantially, the magnitude of the decrease verifying analytical predictions. (auth)« less

Hydrodynamics of a Liquid Poison Scram System

A study is made of the transient motion of a suddenly released column of liquid poison, such as mercury or borated water, as it moves through the control tubes of a reactor. Two basic systems are described. The first system consists of a tank of liquid poison, situated above the reactor, connecting directly to control tubes extending down into the core. The second system is a U-tube arrangement in which the liquid poison enters through tubes that extend upward from the base of the core. By making the tank or standpipe diameter larger than the control tube diameter it is possible to obtain initial fluid column accelerations greater than gravitational acceleration. When solid control rods are used in reactors, it is necessary to employ auxiliary springs to give an initial acceleration greater than gravitational. High initial accelerations are inherent features of liquid poison scram systems. It is found that the initial acceleration is gravitational acceleration multiplied by the ratio of standpipe area to control tube area. An exploitable feature of the U-tube system is that the liquid column undergoes a natural rapid deceleration. It may therefore be possible to do away with damping devices that are often required at the base of a solid control rod channel. The equations of motions of the system are derived and integrated. The general solution in both cases gives the displacement time explicitly in terms of elliptic functions; in the U-tube system the unconfined fluid column is found to execute uniform elliptic oscillations.

International Dental Federation (F. D .I.)

In HTR-PM, gaps between graphite and carbon blocks are widely distributed in the reactor core. The cold helium flows may be bypassed and not completely heated, therefore, the accurate prediction of bypass flows is a key problem related to the reactor safety. The flow network method was employed to represent interconnected bypass flow paths and calculate the bypass flow rates. Bypass paths in irregular shapes were simulated by CFD to get local and friction resistances. Typical bypass flow paths, included flow through sealing ring to the pebble discharge tube, flow through small holes to the control rod channels and flow through vertical gaps to the hot plenum, were modeled to calculate corresponding bypass flow rates. Large gaps, where might induce high flow rates at two ends of the bypass paths, should receive more concerns.