Control Rod Design Research Articles

A Hafnium Material Control Rod Design and Manufacture

Because the metallic hafnium has good combination properties, that makes it the preferred material for control rods of nuclear reactors. To prevent the irradiation deformation and the mechanical property decrease of hafnium plates, and to increase the service life of the control rods, a new type of square-tube control rod is developed. The elemental composition, mechanical property, the corrosion resistance and the physical characteristics of hafnium are researched. Several tests are carried out after the product manufactured, which prove that the design and manufacture of the control rod meet the requirement of the technical conditions.

Structure and properties of hafnium after a 9-year operation in the RBT-6 research reactor

Hafnium control rods in the form of plates 6 mm thick were examined after a 9-year operation in the RBT-6 research reactor in water coolant at 68–86 °С to a maximal fluence 7.5 × 10 25 m −2 ( Е < 0.625 eV) and 2.4 × 10 25 m −2 ( E > 0.8 MeV). The Hf crystalline lattice volume increased by 0.4% and its density decreased. The oxide film thickness did not exceed 1 μm. Linear-type a-dislocations with a concentration of (5 ± 2) × 10 13 m −2 were found in the structure as well as fine precipitates, where iron and chromium were detected, in addition to hafnium. Hf strength changed insignificantly; the material preserved its plasticity at a level of 8–15%. The performed examinations allowed the prolongation of the control rod design operation for up to 20 years.

Developing and modeling of the “Laguna Verde” BWR CRDA benchmark

Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant — unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The “Laguna Verde” (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTRÉE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions.

Advanced Designs and Operating Strategies to Enhance the Safety, Operability, and Efficiency of VVER-1000 Reactors

Advanced fuel, burnable absorber, and control rod designs along with advanced fuel management and power distribution control strategies will be implemented in the first operating cycle of the Czech Republic`s Temelin VVER-1000 nuclear power plants. These improvements increase safety margins, enhance operability, and improve fuel efficiency. The Westinghouse VVANTAGE 6 fuel assembly design incorporates many proven advanced fuel and core design features used extensively in western pressurized water reactors. The fuel assembly incorporates mixing vane structural grids, radial enrichment zoning, ZrB{sub 2} integral fuel burnable absorbers, axial blankets, and Zircaloy guide thimbles and structural grids. Low-leakage loading patterns are also used to reduce radial neutron leakage. The rod cluster control assembly (RCCA) design incorporates two absorber materials. The absorber tip uses silver-indium-cadmium material while the remainder of the absorber material is B{sub 4}C enriched in {sup 10}B. This design increases control rod worth as well as the usable lifetime of RCCA. The Westinghouse constant axial offset control operating strategy, improved RCCA design, modified RCCA overlap, and replacement of part-length RCCA by full-length RCCA are used to improve the axial power distribution control capability for VVER-1000 reactors. These design improvements provide thermal margin benefits, increase shutdown margin by almost 1.0% {Delta}{rho},more » reduce fuel cycle costs by nearly 30%, and improve axial power distribution control.« less

Design and Testing of the Reactor-Internal Hydraulic Control Rod Drive for the Nuclear Heating Plant

A hydraulically driven control rod is being developed at Kraftwerk Union for integration in the primary system of a small nuclear district heating reactor. An elaborate test program, under way for ∼ 3 yr, was initiated with a plexiglass rig to understand the basic principles. A design specification list was prepared, taking reactor boundary conditions and relevant German rules and regulations into account. Subsequently, an atmospheric loop for testing of components at 20 to 90°C was erected. The objectives involved optimization of individual components such as a piston/cylinder drive unit, electromagnetic valves, and an ultrasonic position indication system as well as verification of computer codes. Based on the results obtained, full-scale components were designed and fabricated for a prototype test rig, which is currently in operation. Thus far, all atmospheric tests in this rig have been completed. Investigations under reactor temperature and pressure, followed by endurance tests, are under way. All tests to date have shown a reliable functioning of the hydraulic drive, including a novel ultrasonic position indication system.

Control rod worth in high conversion PWR.

In a High Conversion PWR (HCPWR), the neutron spectrum is hardened in comparison with the conventional PWR and the nuclear characteristics of the control rod is much different. We have calculated the control rod worth in a HCPWR for several materials of strong neutron absorbers and selected B4C for the control rod material. For the B4C control rod, we have investigated the various nuclear characteristics such as the spatial and energy dependency of neutron absorption, the dependency of the control rod worth on moderator to fuel volume ratio and Pu enrichment etc., and obtained the useful data for the nulear design of control rods in a HCPWR.

Reactivity Worths of Annular Control Rods in a Pressure-Tube-Type Heavy Water Lattice

The control rod effect has been experimentally studied in the Deuterium Critical Assembly (DCA) by using annular absorbers that simulate control rods of the FUGEN reactor, a prototype heavy-water-moderated, boiling-light-water-cooled, pressure-tube-type reactor. The DCA cores for this experiment are of the 1.2 percent-/sup 235/U-enriched UO/sub 2/ lattices, and consist of 28-pin fuel clusters arranged in a square array of 22.5-cm lattice pitch. The experiment has been carried out with various control rod patterns and with varying coolant void fraction. Experimental results were analyzed by the absorption method, which was employed in the FUGEN control rod design calculations. The calculated reactivity worth agreed with the experiment within +-10 percent. The calculations somewhat overestimated the absorber worths in the nonvoided core and underestimated them in the voided core. This tendency was found to be greatly improved by considering the anisotropy effect in the migration area of the cluster lattice. The experimental results were also analyzed by the logarithmic derivative method. This method more poorly predicted the worths, but described better the flux shape around the rods.

Irradiation Behavior of Boron Carbide

Boron carbide is the prime candidate material for neutron absorbers in fast breeder reactors and the Fast Test Reactor. Important data required for design of control rods for these reactors concern...

Thermal Buckling of Cruciform Control Rods

A criterion is derived for thermal torsional buckling of a cruciform bar subjected to transverse temperature gradients, and a numerical procedure for the evaluation of this criterion is presented. The problem is of interest in connection with the design and selection of control rods for nuclear reactors. (auth)