A compact Tokamak transmutation reactor

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Tests in an MR reactor of experimental fuel elements of the VVÉR-1000 reactor

Conclusions In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVÉR-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory.The tests in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated that fuel elements of the specified design (with initial helium pressures of 1.96–2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film<3 m thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level).Translated from Atomnaya Énergiya, Vol. 62, No. 5, pp. 312–317, May, 1987.     

Fast reactor progress—slow but sure

The article sketches the history of fast reactor power station development showing how the technical development has been influenced by changes in the expectations of growth of electrical power consumption. All fast reactor development programmes have been affected by the present economic situation and the ways in which individual countries have reacted is examined. The capital cost of fast reactors is a key factor governing their rate of introduction. Areas of possible design improvements which might lead to cost reductions are outlined. Finally there is a brief review of some of the characteristics which contribute to the overall high safety of the fast reactor system and the areas in which future work may further enhance safety are identified.It is concluded that the capital cost of the reactor and the future supplies of uranium are the main factors affecting the rate of introduction of fast reactor power stations.     

Some new viewpoints in reactor noise analysis

It is propsed that the linearity criterion and order criterion via frequency spectrum features without any limitation of the model‘s phase can be used in reactor noise analysis.The time constant,natural frequency as well as the recovered transfer function of reactors can bhe obtained via the analyzable model based on reactor noise.     

Control rod drive for high temperature gas cooled reactor

This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.     

Simulation of a gas-cooled fluidized bed nuclear reactor – Part II: Stability of a fluidized bed reactor with mixed oxide fuels

Fertile materials can be converted by nuclear reaction into a transuranic mixture with a fissile content, mostly plutonium isotopes. Fuel which includes a limited proportion of plutonium is already used in some reactors. Use is restricted by the smaller delayed neutron yield and lower negative temperature coefficient of reactivity compared with uranium fuels.     

Accident source term and radiological consequences of a small modular reactor

Considering the growing global demand for energy and the need for countries to achieve climate goals, there is an increasing global interest in small modular reactors (SMRs) and their applications. Accident source term and radiological consequence evaluations of SMRs are key components of nuclear and radiation safety reviews, which affect the site, exclusion area (EAB),and low population zone outer boundaries. Based on the design characteristics of the SMR and accident analysis results, a theore...     

Using zirconium spacer lattices in the VVÉR-440 reactor

RNTs "Kurchatov Institute." Experimental Design Office "Gidropress." Translated from Atomnaya Énergiya, Vol. 73, No. 5, pp. 410–411, November, 1992.     

Plutonium charging of the VVÉR-440 water-cooled water-moderated reactor

Conclusions Our study has been limited to a neutron-physical analysis of the active zone in the presence of a fresh charge. Allowing for the disposition of the fuel assemblies this reflects a reasonably practical situation. The difficulties associated with local peaks of energy evolution may be quite easily overcome by using different fuel enrichments inside the assembly.Our example of the charging of the active zone with plutonium is of a hypothetical character, since there is far more plutonium in this arrangement than is required in any practical situation. The calculations show that there are no problems as regards the distribution of heat evolution with respect to the radius of the active zone.It would be extremely desirable to pursue this investigation with due allowance for the factors involved in the burnup of the nuclear fuel. Certain experimental work will be required in this connection so as to provide confirmation of the validity of the computer calculations. The nonuniformity of the distribution of energy evolution will be smoothed as the nuclear fuel is impoverished during burnup.Another important aspect to be studied is that of determining the weight of the control rods. We feel that a study of reactivity problems will reveal some more rigorous limitations than those deduced from the study of energy distribution undertaken in the present investigation.State Technical Scientific-Research Center. Laboratory of Nuclear-Power Technology, Helsinki, Finland. Translated from Atomnaya Énergiya, Vol. 40, No. 4, pp. 283–286, April, 1976.     

Calculating reactor transfer functions by Padé approximation via Lanczos algorithm

An effective method of solving the one-speed transport equation in frequency domain is demonstrated in this paper, the so-called Padé approximation via Lanczos algorithm (PVL). The advantage of the PVL method is that implementing the calculation process over a considerably reduced model yields a pseudo-analytical expression of the transfer function over a fairly large range of frequency. As a particular application, the dynamic transfer function of a reactor, i.e. the neutron noise induced by a localised perturbation is calculated in one-speed transport theory. The problem is essentially the same as that of the “detector-field-of-view”; studied by other authors. The PVL algorithm is demonstrated through the solution of the problem and its advantages are described. The quantitative results show that although one-speed theory was used, a local component was found, and thus the local-global decomposition could be reconstructed. This shows that unlike in diffusion theory where at least two-group theory is necessary, the local behaviour can be described already by a one-speed equation in transport theory.     

Xenon poisoning calculation code for miniature neutron source reactor（MNSR）

In line with the actual requirents and based upon the specific characteristics of MNSR,a revised point-reactor model was adopted to model MNSR‘s xenon poisoning.The corresponding calculation code.MNSRXPCC(Xenon Poisoning Calculation Code for MNSR),was developed and tested by the Shanghai MNSR data.     

Investigation on two-phase flow instability in steam generator of integrated nuclear reactor

In the pressure range of 3-18MPa,high pressure steam-water two-phase flow density wave instability in vertical upward parallel pipes with inner diameter of 12mm is studied experimentally.The oscillation curves of two-phase flow instability and the effects of several parameters on the oscillation threshold of the system are obtained.Based on the small pertubation linearization method and the stability principles of automatic control system,a mathematical model is developed to predict the characteristics of density wave instability threshold.The predictions of the model are in good agreement with the experimental results.     

Experimental study on the mechanism of flow blockage formation in fast reactor

Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor (e.g., through chemical interaction between the coolant and impurities, air, or water, through corrosion of structural materials, or from damaged/molten fuel). Such particles may cause flow blockage accidents in a fuel assembly, resulting in a reduction in coolant flow,which potentially causes a local temperature rise in the fuel cladding, cladding failure, and fuel melt. To understand the bl...     

Development and validation of a fast reactor core burnup code – FARCOB

A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X–Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.     

Effect of reprocessing on neutrons of a molten chloride salt fast reactor

Due to their unique features, such as the inherent safety, simplified fuel cycle, and continuous on-line reprocessing, molten salt reactors (MSRs) are regarded as one of the six reference reactors in the Generation IV International Forum (GEN-IV).Molten chloride salt fast reactors (MCFRs) are a type of MSR. Compared to molten fluoride salt reactors (MFSRs), MCFRs have a higher solubility of heavy metal atoms, a harder neutron spectrum, lower accumulation of fission products (FPs), and better bre...     

Composition of reactor radiation—statistically significant radiation damage factor of diamond

Conclusions