Actinide breeding and reactivity variation in a thermal spectrum ADSR – Part 1: Development of a lumped thermal reactor model

The beneficial properties of fast reactor systems in being able to both burn and breed actinides have led to renewed interest in this technology as a means of providing a more sustainable form of nuclear power production. However, despite significant investment over many years in the development of the technology, fast reactors have never been deployed in significant numbers. In view of the difficulties encountered in fast reactor development, enhancements to the existing, well proven light water reactor (LWR) technology may provide a more accessible path to improved sustainability.     

Void distribution and bubble motion in bubbly flows in a 4×4 rod bundle. Part I: Experiments

Lack of local void fraction data in a rod bundle makes it difficult to validate a numerical method for predicting gas–liquid two-phase flow in the bundle. Distributions of local void fraction and bubble velocity in each subchannel in a 4×4 rod bundle were, therefore, measured using a double-sensor conductivity probe. Liquid velocity in the subchannel was also measured using laser Doppler velocimetry (LDV) to obtain relative velocity between bubbles and the liquid phase. The size and pitch of rods were 10 and 12.5 mm, respectively. Air and water at atmospheric pressure and room temperature were used for the gas and liquid phases, respectively. The volume fluxes of gas and liquid phases ranged from 0.06 to 0.15 m/s and from 0.9 to 1.5 m/s, respectively. Experimental results showed that the distributions of void fraction in inner and side subchannels depend not only on lift force acting on bubbles but also on geometrical constraints on bubble dynamics, i.e. the effects of rod walls on bubble shape and rise velocity. The relative velocity between bubbles and the liquid phase in the subchannel forms a non-uniform distribution over the cross-section, and the relative velocity becomes smaller as bubbles approach the wall due to the wall effects.     

Void distribution and bubble motion in bubbly flows in a 4×4 rod bundle. Part II: numerical simulation

Numerical simulations of bubbly flows in a four by four rod bundle are carried out using a multi-fluid model to examine effects of the numerical treatment of phase distribution and drag model. The transport equations of bubble number density and void fraction are used as the continuity equation of the gas phase. Two drag models are tested: one of them accounts for the bubble deformation (aspect ratio), whereas the other does not. The rod diameter, the rod pitch and the hydraulic diameter of the rod bundle are 10, 12.5 and 9.1 mm, respectively. The gas and liquid volume fluxes are J_G = 0.06 m/s and J_L = 0.9 and 1.5 m/s, respectively. The bubble diameter ranges from 1 to 5 mm. Comparisons between the numerical and measured data show that (1) the restriction on bubble lateral motion due to the presence of rods can be taken into account by using the transport equation of bubble number density, whereas that of the void fraction cannot deal with the restriction and causes large errors in the distribution of void fraction and (2) the reduction in the bubble-relative velocity near the wall is predictable by using the drag model accounting for the bubble deformation effect.     

Dynamic impact characteristics of KN-18 SNF transport cask – Part 2: Sensitivity analysis of modeling and design parameters

In Part 1 of this study, an advanced numerical simulation method was proposed to investigate the impact characteristics of the KN-18 spent nuclear fuel (SNF) transport cask recently developed in Korea and verified against the experimental results.     

Experience in operating and upgrading the No. 5 unit of the Novovoronezh nuclear power plant – practical base for developing a reliable source of nuclear energy

The No. 5 unit of the Novovoronezh nuclear power plant, starting commercial operations on September 26, 1980, is the first power-generating unit with a 1000 MW VVER in our country. The assimilation of its power gave invaluable experience to designers, builders, and equipment manufacturers; this experience was taken into account in the design solutions for next-generation power-generating units. A large volume of work on increasing the efficiency, reliability, and safety was performed over a 30-year service life. At present, the power-generating unit has been shut down for a major overhaul for upgrading according a program for extending the service life by 25–30 years.     

Vapor explosions: Prediction of a critical liquid–corium velocity in vapor destabilization mechanism of corium melt in the fuel–coolant interactions – FCI

In this paper, utilizing a simplified physical model an analytical expression for the prediction of most probable corium fragment size as function of relative velocity corium–coolant is derived. Combining this equation with the available experimental data an important consequence for the nuclear safety result: the existence of a critical velocity corium–coolant which maximizes the destabilization of film boiling when an external pressure wave is supplied.     

Swelling of steels and alloys irradiated in the BOR-60 reactor to a fluence of 1.1·1023 neutrons/cm2

Conclusions The structure created by preliminary treatment has a strong influence on the swelling, the nature of pore generation, the dislocation density, and the phase composition of 0Kh16N15M3B steel. The greatest reduction in swelling was produced by 20% cold deformation, which at a high dislocation density led to a sharp refinement of the pores and lamellar precipitations.The lamellar phases precipitated in steels of the Kh16N15 type during the irradiation after special alloying or preliminary treatment do not always determine the nature of the generation and growth of vacancy pores.At a temperature of 460 and 550°C the swelling of nickel is less by a factor of about 3, and the swelling of iron is less by a factor of about 10, than the swelling of austenized 0Kh16N15M3B steel.The least swelling was found in the Kh77TYu alloy (with 77% nickel), which before irradiation contained -phase pre-precipitations, whereas the swelling of alloys with 40 and 60% nickel is 100 times as great. From this it is clear that the nickel content is not the only factor determining the swelling.Journal version of a report delivered at the Conference on the Study of Reactor Materials (Alushta, 1978).Translated from Atomnaya Énergiya, Vol. 45, No. 6, pp. 433–439, December, 1978.     

Early stage of the crystallization in amorphous Fe–Si layers: Formation and growth of metastable α-FeSi2

Crystallization processes of amorphous Fe–Si layers have been investigated using transmission electron microscopy (TEM). Si(1 1 1) substrates were irradiated with 120 keV Fe ions at ?150 °C to a fluence of 1.0 × 10¹⁷ cm². An Fe-rich amorphous layer embedded in an amorphous Si was formed in the as-irradiated sample. Plan-view TEM observations revealed that a part of the amorphous Fe–Si layer crystallized to the metastable α-FeSi₂ after thermal annealing at 350 °C for 8 h. The lattice parameter of c-axis decreased with thermal annealing. It was considered that the change in the lattice parameter originates from the decrease of the Fe occupancy at (0, 0, 1/2) and its equivalent positions in the unit cell of the metastable α-FeSi₂.     

Crystallographic measurement of the β to α phase transformation and δ-hydride precipitation in a laser-welded Zircaloy-2 tube by electron backscattering diffraction

Crystallographic measurement of the β to α phase transformation and δ-hydride precipitation in a laser-welded Zircaloy-2 ferrule tube were carried out using an electron backscattering diffraction pattern (EBSP). A basket-weave structure with sub-micron lath width caused by quenching from the β to α phase was observed in the heat-affected and fusion zones, and mainly showed a grain boundary misorientation angle of 60° with an <1 1  0> rotation axis. This result is consistent with the Burgers orientation relationship of {1 1 0}_β//(0 0 0 1)_α and <1 1 1>_β//<1 1  0>_α for the β to α phase transformation. The texture of the quenched α′ phase was strongly inherited from the original α phase, having a radial (0 0 0 1) basal pole and axial {1 1  0} textures, even in the fusion zone. The primary hydride habit plane in the welded Zircaloy-2 was (0 0 0 1)_α//{1 1 1}_δ, matching previously obtained results for recrystallized cladding tubes. In addition to the primary habit plane, secondary habit planes were observed for the other low-index planes {1 0  0} and {1 0  1} in the fusion zone. The heterogeneous accumulation of hydrides in the transition zone between heat-affected and unaffected zones was mainly due to the residual stress distribution in the narrow region.     

Cellular neural network to the spherical harmonics approximation of neutron transport equation in x–y geometry. Part I: Modeling and verification for time-independent solution

This paper describes a novel method based on using cellular neural networks (CNN) coupled with spherical harmonics method (P_N) to solve the time-independent neutron transport equation in x–y geometry. To achieve this, an equivalent electrical circuit based on second-order form of neutron transport equation and relevant boundary conditions is obtained using CNN method. We use the CNN model to simulate spatial response of scalar flux distribution in the steady state condition for different order of spherical harmonics approximations. The accuracy, stability, and capabilities of CNN model are examined in 2D Cartesian geometry for fixed source and criticality problems.     

Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (III) – a unique simulation code to evaluate time-dependent Cs adsorption/desorption behavior in column system

A simulation code was developed to evaluate the performance of the cesium adsorption instrument operating in Fukushima Daiichi Nuclear Power Station. Since contaminated water contains seawater whose salinity is not constant, a new model was introduced to the conventional zeolite column simulation code to deal with the variable salinity of the seawater. Another feature of the cesium adsorption instrument is that it consists of several columns arranged in both series and parallel. The spent columns are replaced in a unique manner using a merry-go-round system. The code is designed by taking those factors into account. Consequently, it enables the evaluation of the performance characteristics of the cesium adsorption instrument, such as the time history of the decontamination factor, the cesium adsorption amount in each column, and the axial distribution of the adsorbed cesium in the spent columns. The simulation is conducted for different operation patterns and its results are given to Tokyo Electric Power Company (TEPCO) to support the optimization of the operation schedule. The code is also used to investigate the cause of some events that actually occurred in the operation of the cesium adsorption instrument.     

Effects of α-radiation on a direct disposal system for spent nuclear fuel – (2) review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

The Japanese geological disposal programme has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter “direct disposal of SF”) as an alternative management option other than reprocessing followed by vitrification and deep geological disposal of high-level radioactive waste (HLW). In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of α-radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.