An experimental investigation of a passive cooling unit for nuclear plant containment

A set of condensation experiments in the presence of noncondensables (e.g. air, helium) was conducted to evaluate the heat removal capacity of a passive cooling unit in a post-accident containment. Condensation heat transfer coefficients on a vertically mounted smooth tube have been obtained for total pressure ranging from 2.48×10⁵ Pa(abs) to 4.55×10⁵ Pa(abs) and air mass fraction ranging from 0.30 to 0.65. An empirical correlation for heat transfer coefficient (h), has been developed in terms of a parameter group made up of steam mole fraction (Xs), total pressure (P_t), temperature difference between bulk gas and wall surface (dT). This correlation covers all data points within 20%. All data points are also in good agreement with the prediction of the diffusion layer model (DLM) with suction and are approximately 2.2 times the Uchida heat transfer correlation. Experiments with an axial shroud around the test tube to model the restriction on radial flow experienced within a tube bundle demonstrated a reduction of the heat transfer coefficient by a factor of about 0.6. The effect of helium (simulating hydrogen) on the heat transfer coefficient was investigated for helium mole fraction in noncondensable gases (X_He/X_nc) at 15, 30 and 60%. It was found that the condensation heat transfer coefficients are generally lower when introducing helium into noncondensable gas. The difference is within 20% of air-only cases when X_He/X_nc is less than 30% and total pressure is less than 4.55×10⁵ Pa(abs). A gas stratification phenomenon was clearly observed for helium mole fraction in excess of 60%.     

Deposition of boron on fuel rod surface under sub-cooled boiling conditions—An approach toward understanding AOA occurrence

In PWR primary coolant, it has been assumed that Li and B ions deposited on fuel rod surface under sub-cooled boiling conditions and they changed their chemical forms by chemical reaction with nickel iron oxides on the fuel surface. Accumulated boron on the fuel led to axial offset anomaly (AOA). In the present paper, the amount of boron deposited on the fuel surface was evaluated from two directions. The first calculated the amount with the extended micro-layer evaporation and dry-out (MED) model and the other estimated it from the viewpoint of reactor reactivity (neutron economy calculation).The MED model, which was developed for predicting iron crud deposition on the boiling surface of BWR fuel rods, was extended for application to metallic ion deposition, and modified to evaluate deposition of crud and metallic ions on sub-cooled boiling surface. Processes of growth and collapse of bubbles were calculated to determine the time from bubble generation to collapse and total evaporation volume and deposition amount of boron and metallic ions and their oxides on the fuel rod surface for a bubble. Finally chemical reaction rates of boron and metallic ions were calculated in the deposits.From the evaluation, it was concluded that: (i) the calculated deposition amount of boron on the fuel rod surface, which was four or forty times larger than measured amounts of boron and nickel oxides compounds, was seldom measured in the fuel deposits due to its high release rate; (ii) its hideout return during the reactor shutdown period was seldom observed due to its high concentration in the primary coolant; (iii) one of the most promising approaches to evaluate its accumulation on the fuel rod surface during plant operation was the MED model calculation; and (iv) control of nickel concentration in the primary coolant resulted in decreased nickel oxide deposition and then mitigation of AOA occurrence due to decreasing average residence time of boron on the fuel rod surface.     

A numerical investigation of three-dimensional flows in large volumes in the context of passive containment cooling in BWRs

The paper describes Computational Fluid Dynamics (CFD) calculations undertaken in support of analyses of three-dimensional flows that take place in the drywell volumes of advanced boiling water reactors with passive decay–heat removal systems. Data for comparison are taken from the 1/40th-scale European Simplified Boiling Water Reactor (ESBWR) mock-up facility PANDA under conditions of symmetric steam injection and asymmetric outflow. Steady-state simulations for pure steam conditions illustrate how the separate flow streams mix to ensure balanced outflow conditions to the condenser units. A transient calculation has also been performed to examine how air, assumed to be released from solution in the PANDA boiler, would ultimately accumulate in the separate condenser units. Results provide a possible explanation for the rundown in performance of one of the condensers, behaviour which was repeatedly observed in some of the earlier PANDA tests. The work also provides more general insights on how trace amounts of non-condensable gases may accumulate in passive cooling equipment.     

Research on operational characteristics of passive residual heat removal system under rolling motion

The operational characteristics of passive residual heat removal system under rolling motion were investigated experimentally. The passive residual heat removal system under rolling motion was simulated with the advanced RELAP5 code. The results are consistent with experiments. The relative discrepancy between calculating and experimental results is less than 10%. The modified condensation heat transfer model can also be used to calculate the condensation heat transfer coefficient with droplet carryover precisely. The fluctuation of condensate temperature and steam pressure is not noticeable. As the power becomes larger for the same rolling motion, the oscillation amplitude of condensate flow rate becomes larger. The effect of rolling motion upon heat transfer coefficient and flow resistance was investigated with experimental results. Rolling motion can increase the flow resistance in a great extent. The more serious the rolling is, the more the flow resistance is. Additional pressure drop does not effect on average flow velocity. The decreasing of average flow velocity is due to the decreasing average gravity pressure drop and the increasing of flow resistance. The contribution of gravity pressure drop on the decrement of average flow velocity is less than 20%. The other is due to the increasing flow resistance. In the present paper, the experimental results are listed first, and then the simulation results comparing with the experimental results are listed in the second part. At last, the effect of rolling motion is investigated theoretically.     

An experimental investigation on the local behaviour of steel-concrete interfaces at large openings in the PSC inner containment dome

An experimental study has been conducted to know the structural behaviour of a representative test specimen for a typical Prestressed Concrete Inner Containment dome with due regards to the local behaviour at the steel plate-concrete interfaces at large steam generator openings, under initial condition at prestress transfer. The primary thrust of the work has been in the objective of predicting possibilities of separation at the steel-concrete interface zones adjacent to the embedded plates of the steam generator openings of the inner containment dome. Important observations in line with the desired objectives have been made based on the results of the experiment.     

An investigation of transition boiling mechanisms of subcooled water under forced convective conditions

A mechanistic model for forced convective transition boiling has been developed to investigate transition boiling mechanisms and to predict transition boiling heat flux realistically. This model is based on a postulated multi-stage boiling process occurring during the passage time of the elongated vapor blanket specified at a critical heat flux (CHF) condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling characterized by the frequent touches of the interface and the heated wall. The total heat transfer rates after the DNB is weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. The parametric effects of pressure, mass flux, inlet subcooling on the transition boiling heat transfer are also investigated. From these comparisons, it can be seen that this model can identify the crucial mechanisms of forced convective transition boiling, and that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are well predicted at low qualities/high pressures near 10 bar. In future, this model will be improved in the unstable film boiling stage and generalized for high quality and low pressure situations.     

An investigation of the degradation characteristics for casting stainless steel, CF8M, under high temperatures

Stainless steel castings are used in pipes and valves subjected to high pressure and temperatures. The primary coolant system of a nuclear power plant is made of a stainless steel casting and the operating temperatures are in the range of 290–330°C. If the coolant system is exposed to these temperature ranges for a long period, it may be possible to experience degradation of the material. The present investigation is concerned with the degradation characteristics of CF8M (cast duplex stainless steel), exposed to the thermal and σ-phase degradation temperatures, 430 and 700°C, respectively. After the CF8M specimens are held 100–3600 h at 430°C for the thermally degraded specimens and maintained 20 min to 150 h at 700°C for the σ-phase degraded specimens, respectively, all specimens are water quenched. Each specimen of the thermally and σ-phase degraded materials is classified into five classes depending on the holding time at the given temperatures. In order to investigate the characteristics of degradation, microstructure, micro Vickers hardness, tensile, impact tests, and fatigue crack growth tests are performed for each class of the specimens. From the present investigation the following results were obtained: (1) the difference between the thermally and σ-phase degraded specimens can be distinguished through their microstructures, (2) hardness and tensile strength are increased with degradation, while elongation, reduction area, and impact energy are decreased by increasing the degradation, (3) the fatigue crack growth rate (FCG) of the σ-phase degradation at 700°C is larger than that of the thermally degraded specimens, and (4) the FCG for both thermally and σ-phase degraded specimens are larger than those of the virgin (nondegraded) specimens.     

Development and investigation of the regulator of the passive heat-removal system for the Kudankulam nuclear power plant (India)

A system for passive removal of the residual heat released in VVéR-1000 with a heat-removal regulator in the form of an air damper operating passively is described. The main results of the computational and experimental validation of the operability of the regulator are presented. It is shown that the design proposed for the regulator gives the nominal characteristics and enables stable operation of the system in hot stand-by and cool-down regimes of the reactor system, and it can be used in the Kudankulam nuclear power plant (India) which is now under construction. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 49–52, January, 2007.     

An investigation of uranium sorption on minerals of the surrounding rock

Present technological methods do not ensure complete extraction of uranium from ores, because a definite amount of it remains in the solid part of the residue, even under conditions of severe leaching. The bulk of the over-all uranium losses during hydrometallurgical processing of ores is due to this fact.This paper gives experimental data confirming that one of the causes of uranium losses with the solid part of the residue is its sorption on minerals of the surrounding rock. The sorption capacity of such minerals as montmorillonite, kaolinite, bauxite, and albite, and the possibility of uranium desorption by solutiom of various salts and acids, were investigated.Translated from Atomnaya Énergiya, Vol. 16, No. 1, pp. 51–55, January, 1964     

An investigation of the L-shell x-ray conversion efficiency for laser-irradiated tin foils

We have used the Shenguang Ⅱ laser in third harmonic(351 nm) to investigate the emission of L-shell radiation in the 3.3–4.4 ke V range generated using thin foils of Sn coated onto a parylene substrate with irradiation of order 10~(15) W cm~(-2) and nanosecond pulse duration. In our experiment, we have concentrated on assessing the emission on the non-laser irradiated side as this allows an experimental geometry relevant to experiments on photo-ionised plasmas where a secondary target must be placed close to the source, to achieve x-ray fluxes appropriate to astrophysical objects. Overall L-shell conversion efficiencies are estimated to be of order 1%,with little dependence on Sn thickness between 400 and 800 nm.     

An investigation into the dependence of rod response coefficients on end effects

Perturbations caused by homogeneous neutron absorbing and scattering rods to the thermal neutron flux distribution within a volume of moderator are measured and compared with response coefficient theory, which has been developed for idealized problems. It is found that although the theory predicts the shape of the asymptotic flux perturbation correctly in general, it overestimates the size of the perturbation. It is shown that this discrepancy can be attributed primarily to rod end effects, which in turn suggests that the theory should be used with caution in reactor calculations where these effects are important.     

Experimental investigation of the buckling of nuclear containment-like cylindrical geometries under combined shear and bending

This paper describes the method for and results of, determining the static buckling interaction curves for both ring-stiffened and unstiffened cylindrical geometries that have radius-to-thickness (R/t) ratios and other parameters characteristic of nuclear steel containments. The purpose of developing these test methods and interaction curves is to have this information available for a dynamic buckling study on the same or similar shells that will be directed at answering questions regarding the freezing-in-time analysis method.     

An investigation of the properties of metals and some steels after irradiation by fast neutrons

The article considers the effect of the irradiation by fast neutrons and of a subsequent heat-treatment on the properties of some metallic materials. The change of the properties of materials upon irradiation is explained in terms of the formation of various types of defects on the crystal lattice, which are annealed at appropriate temperatures. The kinetics of the defect elimination processes leading to a strengthening of the material is studied, and their activation energy is determined.The authors express their gratitude to V. M. Agranovich for his useful participation in the discussion of the results of this work.     

As离子束混合形成硅化钨的卢瑟福背散射分析

本工作用卢瑟福背散射研究了离子束混合方法形成硅化钨的条件。通过对热烧结和离子束混合法的比较,结合X射线衍射分析相结构的结果,指出离子束混合法可降低形成二硅化钨所需的退火温度。观察到氧杂质对形成条件的影响,简单讨论了离子束混合形成硅化物的机理。     

An investigation of the effect of Co60 γ-radiation on strongly basic anionites AV-17 and AM

The effect of Co⁶⁰ -radiation on strongly basic anionites AV-17 (with 6 and 16% divinyl benzene) and AM in water was investigated. It was found that as a result of irradiation by 4.7.10⁸ tad, the average decrease in weight of the anionites is 23%; the capacity of AV-17Kh16, AV-17Kh6 and AM anionites decreases at the same time by 73%, 85% and 940]0 respectively, and their specific volumes by 35%, 45%, and 43% respectively. When AV'lTKh6 anionite is irradiated with 1.9·10⁸ rad, its capacity and specific volume are reduced by 56% and 27% respectively. AV-17Kh6 anionite may evidently be used for practical purposes at doses not exceeding 1·10⁸ rad.There are grounds for assuming that the main water-soluble alkaline radiolysis products of AV-17Kh6 and AV-17Kh16,are ammonia and trimethylamine, respectively. AV-17Kh16 is more resistant to the effect of radiation.Translated from Atomnaya Énergiya, Vol. 16, No. 3, pp. 245–249, March, 1963     

<Superscript>59</Superscript>Ni Accumulation in Corrosion-Resistant Steel under the Action of Neutron Flux

A γ-ray line with energy E_γ = 11.3 MeV was detected during an experiment, performed on a nuclear reactor, investigating the characteristics of the energy spectrum of γ-rays. The most likely source of this line is radiative capture of thermal neutrons by ⁵⁹Ni nuclei, which accumulated in the corrosion-resistance steel as a result of the more than 20 years of irradiation in the reactor, via the reaction ⁵⁸Ni(n, γ)⁵⁹Ni. It was found that for thermal-neutron fluence 10²¹ cm⁻² the ⁵⁹Ni concentration is 0.47% of the ⁵⁸Ni concentration. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 268–272, October, 2005.     

Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of ‘overpressure-molten pool structure’ when the vessel failure started at the ‘hot’ layers of the vessel. It was shown in this study that the processes in the molten pools reach a quasistationary state at 2000…3000 s after molten pool formation. Numerical results in this paper illustrate that the large creep deformations of the vessel lower head can lead to an appearance of the gaps between the vessel surface and the molten pool crust. It is obvious that the joint thermal and structural analyses are needed for the accurate tracing of the initial bounds of the vessel and molten pool during simulations.