Water electrolyzer for renewable energy systems

The article is devoted to the features of the alkaline water electrolyzers use in power plants with a hydrogen energy storage systems based on renewable energy sources. The technology of nickel–cobalt electrodes electrochemical formation according to a printed 2–dimensional sketch is proposed. A new technique for the synthesis of diaphragms with a zirconium hydroxide hydrogel as a hydrophilic filler is considered. The current–voltage characteristics of an electrolytic cell located inside outer containment shell, designed for pressures up to 160 atm, are investigated.

Methodical Apparatus for the Development of Typical Scenarios of Beyond Design Basis Accidents with VVER Reactors in the Planning and Implementation of Protective Measures for the Population

To develop a simple methodological apparatus for constructing typical scenarios of accidents and estimate of emissions of radioactive substances from nuclear power plants with VVER reactors in the planning and implementation of protective measures for the population.Material and methods_ To build the model of release of radionuclides into the environment in case of an accident used literary experimental data of outputs of the radioactive substances from the fuel when it is heating and melting, the destruction of the fuel rod cladding, the use of technical means retention of radioactive substances in the containment of the reactor and their behaviour in the containment (sedimentation, sorption, etc.).Results_ Mathematical apparatus has been developed to quantify the magnitude of emissions of radioactive substances at nuclear power plants with VVER reactors depending on the time for dose-forming radionuclides in the instantaneous rupture of cooling pipelines in the presence of additional failure of NPP safety systems. The amount of emission of each deterministic of an accident correspond to the level 4 to 7 of the INES scale. Radionuclide release into the environment was calculated in the following physical and chemical processes_ heating and melting of fuel, depressurization of fuel rod claddings, peculiarities of the behavior of radionuclides in the containment shell (deposition, etc.) and operation of technical means to ensure radiation safety, structural leakages of the buildings. As an example, the assessment of the release of radionuclides into the environment depending on the time for level 5 of the INES accident scale is given. Conclusion_ Methodical apparatus can be used in the construction of typical scenarios for the development of basis accidents and assessment of radioactive emissions at nuclear power plants with VVER reactors in the planning and implementation of protective measures for the population and emergency exercises and training_

Air Velocity Influence on Heat Transfer in the PCCS

The passive containment cooling system (PCCS) is important passive safety systems in the Advanced Pressurized Water Reactor (APWR), which belongs to the generation III of nuclear reactors. In design basis accident (DBA), the steam condenses on the inner surface of the containment shell and the cooling water evaporates from the outer surface of the containment shell. In this process, the heat is transferred from the inside of the containment to the outside. To span the expected range of conditions and provide a proper model for evaluating the inner steam condensation coupled with outer evaporation heat transfer process, the inner steam condensation coupled outer evaporation experimental test (ISCOE) is developed by State Nuclear Power Technology Research & Development Centre (SNPTRD). Several tests have been done on the ISCOE experimental test facility. The influence of different key factors for the capability of the heat transfer of the containment steel shell wall has been researched. Key factors include steam pressure, steam temperature, water film velocity, air velocity, steel shell wall angle, and so on. The result of these tests has an important significance to the research of heat transfer capability of the containment steel shell wall. In this paper, several tests are introduced, including details, results and analysis. The influence of air velocity for the capability of the heat transfer of the containment steel shell wall is also analyzed.

3D Linear Bifurcation Analysis of Steel Containment Vessel

The AP1000® Containment Vessel (CV) is a freestanding steel containment designed to protect the public from radiation release. The CV consists of 2 ellipsoidal heads connected by a cylindrical shell and is constructed of carbon steel. The AP1000 plant design has four large penetrations (two airlocks and two equipment hatches) located in approximately the same quadrant of the circumference of the shell which imposes asymmetric effects in the shell structure. The CV is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Subsection NE. Traditionally, the local and global stability of freestanding steel containments have been designed by use of formulae using conservative assumptions based on an axisymmetric structure. ASME Code Case N-284 “Metal Containment Shell Buckling Design Methods, Class MC Section III, Division 1” outlines methodology for satisfying the stability of the CV using two approaches. Section 1710 provides a stress based buckling approach using detailed formulae that assumes an axisymmetric structure. The second approach provides guidance and acceptability based on a linear bifurcation analysis (2D (1720) or 3D (1730)). Due to the asymmetric structure of the CV, the 3D linear bifurcation method delivers the most accurate results. The methodology and assumptions implemented by Westinghouse to qualify stability of the CV via Code Case N-284 are outlined. Also, the procedure to properly amplify the stresses as required by N-284 is included as justification of the methods used. This justification was thoroughly investigated by the Nuclear Regulatory Commission (NRC) and deemed acceptable.

On Features of Thermal Design of Passive Evaporation-and-Condensation Systems of Reactor Thermal Shielding

Further development of nuclear engineering is inseparably linked with the requirement of vast application of the passive systems of heat removal running without human intervention. Creation of such systems is impossible, if only conventional engineering solutions are used. As known, to prevent propagation of the fission products into the environment there are three safety barriers. To provide operation of the third safety barrier (containment shell), in particular, of the reactor cavities both in operational and emergency modes a passive evaporation-and-condensation (EC) system of heat removal is proposed. The features of thermal design of the EC systems for thermal shielding of the reactor cavities are considered. They make it possible to determine the optimal main design variables of the EC systems and prove reasonability and efficiency of their application. The performed study validates engineering feasibility of an efficient EC system for thermal shielding of the reactor equipment.

A Study of GOTHIC 8.0 Code Application to AP1000 Containment Response

The AP1000 containment model has been developed by using WGOTHIC version 4.2 code. Condensation heat and mass transfer from the volumes to the containment shell, conduction through the shell, and evaporation from the shell to the riser were all calculated by using the special CLIMEs model. In this paper, the latest GOTHIC version 8.0 code is used to model both condensation and evaporation heat and mass transfer process. An improved heat and mass transfer model, the diffusion layer model (DLM), is adopted to model the condensation on the inside wall of containment. The Film heat transfer coefficient option is used to model the evaporation on the outside wall of containment. As a preliminary code consolidation effort, it is possible to use GOTHIC 8.0 code as a tool to analysis the AP1000 containment response.