[en] The chapter discusses the radiation aspects of the Belene NPP and also the possible radiological contamination of the air, the Danube river and the environment, according to the information in the technical project and the national standards on radiation safety for three specified population categories (personnel, people in the sanitary zone and the rest of the population of the country). A paragraph elaborates on what is foreseen in the project for the three categories in cases of emergency release of radioactive materials from the containment of a given unit. The conclusions are that in case of normal operation (no accidents), the NPP would not cause considerable negative consequences for the country and the population, but judging from the available information on the project the plant would not match the performance values of those in the developed countries. The authors suppose that in case of a serious accident the consequences would be heavy, because of the densely populated area in the 10 km zone around the site (3 towns and 3 villages or altogether several hundred thousand people living in the 50 km zone) and in this sense they consider the site as not well chosen. 17 refs., 7 tabs. (R.Ts.)

[en] Presented is a study of the accident confinement system of the Ignalina nuclear power plant (NPP) with Russian acronym for 'channelled large power reactor' (RBMK)-1500 reactors for the ultimate design basis accident. The computation techniques described pertain to the strength analysis of the accident confinement system (ACS): the design of the ACS compartments, materials, and manner of reinforcement. In strength predictions of reinforced concrete to evaluate the carrying-ability of the structure and its performance by the technique of limit analysis. The ACS analysis with finite elements (FE) of several types yielded instantaneous reserve strength factors for static loads. Numerical results of dynamic effects to the strength evaluation for compartment which houses the downcomers of drum separators are presented. The behaviour of pressure and temperature were taken from the results of thermal hydraulics models. Because of the relatively slow rate of loading, the investigation has confirmed that dynamic effects do not contribute additional insight to regular static analysis. (orig.)

[en] The results of prediction of the radiation conditions and damage due to fire at the unit-IV of the Chernobyl NPP are analysed. The distributions of the expected equivalent dose in the case of accidental releases of radionuclides as well as plutonium and other fusion products are obtained. Some after-effects of these accidents for the Belarus lands have been studied

[en] An original method of safety assessment has been used in order to analyze the technological risks of accident, bound with the installation of a 300 MW gas turbine power plant, on the site of Gentilly nuclear power plant. This process is based on opinions of international experts committee, and also on calculations of accident consequences. After this procedure, the Environment minister of Quebec has finally accepted this installation. In this article, the authors detail this procedure

[en] Because of the special requirements for transient multi-channel BWR core hydraulics, a completely new numerical solution method for the basic thermal-hydraulic equations had to be developed. The new numerics resides in the module, BRAN, the basic hydraulics kernel of FIBWR2. The method is based on the integral momentum approach, and is fully explicit. This paper presents the BRAN numerical method, and describes results for several test problems

[en] The feasibility of physically modeling B and W once through steam generator (OTSG) secondary side main steamline break (MSLB) phenomena in a adiabatic air-water facility is investigated in this paper. This is accomplished by use of scaling relationships that relate phenomena in air-water at atmospheric pressure to steam-water at secondary side pressure, where secondary side pressure is greatly reduced from normal operating pressure for the double ended offset shear MSLB addressed in this analysis. Scaling in an air-water flow system is feasible because of the greatly reduced secondary side pressure during a MSLB and because the phenomena of greatest importance for code development needs is hydrodynamic in nature rather than heat transfer

[en] In this paper, a generalized theoretical model is developed and numerically implemented to predict the thermal hydraulic response of a U-tube steam generator during various transients such as steam line break, steam generator tube rupture, loss of feed water, and primary side transients. Comparisons are made to the data of the Westinghouse Model Boiler No. 2 (MB-2) test for five types of transient: 100% steam line break from hot standby condition, 50% steam line break from hot standby condition, 8% steam line break with steam generator tube rupture, loss of feedwater from 100% power without reactor scram, and loss of feed water from 100% power with reactor scram. Good agreement is observed between the model predictions and the test data

[en] With the advancement of the new generation of inherently safe reactors, increasing demands are being made of the containment and its role as a safety system. Of particular importance, is the need for the safety related equipment, located within both the current and advanced containment designs, to operate and perform their intended role in mitigating the sequences of an accident. One parameter which affects the containment atmosphere and the equipment contained therein is condensation heat transfer. The second is a correlation based upon boundary layer theory and includes the effects of sensible heat transfer due to a temperature difference as well as latent heat transfer due to a concentration difference. This paper describes the two condensation heat transfer correlations discussed above as well as the Uchida correlation, evaluates each correlation, and then presents the results of a sensitivity study over a range of environmental conditions. Although the results of each correlation are comparable, the sensitivity of each is not. As the advanced reactor concepts continue to be introduced, an accurate evaluation of the interaction between the equipment, the containment and the environment within the containment is essential to assess their safety

[en] EPRI and Toshiba have jointly carried out a piping system damping evaluation study This paper reports on the study which was conducted by applying the single variable Auto Regressive Method (ARM) and the Curve-Fit Method (CFM), currently used in Japan, and the Logarithmic Decrement Method (LDM), used in the EPRI study, to the impulse test data obtained using actual plant pipe and the test data obtained using mockup prototypical nuclear pipe. Although, the dampings evaluated for actual plant pipe, using ARM, indicate good consistency with those evaluated in the EPRI study, the result should be situated as a reference only, because the applicability of ARM to the impulse test data is still questionable. The first mode damping for the mockup pipe was calculated to be in the range of 5-10%. Generally, difference sin dampings due to different evaluation methods are small when compared with the spread of the evaluated dampings due to test data variations

[en] This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion