scholarly journals Deterministic Analysis of Beyond Design Basis Accidents in RBMK Reactors

10.5772/34501 ◽  
2012 ◽  
Author(s):  
Eugenijus Uspuras ◽  
Algirdas Kaliatk
Keyword(s):  
Deterministic Analysis ◽  
Design Basis

Safety cases for design-basis accidents in LWRs featuring passive systems

2021 ◽  
pp. 111095
Author(s):  
Thomas Mull ◽  
Thomas Wagner ◽  
Giuseppe Bonfigli ◽  
Sebastian Buchholz ◽  
Frank Schäfer ◽  
...  
Keyword(s):  
Passive Systems ◽  
Design Basis ◽  
Safety Cases

scholarly journals The SPES3 Facility for Testing an Integral Layout SMR: BDBE Simulation Analysis

2011 ◽  
Author(s):  
Roberta Ferri ◽  
Andrea Achilli ◽  
Cinzia Congiu ◽  
Gustavo Cattadori ◽  
Fosco Bianchi ◽  
...  
Keyword(s):  
Simulation Analysis ◽  
Nuclear Fission ◽  
Sensitivity Analyses ◽  
Medium Size ◽  
Design Calculation ◽  
Pressurized Water ◽  
Pressure Equalization ◽  
Plant Safety ◽  
Design Basis ◽  
Safety Features

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.

scholarly journals Computational Fluid Dynamics Modeling of Steam Condensation on Nuclear Containment Wall Surfaces Based on Semiempirical Generalized Correlations

2012 ◽  
Vol 2012 ◽  
pp. 1-7 ◽  
Author(s):  
Pavan K. Sharma ◽  
B. Gera ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Pressure Rise ◽  
Dynamics Modeling ◽  
Steam Condensation ◽  
Noncondensable Gases ◽  
Nuclear Containment ◽  
Design Basis ◽  
Lumped Parameter ◽  
Transport Calculation

In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.

Impact of random soil properties on stress–strain response

2004 ◽  
Vol 41 (2) ◽  
pp. 351-355 ◽  
Author(s):  
Dieter Stolle ◽  
Peijun Guo ◽  
Gabriel Sedran
Keyword(s):  
Soil Properties ◽  
Constitutive Modelling ◽  
Soil Parameters ◽  
Strain Response ◽  
Constitutive Behaviour ◽  
Stress Strain ◽  
Deterministic Analysis ◽  
Model Predictions ◽  
Random Variability ◽  
The Impact

This paper analyzes the impact of natural random variation of soil properties on the constitutive modelling of geomaterial behaviour. A theoretical framework for accommodating variation in soil properties is presented. The framework is then used to examine the consequence of parameter variability on stress–strain relations. An important observation is that average soil parameters from a series of tests on small specimens, in which density of the specimens varies randomly, do not necessarily reflect the average constitutive behaviour of soil. Model predictions are shown to be consistent with the experimental data.Key words: random variability, deterministic analysis, soil parameters, constitutive model.

Current status of design basis earthquake level for nuclear power plant sites in several countries

2021 ◽  
Author(s):  
Hoseon Choi ◽  
Seung Gyu Hyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
South Korea ◽  
Seismic Activity ◽  
Nuclear Power ◽  
Current Status ◽  
Seismic Safety ◽  
Design Basis ◽  
Design Earthquake ◽  
Design Construction

<p>According to strict criteria step by step for site selection, design, construction and operation, the seismic safety of nuclear power plant (NPP) sites in South Korea are secured by considering design basis earthquake (DBE) level capable of withstanding the maximum ground motions that can occur on the site. Therefore, it is intended to summarize DBE level and its evaluation details for NPP sites in several countries.</p><p>Similar but different terms are used for DBE from country to country, i.e. safe shutdown earthquake (SSE), design earthquake (DE), SL2, Ss, and maximum calculated earthquake (MCE). They may differ when applied to actual seismic design process, and only refer to approximate comparisons. This script used DBE as a representative term, and DBE level was based on horizontal values.</p><p>The DBE level of NPP sites depends on seismic activity of the area. Japan and Western United States, where earthquakes occur more frequently than South Korea, have high DBE values. The DBE level of NPP sites in South Korea has been confirmed to be similar or higher compared to that of Central and Eastern Unites Sates and Europe, which have similar seismic activity.</p>

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