Lorenzo Santini

Davide Papini

Marco E. Ricotti

Experimental Characterization of a Passive Emergency Heat Removal System for a GenIII<svg style="vertical-align:-0.04094pt" height="8.5" version="1.1" viewBox="0 0 10.1 8.5" width="10.1" xmlns:xlink="http://www.w3.org/1999/xlink" xmlns="http://www.w3.org/2000/svg"><g transform="matrix(.017,-0,0,-.017,.062,8.387)"><path id="x2B" d="M535 230h-212v-233h-58v233h-213v50h213v210h58v-210h212v-50z" /></g></svg> Reactor

To estimate the success criteria time windows of operator actions the conservative approach was used in the conventional probabilistic safety assessment (PSA). The current PSA standard recommends the use of best-estimate codes. The purpose of the study was to estimate the operator action success criteria time windows in scenarios in which the human actions are supplement to safety systems actuations, needed for updated human reliability analysis (HRA). For calculations the RELAP5/MOD3.3 best estimate thermal-hydraulic computer code and the qualified RELAP5 input model representing a two-loop pressurized water reactor, Westinghouse type, were used. The results of deterministic safety analysis were examined what is the latest time to perform the operator action and still satisfy the safety criteria. The results showed that uncertainty analysis of realistic calculation in general is not needed for human reliability analysis when additional time is available and/or the event is not significant contributor to the risk.

Andrej Prošek

Borut Mavko

RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis

In the course of a postulated severe accident in an NPP, Direct Containment Heating (DCH) may occur after an eventual failure of the vessel. DCH is related to dynamical, thermal, and chemical phenomena involved by the eventual fine fragmentation and dispersal of the corium melt out of the vessel pit. It may threaten the integrity of the containment by pressurization of its atmosphere. Several simplified modellings have been proposed in the past but they require a very strong fitting which renders any extrapolation regarding geometry, material, and scales rather doubtful. With the development of multidimensional multiphase flow computer codes, it is now possible to investigate the phenomenon numerically with more details. We present an analysis of the potential of the MC3D code to support the analysis of this phenomenon, restricting our discussion to the dynamical processes. The analysis is applied to the case of French 1300 MWe PWR reactors for which we derive a correlation for the corium dispersal rate for application in a Probabilistic Safety Analysis (PSA) level 2 study.

Renaud Meignen

Tanguy Janin

On the Analysis and Evaluation of Direct Containment Heating with the Multidimensional Multiphase Flow Code MC3D

RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Kr&#x161;ko with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Sini&#x161;a &#x160;adek

Sr&#x111;an &#x160;palj

Bruno Glaser

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety (NRS) analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz  and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

T. Höhne

E. Krepper

U. Rohde

Application of CFD Codes in Nuclear Reactor Safety Analysis

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Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.

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Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Science and Technology of Nuclear Installations

Selected Papers from TopSafe 2008: Safety at Nuclear Installations

Selected Papers from TopSafe 2008: Safety at Nuclear Installations

Articles

Experimental Characterization of a Passive Emergency Heat Removal System for a GenIII Reactor

RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis

On the Analysis and Evaluation of Direct Containment Heating with the Multidimensional Multiphase Flow Code MC3D

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

Application of CFD Codes in Nuclear Reactor Safety Analysis