Michel Giot

Nikola &#268;avlina

Dubravko Pevec

Alessandro Petruzzi

Selected Papers from TopSafe 2008: Safety at Nuclear Installations

This paper presents a methodology and its application to a Level 2 Probabilistic Safety Assessment (PSA-2), to evaluate the impact of the Passive Autocatalytic Recombiners (PARs) on the hydrogen risk in the reactor containment in case of a severe accident. Among the whole set of accidental scenarios calculated in the framework of the PSA-2, nine have been selected as representative in terms of the in-vessel hydrogen production rate and in-vessel total produced hydrogen mass. Five complementary scenarios have been added as representative of the core reflooding situations.  For this set of selected scenarios the evolution of the conditions in the containment (i.e., pressure, temperature, and composition) during the in-vessel phase of the accident has been evaluated by means of a lumped parameter approach. The use of spray systems in the containment has also been considered as well as the presence of recombiners. Moreover, the ignition by recombiners of the flammable atmosphere has been considered.

Ahmed Bentaib

Cataldo Caroli

Bernard Chaumont

Karine Chevalier-Jabet

Evaluation of the Impact That PARs Have on the Hydrogen Risk in the Reactor Containment: Methodology and Application to PSA Level 2

This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work) of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER). The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model), represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection.  The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

Luben Sabotinov

Heikki Purhonen

Vesa Riikonen

CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). SDTP consists of nearly 60 organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, and the United States. Innovative Systems Software (ISS) is the administrator for the program. RELAP/SCDAPSIM is used by program members and licensed users to support a variety of activities. The paper provides a brief review of some of the more important activities including the analysis of research reactors and Nuclear Power Plants, design and analysis of experiments, and training.

C. M. Allison

J. K. Hohorst

Role of RELAP/SCDAPSIM in Nuclear Safety

Phase IV of BEMUSE Program is a necessary step for a subsequent uncertainty analysis. It includes the simulation of the reference scenario and a sensitivity study. The scenario is a LBLOCA and the reference plant is Zion 1 NPP, a 4 loop PWR unit. Thirteen participants coming from ten different countries have taken part in the exercise. The BEMUSE (Best Estimate Methods plus Uncertainty and Sensitivity Evaluation) Program has been promoted by the Working Group on Accident Management and Analysis (WGAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI).
The paper presents the results of the calculations performed by participants and emphasizes its usefulness for future uncertainty evaluation, to be performed in next phase. The objectives of the activity are basically to simulate the LBLOCA reproducing the phenomena associated to the scenario and also to build a common, well-known, basis for the future comparison of uncertainty evaluation results among different methodologies and codes. The sensitivity calculations performed by participants are also presented. They allow studying the influence of different parameters such as material properties or initial and boundary conditions, upon the behaviour of the most relevant parameters related to the scenario.

M. Pérez

F. Reventós

L. Batet

R. Pericas

I. Toth

P. Bazin

A. de Crécy

P. Germain

S. Borisov

H. Glaeser

T. Skorek

J. Joucla

P. Probst

A. Ui

B. Chung

D. Y. Oh

M. Kyncl

R. Pernica

A. Manera

F. D'Auria

A. Petruzzi

A. Del Nevo

Main Results of Phase IV BEMUSE Project: Simulation of LBLOCA in an NPP

Decay Heat Removal (DHR) is a fundamental safety function which is often accomplished in the advanced LWRs relying on natural phenomena. A typical passive DHR system is the two-phase flow, natural circulation, closed loop system, where heat is removed by means of a steam generator or heat exchanger, a condenser, and a pool.
Different condenser tube arrangements have been developed for applications to the next generation NPPs. The two most used configurations, namely, horizontal and vertical tube condensers, are thoroughly investigated in this paper. Several thermal-hydraulic features were explored, being the analysis mainly devoted to the description of the best-estimate correlations and models for heat transfer coefficient prediction. In spite of a more critical behaviour concerning thermal expansion issues, vertical tube condensers offer remarkably better thermal-hydraulic performances. 
An experimental validation of the vertical tube correlations is provided by PERSEO facility (SIET labs, Piacenza), showing a fairly good agreement.

Davide Papini

Antonio Cammi

Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

The study of fuel behaviour under accidental conditions is a major concern in the safety analysis of the Pressurised Water Reactors. The consequences of Design Basis Accidents, such as Loss of Coolant Accident and Reactivity Initiated Accident, have to be quantified in comparison to the safety criteria. Those criteria have been established in the 1970s on the basis of experiments performed with fresh or low irradiated fuel. Starting in the 1990s, the increased industrial competition and constraints led utilities to use fuel in more and more aggressive conditions (higher discharge burnup, higher power, load follow, etc.) and create incentive conditions for the development of advanced fuel designs with improved performance (new fuel types with additives, cladding material with better resistance to corrosion, etc.). These long anticipated developments involved the need for new investigations of irradiated fuel behaviour in order to check the adequacy of the current criteria, evaluate the safety margins, provide new technical bases for modelling and allow an evolution of these criteria. Such an evolution is presently under discussion in France and several other countries, in view of a revision in the next coming years. For this purpose, a R&#38;D strategy has been defined at IRSN.

Fran&#231;ois Barr&#233;

Claude Grandjean

Marc Petit

Jean-Claude Micaelli

Fuel R&#38;D Needs and Strategy towards a Revision of Acceptance Criteria

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shut down for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. The paper summarizing the results of deterministic and probabilistic analyses is developed within 1991&#8211;2007 by specialists from Lithuanian Energy Institute. The main operational safety aspects, including analyses performed according the Ignalina Safety Improvement Programs, development and installation of the Second Shutdown System and Guidelines on Severe Accidents Management are discussed. Also the phenomena related to the closure of the gap between fuel channel and graphite bricks, multiple fuel channel tube rupture, and containment issues as well as implication of the external events to the Ignalina NPP safety are discussed separately.

E. U&#353;puras

State of the Art of the Ignalina RBMK-1500 Safety

The paper presents validation results for multichannel vessel thermal-hydraulic models in CATHARE used in coupled 3D neutronic/thermal hydraulic calculations. The mixing is modeled with cross flows governed by local pressure drops. The test cases are from the OECD VVER-1000 coolant transient benchmark (V1000CT) and include asymmetric vessel flow transients and main steam line break (MSLB) transients. Plant data from flow mixing experiments are available for comparison.  Sufficient mesh refinement with up to 24 sectors in the vessel is considered for acceptable resolution. The results demonstrate the applicability of such validated thermal-hydraulic models to MSLB scenarios involving thermal mixing, azimuthal flow rotation, and primary pump trip. An acceptable trade-off between accuracy and computational efficiency can be obtained.

I. Spasov

J. Donov

N. P. Kolev

L. Sabotinov

CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis

The Institute of Neutron Physics and Reactor Technology (INR) is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV) as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

V. S&#225;nchez

W. Jaeger

M. Boettcher

B. Truong

Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

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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

Selected Papers from TopSafe 2008: Safety at Nuclear Installations

Selected Papers from TopSafe 2008: Safety at Nuclear Installations

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