Journal Publications 2024

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J. Zhang a, M. Havet a, J. Zheng a, A. Bousbia Salah b, M. Ševeček c, P. Kral d, J. Krhounkova d, A. Guba e, Z. Hozer e, A. Bersano f, F. Mascari f, M. Cherubini g, T. Nemec h, M. Sánchez i,
R. Mendizábal i, C. Queral j, L.E. Herranz k, M. Adorni l, M. Bales l

a Tractebel (ENGIE), Boulevard Simon Bolivar 36, 1000 Brussels, Belgium
b BEL V (subsidiary of FANC), 148 rue Walcourt, 1070 Brussels, Belgium
c Czech Technical University in Prague, Brehova 7, 11519 Praha 1, Czech Republic
d ÚJV Rez, a. s., Hlavní 130, 25068 Husinec – Rez, Czech Republic
e Centre for Energy Research (EK), Konkoly-Thege 25-33, Budapest 1121, Hungary
f ENEA, FSN-SICNUC-SIN, Via Martiri di Monte Sole, 4, 40129 Bologna, Italy
g N.IN.E.-Nuclear and INdustrial Engineering S.r.l., Via della Chiesa XXXII, 759, 55100 Lucca, Italy
h Slovenian Nuclear Safety Administration (SNSA), Litostrojska 54, 1000 Ljubljana, Slovenia
i Consejo de Seguridad Nuclear (CSN), Pedro Justo Dorado Dellmans 11, 28040 Madrid, Spain
j Universidad Politécnica de Madrid (UPM), Ramiro de Maeztu, 7, 28040 Madrid, Spain
k CIEMAT, Avda. Complutense, 40, 28040 Madrid, Spain
l OECD Nuclear Energy Agency (NEA), 46 quai Alphonse Le Gallo, 92100 Boulogne-Billancourt, France

Nuclear Engineering and Design, Volume 425, August 2024, 113320

Abstract — Since 2012, many NEA member countries have implemented deterministic safety analyses for operating nuclear power plants under design extension conditions without significant fuel degradation or core melt (DEC-A). However, variations persist among these countries in defining DEC-A scenarios and acceptance criteria, validation and application of computer codes, development and application of deterministic safety analysis methods. Furthermore, there is a dearth of shared international experience and methodologies among various stakeholders, including regulatory authorities, technical safety or support organizations, utilities, engineering and consulting companies. To address these gaps, the OECD/NEA initiated a project in 2021, titled “Good Practices for Analyses of Design Extension Conditions without Significant Fuel Degradation for Operating Nuclear Power Plants” (or “DEC-A”), under the auspices of the Working Group on Accident Management and Analysis (WGAMA) and the Working Group on Fuel Safety (WGFS). The DEC-A project aims to review and summarize the current requirements, knowledge status, and best practices in NEA member countries. This paper outlines the objectives and scope of the OECD/NEA DEC-A project, and presents the findings from the review and discussions for each task..



Bernd Hrdy a, Raphael Zimmerl b, Marco Cherubini c, Nikolaus Müllner a


a University of Natural Resources and Life Sciences, Vienna, Department of Water, Atmosphere, and Environment, Institute of Safety and Risk Sciences, Peter-Jordan-Straße 76/1, 1190 Vienna, Austria
b Vienna Ombuds Office for Environmental Protection, Muthgasse 62, Vienna, 1190, Austria
c Nuclear and Industrial Engineering NINE Srl, Via della Chiesa XXXII, 759, Lucca, 55100, Italy

Annals of Nuclear Energy, Volume 203, August 2024, 110507

Abstract — Steam generator tube rupture (SGTR) accidents create a bypass of the containment of a pressurised water reactor (PWR) and can therefore result in the release of primary system coolant to the atmosphere via the steam relief or safety valves. In general, primary system coolant will transport radionuclides such as iodine-131. Accident management strategies for SGTR accidents therefore aim to reduce releases to the environment while ensuring core cooling.
The Downhill Simplex algorithm is used in this paper to optimise the timing of accident management measures during a SGTR accident. The secondary system steam relief and safety valves (SRV) are assumed to fail in the stuck open position at the first opening. Depressurisation of the primary system by opening the pressuriser pilot operated relief valve (PORV) and keeping the primary pressure low by shutting down two of the three trains of the high pressure injection system (HPIS) is assumed as the accident management procedure. The success of the measures is evaluated by a Relap5-3D simulation, which calculates the thermal hydraulic behaviour of the system. One of the key parameters used to assess success is the amount of iodine-131 released into the environment. The algorithm varies the timing of a set of three operator actions — opening the PORV and shutdown of HPIS trains one and two. In addition to iodine release, two other parameters are evaluated — reactor core coolant level and primary system pressure. Three normalisation functions are used to convert these parameters into a single target value, which is low when the core is covered and both primary system pressure and iodine release are low. The simplex algorithm then modifies the timing of operator actions to achieve a local minimum of the target value.
The results show that the Downhill Simplex algorithm can be used to optimise the timing of operator actions. Although timing cannot be directly implemented in Accident Management Procedures (AMPs), it is important to be aware of time sensitivity when designing AMPs. In addition, the algorithm can be adapted to optimise design parameters such as valve sizes, hydro accumulator nominal pressure levels.
The work was performed within the EURATOM R2CA project.