INTERNATIONAL FUEL PERFORMANCE STUDY OF FRESH FUEL EXPERIMENTS FOR PCMI EFFECTS DURING RIA EXPERIMENTS

Seokbin Seo ^a, Charles Folsom ^a, Colby Jensen ^a, David Kamerman ^a, Luana Giaccardi ^b, Marco Cherubini ^b, Pavel Suk ^b, Martin Sevecek ^c, Jerome Sercombe ^d, Isabelle Guenot-Delahaie ^d, Alessandro Scolaro ^e, Matthieu Reymond ^e, Katalin Kulacsy ^f, Luis Herranz ^g, Francisco Feria ^g, Pau Aragón ^g, Grigori Khvostov ^h, Imran Khan ⁱ, Anuj Kumar Deo ^j, Srinivasa Rao Ravva ^j, Rolando Calabrese ^k, Felix Boldt ^l, Jonathan Sappl ^l, Florian Falk ^l, Asko Arkoma ^m, Georgenthum Vincent ⁿ, Yudai Tasaki ^o, Kazuo Kakiuchi ^o, Yutaka Udagawa ^o, Gregory Delipei ^p, Charles Cheron ^p, James Corson ^q, Jinzhao Zhang ^r, Thomas Drieu ^r, Jan Klouzal ^s, Martin Dostal ^s, Vitezslav Matocha ^s, Tereza Kinkorová ^t, Carlo Fiorina ^u

^a Idaho National Laboratory (INL), USA
^b Nuclear and Industrial Engineering (NINE), Italy
^c ALVEL, Czech Republic
^d French Alternative Energies and Atomic Energy Commission(CEA), France
^e EPFL, Switzerland
^f HUN-REN Centre for Energy Research (HUN-REN EK-CER), Hungary
^g Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas(CIEMAT), Spain
^h Paul Scherrer Institute(PSI), Switzerland
ⁱ Bhabha Atomic Research Centre(BARC), India
^j Atomic Energy Regulatory Board(AERB), India
^k ENEA, Italy
^l Gesellschaft für Anlagen und Reaktorsicherheit gGmbH(GRS), Germany
^m Technical Research Centre of Finland(VTT), Finland
ⁿ Institut de Radioprotection et de Surete Nucleaire(IRSN), France
^o Japan Atomic Energy Agency(JAEA), Japan
^p North Carolina State University(NCSU), USA
^q Nuclear Regulatory Commission(NRC), USA
^r TRACTEBEL, Belgium
^s UJV, Czech Republic
^t Czech Technical University(CTU), Czech Republic
^u Texas A&M University (TAMU), USA

Nuclear Engineering and Design, Volume 430, 15 December 2024, 113673

Abstract — This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling & Simulation (M&S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. The blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal–mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal–mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.

TOWARD A BETTER UNDERSTANDING OF REFLOOD THERMAL HYDRAULICS: A SUMMARY OF THE OECD/NEA RBHT PROJECT

Stephen M. Bajorek ^a, Brian Lowery ^b, Fan-Bill Cheung ^c, Alessandro Del Ferraro ^d, Marco Cherubini ^d, Alessandro Petruzzi ^d, Jinzhao Zhang ^e, and Martina Adorni ^f

^a U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001
^b The Pennsylvania State University, Applied Research Laboratory, University Park, Pennsylvania 16802
^c The Pennsylvania State University, Department of Mechanical Engineering and Nuclear Engineering, UniversityPark, Pennsylvania 16802
^d Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, Lucca, Italy
^e Tractebel (ENGIE), Boulevard Simon Bolivar 34-36, 1000 Brussels, Belgium
^f Organisation for Economic Co-operation and Development, Nuclear Energy Agency, Paris, France

Nuclear Technology, 18 October 2024

Abstract — Reflood thermal hydraulics remains a difficult and complex subject, and understanding thephysical phenomena that occur during a reflood transient is important to nuclear safety. The Organisation forEconomic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Rod Bundle Heat Transfer(RBHT) project was designed to provide unique experimental data for code assessment and model development.Participants, which came from 21 international organizations, used analysis codes including APROS, ATHLET, CATHARE, CTF, MARS, RELAP5, TRACE, and SPACE to simulate the tests performed in the RBHT facility. The experimental campaign carried out within the OECD/NEA RBHT project produced data for a total of 16 reflood tests conducted in two test series. An “open” test series consisted of 11 experiments, and a “blind” test series consisted of 5 experiments. In the blind tests, only the initial and boundary conditions were providedto participants prior to simulation of those experiments. Reflood rates ranged from 0.5 to 15 cm/s, thusproducing data applicable to dispersed flow film boiling and inverted annular flow film boiling. Inlet subcoolingranged from 2.8 to 80 K. Tests with variable reflood rates and oscillatory reflood rates were included in the testmatrix. This paper describes the project and presents a summary of major experimental and analytical findings.

ANALYSES OF DESIGN EXTENSION CONDITIONS WITHOUT SIGNIFICANT FUEL DEGRADATION FOR OPERATING NUCLEAR POWER PLANTS: AN OECD/NEA REVIEW

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.