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The Atucha-2 Pressurized Heavy Water Reactor is equipped with a back-up shutdown system based on the fast injection of boron into the moderator tank. Such system had initially been designed to cope with a 10%-area (0.1A) break Loss Of Coolant Accident (LOCA) scenario, but based on upgraded licensing requirements the design had to be revised and possibly improved against a double-ended guillotine (2A) break LOCA. In particular, the boron injection had to be proven fast enough to allow a timely shutdown of the reactor, even in the case of a failure of the primary shutdown system (control rods). A full-scale test facility was built for such “design validation” purpose, in the framework of a cooperation program between the University of Pisa – San Piero a Grado Nuclear Research Group (GRNSPG) and the utility Nucleoelectrica Argentina S.A. (NA-SA). A special instrumentation system, based on conductivity probes designed on purpose by the Helmholtz Zentrum Dresden-Rossendorf (HZDR), was adopted for the measurement of the injection delay, as well as for the monitoring of pressure at several key locations. Care was taken to reproduce the relevant NPP conditions as closely as possible to those expected on the basis of extensive safety analyses performed adopting a Best Estimate Plus Uncertainty (BEPU) approach. In this respect, not only the test facility is full-scale, but also the key components (such as the fast opening air valves, the boric acid tanks, the rupture device, the injection lance) were directly borrowed from the Atucha-2 NPP. This paper provides an overview of the test facility, with particular emphasis on the Authors’ contributions to its design, implementation and operation. Then, it highlights the final outcomes of the experimental campaign carried out by NA-SA, namely: allowing to improve the design of the boron injection system (especially as to some fluid–structure interaction issues) and – what was the main goal – demonstrating that the system’s performance is fast enough to assure a timely and safe shutdown of the reactor, thus contributing to the successful completion of the NPP licensing process.

Atucha-2 is a Siemens-designed pressurized heavy water reactor located in Argentina. Its geometrical complexity and peculiarity [e.g., oblique control rods (CRs) and positive void coefficient] requ...

Atucha-2 is a Siemens-designed pressurized heavy water reactor in the Republic of Argentina. The correct prediction of the negative reactivity introduced in the moderator by an Emergency Boron Shutdown System (EBSS) is of great relevance for the correct safety evaluation of a double-ended guillotine large break LOCA scenario. During such event the EBSS is in charge to compensate the insertion of positive reactivity, caused by the void generated in the coolant channels by a sharp system pressure drop, in order to avoid severe core damage. The correct simulation of such event implies the minimization of the so called "numeric boron self-shielding effect" or the over-estimation of the inserted negative reactivity caused by the adoption of relatively large numerical meshes. In fact, because during the first phases of the injection, a very high concentrated boron solution is introduced in a small volume of the moderator tank, non-conservative reactivity estimation can be calculated if a "numeric boron dilution" is resulting by the adoption of large thermal-hydraulic and neutronic meshes. A methodology based on Monte Carlo transport code MCNP5 has been developed in order to predict power and reactivity excursions, representing a boron distribution in the moderator with different spatial resolutions. In such a way, it was possible to investigate the negative reactivity over-estimation due to the "boron self-shielding effect". This investigation is generally not possible by system codes used for safety analysis. The methodology was extended also to benchmark the EBSS negative reactivity insertion evaluated by the reference system code RELAP-3D©/NESTLE. Final results were used for the development of the Chapter 15 of the Final Safety Analysis Report of Atucha-2 NPP, demonstrating the conservative predictions of the reference system code employed for the safety evaluation. © 2015 Elsevier Ltd. All rights reserved.

Atucha-2 is a Siemens-designed Pressurized Heavy Water Reactor (PHWR) reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarity (e.g. oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank. The boron clouds reconstruction is obtained using a Computational Fluid Dynamics (CFD) CFX code calculation. A complete Large Break Loss Of Coolant Accident (LBLOCA) calculation implies the application of the RELAP5-3D © system code. Within the framework of the third Agreement “Nucleoelectrica Argentina-Sociedad Anonima (NA-SA) – University of Pisa/GRNSPG” ( Contract, 2009 ), a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions.

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis. This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.