2018

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Uncertainty calculation in small break LOCA in the emergency core cooling system connected to the hot leg of Angra 2 nuclear power plant

 

 

Eduardo Madeira Borgesa, Gaianê Sabundjiana, F. D'Auriab, A. Petruzzic


a
Instituto de Pesquisas Energéticas e Nucleares (IPEN-CNEN-SP), Av. Professor Lineu Prestes, 2242, São Paulo, Brazil

bUniversity of Pisa, Largo L. Lazzarino 2, Pisa, Italy
c
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy


Nuclear Energy Science and Technology, Vol. 12, No. 2, 2018


Abstract — 
Owing to the occurrence of nuclear accidents, worldwide nuclear regulatory organisations included the analysis of accidents considered as design basis accidents – Loss of Coolant Accident (large and small-break, LBLOCA or SBLOCA) – in the safety analysis reports of nuclear facilities. In Brazil, the tool selected by the licensing authority, Comissão Nacional de Energia Nuclear (CNEN), is RELAP5 Code. The aim of this paper is the evaluation of the performance of the Emergency Core Cooling System (ECCS) of Angra 2 nuclear reactor during SBLOCA. In this study, the RELAP5 code and the Code Internal Assessment of Uncertainty (CIAU) were used to simulate and analyse the uncertainties of the results. The postulated accident is the SBLOCA in the hot leg connected to the ECCS described in the Final Safety Analysis Report of Angra 2 (FSAR/A2). The results from this study were satisfactory when compared with the FSAR/A2. 


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Uncertainty and sensitivity analysis in reactivity-initiated accident fuel modeling: synthesis of organisation for economic co-operation and development (OECD)/nuclear energy agency (NEA) benchmark on reactivity-initiated accident codes phase-II

 

Olivier Marchanda, Jinzhao Zhang b, Marco Cherubinic 


a
Institut de Radioprotection et de Sûrete Nucleaire (IRSN), PSN-RES, SEMIA, Cadarache, St Paul-Lez-Durance, 13115, France

bTractebel (ENGIE), Avenue Ariane 7, 1200 Brussels, Belgium
c
Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

 

Nuclear Energy Science and Technology, 50 (2018) 280-291


Abstract — 
In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010e2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made. © 2018 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the 

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Atucha-1 NPP containment venting analysis following SBO and LBLOCA events by GOTHIC code

 

 

 

A. Popa,b, W. Giannottia, A. Petruzzia , R. Garberoc, O. Mazzantinic


a
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bUniversity of Pisa, Largo L. Lazzarino 2, Pisa, Italy
c
NA-SA, Nucleoelectrica Argentina, Buonos Aires, Argentina

 

Nuclear Engineering and Design, October 2018


Abstract — 
Containment behaviour plays a key role in the safety framework of a Nuclear Power Plant (NPP). The GOTHIC thermal hydraulic code has been adopted to evaluate the Atucha-1 NPP containment responses during two postulated severe accident scenarios, Station Black Out and Large Break Loss of Coolant Accident without Safety Injection Pumps (SIPs), while assuming that the external cooling of the Reactor Pressure Vessel is carried out during the transients.

The Atucha-1 NPP has a containment designed to work at full pressure, constituted by a steel sphere enveloped by a concrete shell, and having an annular gap of air in between.

The target of the analysis is the evaluation of the effects caused by the additional production of steam in the reactor cavity as a consequence of the external vessel cooling, which could cause an increase in containment pressure, and lead to pressure values above the safety limit. The containment pressure and temperature, the distribution of hydrogen in the containment atmosphere and the water hold-up in the most relevant rooms have been analysed as target variables. Each accident scenario was simulated using two different nodalizations, characterized by a different level of refinement. The “detailed” nodalization is meant to be the most refined nodalization according to the available computational resources; having high fidelity three dimensional details, with a high number of cells. Taking into consideration that several sensitivities were performed, the “coarse” nodalization was developed in order to lower the demand for computational resources without significantly compromising the global scenario response. Both nodalizations are characterized by high complexity in the representation of rooms and their connections.

Both accident transients, for each type of nodalization, were simulated for 200,000 s. At the end of the simulated transient, results showed that for the Large Break Loss of Coolant Accident pressure is predicted to surpass 5 bar, while the Station Black Out scenario is calculated to reach 4.4 bar. The performed sensitivities were simulated for 100,000 s and were meant to understand and characterize the impact of the different nodalization parameters (geometrical aspects, material properties, BCs). In addition, due to several code anomalies identified, several other sensitivity calculations were performed in order to find a way to analyse and mitigate the issues.


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Instrumenting full-scale Boron Injection Test Facility to support Atucha-2 NPP licensing

 

 

 

F. Morettia, F. Terzuolia, F. D’Auria,b O. Mazzantinic


a
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bGRNSPG—University of Pisa, Via Livornese 1219, San Piero a Grado (PI), Pisa, Italy
c
NA-SA, Nucleoelectrica Argentina S.A., UG-CNAII, 2806 Lima, Argentina

 

Nuclear Engineering and Design, February 2018

 

Abstract —  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 Nucleoeléctrica 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.