1,684 research outputs found
Fonti per i materiali e le tecniche del disegno genovese a lapis, in (a cura di) M. C. Galassi e M. Priarone, L’underdrawing del disegno genovese. Dentro la genesi dell’opera grafica attraverso l’esame nell’infrarosso, Genova University Press, 2014, pp. 13-16
No full textINTELLIMAN. WP5 T5-3-1. Attention-Based Cloth Manipulation from Model-free Topological Representation
No full textThis dataset contains data related to a novel attention-based neural architecture capable of solving a smoothing task for deformable objects, such as clothes and fabric, by means of a single robotic arm. In particular, the dataset contains the data used for the training on a model-free learning-based policy for cloth smoothing task and the script used to collect the resulting data. The data were produced in the framework of Horizon Europe INTELLIMAN project and are presented in the publication:
K. Galassi, B. Wu, J. Perez, G. Palli and J. -M. Renders, "Attention-Based Cloth Manipulation from Model-free Topological Representation", 2024 IEEE International Conference on Robotics and Automation (ICRA), Yokohama, Japan, 2024, pp. 18207-18213, doi: 10.1109/ICRA57147.2024.10610241
Atucha II Nuclear Reactor: Design Safety and Licensing
No full textAtucha II is the third reactor built and operated in Argentina, following Atucha I and Embalse. This is the First-Of-A-Kind (FOAK) in the world notwithstanding Atucha I. The following main facts relate to Atucha II:
This is a natural uranium nuclear reactor (like a CANDU, i.e. Canadian Deuterium Uranium), vessel equipped (like a PWR, i.e. Pressurized Water Reactor) and with established safety features (namely, basic philosophy and protection and control systems) of latest German PWR.
It has a positive coolant void reactivity coefficient (fission power tends to increase if coolant is removed from the core): excluding one (standard) CANDU reactor built in China and CANDU-type reactors in India, this is the only reactor entered in operation with a such a feature after the Chernobyl accident in 1986.
A relationship establishes among Break Opening Time (BOT), fast Boron Injection Time (BIT), Peak Clad Temperature (PCT) and Fission Power Peak (FPP) following a loss of coolant accident. This puts a challenge to the safety of the Atucha II reactor. The availability of computational tools, including uncertainty evaluation, allowed the characterization of the relationship BOT-BIT-PCT-FPP: not necessarily, the highest PCT coincides with the highest FPP.
The construction of the reactor, started at the end of 1980 has stopped and finally restarted in 2006: all main components were stored at the site, but no industry (noticeably, not even the initial designer who supplied the components) was available for planning or supervising the construction.
The owner and operator of the Plant, Nucleoléctrica Argentina SA (NA-SA), created a single-purpose multidisciplinary team, headed by José Luis Antúnez (JLA) to undertake the completion of the construction and the start-up of Atucha II. The team completed the commissioning in due time.
A system encompassing the complexity of CANDU and PWR had to be built (in 2006) in a Country having a long-lasting history within nuclear technology. Furthermore, a competent regulator was ready to act based on the latest findings by the international community in terms of nuclear safety technology. Then, the utility recognized the need for a deep understanding of the design principles of the reactor beyond and over the capability to install the various components.
In order to accomplish the construction of Atucha II, international groups of experts formed to support the utility and the regulators, respectively. Those groups, under the guidance of regulator and utility expert-staff, confronted each other in relation to the critical issues of the system.
An unprecedented and unrepeatable endeavour took place: the utility leaded group accessed and reviewed the documentation of Atucha II (i.e. a few million pages) and confirmed, independently of the original designers, the validity of the design choices and the safety margins. The blue prints, the measurements taken on the site and the material properties were the only basis of the analyses performed by the latest computational tools
Atucha II Nuclear Reactor: Design, Safety and Licensing @2021
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available from the scientific community. All of this allowed the reconstruction of the steps of the reactor design without adopting any assumption or coefficient proposed by original designers: a pioneering application was completed of the Best Estimate Plus Uncertainty (BEPU) approach. The regulators timely assessed and, following deep discussions, endorsed the approach and the results.
The BEPU application, first of a kind to the analysis of all transients, part of FSAR Chapter 15, is the result of a technology grow-up in the area of accident investigations that started in the middle of 1970, i.e. at OECD/NEA/CSNI. Planning and analysis of complex experiments, code, code-user and nodalization qualification, accuracy quantification, scaling, uncertainty evaluation, code coupling, probabilistic safety assessment and licensing connection of thermal hydraulics are among the concerned topics.
Therefore, we deemed the description of that endeavour worthy for the present book. The current authors directed a few dozen scientists (see acknowledgments where not all of them are listed) contributing to the efforts during less than ten years to accomplish the mission proposed by JLA. Four parts constitute the book.
Part 1, Chapters 1 and 2 – Atucha II reactor description.
Part 2, Chapters 3 to 8 – The BEPU approach.
Part 3, Chapters 9 to 13 – The Large Break Loss of Coolant (LBLOCA) issue.
Part 4, Chapters 14 to 24 – Insights from Accident Analysis.
The Part 1 deals with a short history of the Atucha II project and provides key reactor features. The heavy water moderator and coolant fluids enter in contact into the vessel through proper bypass flow paths and ensure cooling of the fuel rods and moderation of neutrons at high and reduced average temperatures, respectively.
The Part 2 discusses the BEPU approach, i.e. a prerequisite for understanding the reactor features. The BEPU approach, established in nuclear thermal hydraulics, includes the application of qualified numerical codes and uncertainty evaluation procedures. A pioneering effort brought to adopting the approach for the entire Accident Analysis (AA), part of ‘Chapter 15’ part of the standard Final Safety Analysis Report (FSAR). This required a previous acceptance of the approach by regulators.
The Part 3 deals with the LBLOCA. Following the initiating event or the double-ended guillotine break, the hot fluid in the core, which behaves as a neutron absorber in nominal conditions, vaporizes starting at a few milliseconds. The moderator remains liquid for a few seconds and induces a positive reactivity input. A power excursion follows in a situation of degraded cooling (FPP).
The BEPU application to the envelope of transients addressed in AA, Part 4, implied the coupling of a wide variety of numerical codes in the areas of neutron physics, nuclear fuel performance, structural mechanics, fission products source term in the core and radiological impact on the environment, in addition to thermal hydraulics.
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A dozen papers dealing with the above topics have been published, e.g. Adorni et al., 2011; Araneo et al., 2011; D’Auria et al., 2012; D’Auria et al., 2012a; Pecchia et al., 2015; Petruzzi et al., 2016; Moretti et al., 2018; Mazzantini et al. 2019; D’Auria, 2019; D’Auria et al., 2019.
The papers by Adorni et al., 2011; Araneo et al. 2011; D’Auria et al., 2012; and Pecchia et al., 2015, deal with nuclear fuel, Pressurized Thermal Shock (PTS), Instrumentation and Control (I & C) and neutron physics modelling, respectively: related matter is part of the BEPU approach of Atucha II.
The paper by Moretti et al., 2018, deals with a scale-1 experiment performed to confirm the quality of the fast boron injection system of Atucha II: this is essential to mitigate the fission power excursion that is consequent to the LBLOCA.
The paper by Mazzantini et al., 2019, discusses various issues connected with the Large Break LOCA analysis.
The papers by D’Auria et al., 2012a; Petruzzi et al., 2016, and D’Auria, 2019 provide a summary of the BEPU use in licensing, including achievements and perspectives in the area. Finally, BEPU constituted the key element for a proposal aimed at creating a connection between safety of existing rectors and advancements in nuclear science and technology, paper by D’Auria et al., 2019.
We issued several hundred documents within the framework of the cooperation between NA-SA and University of Pisa. Three summary documents are GRNSPG, 2008; GRNSPG 2008a, and GRNSPG, 2010. Industry property information is involved.
All those papers and documents guided the planning of the present book where we avoided infringing the property rights. Thus, the idea here is not to replicate the contents of those papers and documents, with one noticeable exception that is the paper by Mazzantini et al., 2019: this constitutes the basis for issuing the Part 3 of the book
Methodology for the evaluation of the reliability of passive systems
No full textWithin the co-operation between ENEA and University of Pisa (Contract No. 9840, serie 3A, signed in 1998),
a synergic study including applications of PSA (Probabilistic Safety Assessment) and thermohydraulics to the design of
new reactor concept, is foreseen. An application-oriented project was proposed by ENEA. This is identified hereafter as
PSA-TH Project. In this framework a working group has been set-up and supported by ENEA, involving scientific
personnel at ENEA, at Polytechnic of Milan (Department of Nuclear Engineering – CESNEF) and at University of Pisa
(Department of Mechanical, Nuclear and Production Engineering - DIMNP). Specialists in safety assessment and in
thermohydraulics are part of the working group.
Following the first meeting ‘exploratory activities’ have been conducted by each of the three branches (i.e.
ENEA, CESNEF and DIMNP) of the working group, with the aim of identifying a common way for the study and of
finalizing the PSA-TH Project proposed by the ENEA.
The ‘exploratory activity’ performed at DIMNP is documented in ref. [1] (see also App. 4 of the present
report).
At the end of the year 1999, a draft procedure was agreed within the group. The procedure and the results from
its application are discussed in the report.
The attention was focused toward passive systems. The general objective of the activity was to characterize the
transient thermalhydraulic performance of the passive system in probabilistic terms. A natural circulation loop including
an Isolation Condenser and the Reactor Pressure Vessel where power production occurs, has been selected for the
analysis. The following steps of the procedure can be distinguished:
(a) Characterization of design/operational status for the system,
(b) Characterization of parameters that are critical for the operation of the system,
(c) Assigning probability values to the status and to the parameters from steps (a) and (b),
(d) Modeling of the system by a qualified thermal-hydraulic system code,
(e) Performing a best estimate calculation and assigning failure criteria for the system performance,
(f) Deterministic selection of sets of boundary and initial conditions (6 sets),
(g) Statistic selection of sets of boundary and initial conditions – discrete probability distribution (69 sets),
(h) Statistic selection of sets of boundary and initial conditions – continuous probability distribution (69 sets),
(i) Association of each set to a code run (each set is also characterized by a probability value) and execution of 144
code runs (6+69+69),
(j) Based upon failure criteria defined under item (e) and upon results of code calculations, item (i), system
performance indicators could be derived (these consisted of ensembles of tables and plots). The deterministic
selection, item (f), combined with each of the statistic selections, items (g) or (h), causes the achievement of two
ensembles of system performance indicators.
‘Curves of merit’ for the system performance are significant results achieved from the application of the
procedure. These are assumed to be characteristic of the selected thermalhydraulic system and can be used to judge the
system acceptability and to compare the selected system with different systems
IgM MGUS anti myelin-associated glycoprotein neuropathy can rarely express as a predominantly distal motor neuropathy
No full textResults of RELAP4/MOD6 code applications", Lectures L8, L9, L10, L13, L16 at Course on Thermal-hydraulic Phenomena in Nuclear Reactor Technology - Sofia (BG),
No full textA Course in Nuclear thermal-hydraulics was organized by UNIPI in Sofia at the time of the cold war (Soviet Union collapsed in 1991). Contacts crossing the iron curtain were extremely complex. The entire Course consists of several hundred slides (all preserved in paper format by the corresponding author) and a couple dozen lectures (see copies below). Two pages from the Course are reported below. The current lecture deals with the selected topics relevant in nuclear thermal-hydraulics –more details can be found in the copied program below
"Analysis of natural circulation test A2-77 performed in LOBI facility, by RELAP5/MOD2 code"
No full textThe document deals with the analysis of an experiment to be performed in LOBI ITF in the configuration Mode2. LOBI was a Pressurized Water Reactor (PWR) simulator installed in the EC center of Ispra Italy financially supported by EURATOM.
The configuration Mod2 of LOBI was designed as a consequence of the Three Mile accident in US and is the follow-up of the configuration Mod1. Mod2 was designed to simulate stratification phenomena and natural circulation which are important during Small Break Loss of Coolant Accidents (SBLOCA); as a difference Mod1 was designed to simulate phenomena expected during Large Break Loss of Coolant (LBLOCA) scenarios. Namely downcomer size in Mod2 was much smaller than in the case of Mod1 and a large number of Emergency Core Cooling Systems (ECCS) were installed in Mod2. The concerned A2-77 Natural Circulation (NC) experiment was performed to model as far as possible the PWR performance. Various phase of NC were identified and brought to important findings in Nuclear Reactor Safety, including the characterization of instabilities and the reflux condensing mode.
The work was discussed in a specific meeting in Ispra of the so-called LPTF research group where University of Pisa represented Italy
Accuracy quantification by the FFT method in FARO L-14 (ISP 39) open calculations, Ispra (I), April 22-23, 1997
No full textThe Fast Fourier Transform Based Method (FFTBM) was developed at University of Pisa to achieve a quantitative evaluation of the accuracy of thermal-hydraulic system code calculations. International cooperation was established to transfer the method with various Institutions all over the world. In the case documented in this report one scientists from University of Pisa (Mario Leonardi) has been invited at the European Commission (EC) Joint Research Center (JRC) of Ispra (Varese, Italy) to implement and to apply the method. The present document has been issued by JSI and describes the application of the method to the severe accident (corium interaction with water) FARO L-14 carried out in the framework of EC EURATOM researches. The L-14 experiment was selected at the basis of International Standard Problem 39 (ISP-39) by the Organization for Economic Cooperation and Development / Nuclear Energy Agency / Committee on the Safety of Nuclear Installations (OECD/NEA/CSNI)
Best Estimate Analysis and Uncertainty evaluation of the Angra-2 LBLOCA DBA
No full textThe licensing is a legal process to transfer the nuclear safety technology (developed and supported by researches) into a legal-fixed environment. Competences in the legal nuclear safety framework (typically part of the Atomic Act in each Country) and of nuclear reactor safety are needed in such a context.
University of Pisa had the opportunity and the challenge to participate, on the behalf of the local Regulatory body (a Governmental Institution) in Brazil, to the review of the safety report submitted by the German Industry (namely, KWU-Siemens) for the licensing of the Angra-2, 1400 Mwe (the largest Nuclear Power plant in the world in terms of reactor power) o Nuclear Power Plant.
The present document deals with the independent analysis of the Large Break Loss of Coolant Accident (LBLOCA) performed at University of Pisa and to be compared with the results for the same accident obtained by the industry
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