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Francesco Saverio Dauria - One of the best experts on this subject based on the ideXlab platform.
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strengthening nuclear Reactor Safety and analysis
Nuclear Engineering and Design, 2017Co-Authors: Francesco Saverio Dauria, Nenad Debrecin, H GlaeserAbstract:Abstract The paper aims at fixing bases for possible strengthening of current Nuclear Reactor Safety (NRS) and Safety analysis: this is done by combining the logical frameworks connected with the terms As-Low-As-Reasonably-Achievable (ALARA), Best-Estimate-Plus-Uncertainty (BEPU), Extended-Safety-Margin (E-SM) and Independent-Assessment (IA). ALARA is an early principle in Nuclear Reactor Safety: designers and operators must do their best to minimize doses to the humans. BEPU is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the Safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margins constitute an established concept in NRS: a few dozen SM values must be calculated in current Safety analyses and demonstrated to be acceptable. The concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit, thus forming the E-SM. Here ‘the environment’ includes the personnel in charge of activities connected with the NPP. The E-SM implies the formulation of some ten-thousands Safety margins values, which shall correspond to a similar number of monitored variables. IA is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design. The implementation of the idea in the paper brings to an additional Safety barrier for existing and future nuclear Reactors which may reduce the probability of core melt, presumably at an affordable cost for the industry.
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preservation and use of integral system test facilities data the experience of the lobi data and the stresa database
Progress in Nuclear Energy, 2012Co-Authors: P Pla, A Annunziato, C Addabbo, G M Galassi, Francesco Saverio DauriaAbstract:Abstract Experimental data recorded in Integral Effect Test Facilities (ITFs) are traditionally used in order to validate Best Estimate (BE) system codes and to investigate the behaviour of Nuclear Power Plants (NPP) under accident scenarios. The extent to which the existing Reactor Safety experimental databases are preserved was well known and frequently debated and questioned in the nuclear community. The Joint Research Centre (JRC) of the European Commission (EC) has been deeply involved during years in several projects for experimental data production and experimental data preservation; in particular a big initiative was the LOBI ITF project. In this context the STRESA (Storage of Thermal Reactor Safety Analysis Data) web-based informatic platform was initially planned by JRC-Ispra with the main objective to disseminate documents and experimental data from large in-house JRC scientific projects, as LOBI ITF data, and later it was extensively used in order to provide a secure repository of ITF data exploiting modern computer information technologies for access and retrieve of the information. The paper is focused in presenting one of the largest EC initiatives on the production of ITF data (the LOBI project), its use for system thermal hydraulic code assessment and its storage in the JRC STRESA node web platform ( http://stresa.jrc.ec.europa.eu/stresa/ ). The objective of the paper is to further disseminate and promote the usage of the database containing these LOBI ITF data and to demonstrate long-term importance of well maintained ITF databases. At present the JRC STRESA database is maintained by JRC-Petten.
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thermal hydraulic system codes in nulcear Reactor Safety and qualification procedures
Science and Technology of Nuclear Installations, 2008Co-Authors: A Petruzzi, Francesco Saverio DauriaAbstract:In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.
János Varga - One of the best experts on this subject based on the ideXlab platform.
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TELLER Edéről mondták – pályatársak, barátok, ellenségek véleménye
Erdélyi Magyar Műszaki Tudományos Társaság – EMT Hungarian Technical Scientific Society of Transylvania Societatea Maghiară Tehnico-Ştiinţifică din Tr, 2020Co-Authors: János VargaAbstract:Edward Teller, the Hungarian-born American theoretical physicist is known as „the father of the hydrogen bomb”; he liked to be called "the father of Reactor Safety" or "the father of two children". He made numerous contributions to nuclear physics, chemistry, quantum chemistry, Reactor physics, Reactor Safety, atomic and molecular physics, solid state physics, and space and plasma physics, statistical physics and many other disciplines. Based on his achievements he is rightly called one of the major scholars of the last century. His contradictory personality, arrogant attitude also gained him a lot of opponents. This collection of quotes gives you the views of his friends, enemies, and from those you get a glimpse of his colourful personality. Rezumat Ede Teller, fizicianul teoretic american de origine maghiară este cunoscut ca „tatăl bombei de hidrogen”; dar el poate fi denumit drept „tatăl securităţii reactoarelor nucleare” sau „tatăl a doi copii”. El a adus numeroase contribuţii la fizica nucleară, chimie, chimie cuantică, fizica reactoarelor, securitatea reactoarelor, la fizica atomică şi nucleară, fizica corpurilor solide, precum şi la fizica spaţială, fizica plasmei, fizica statistică şi la multe alte discipline. Pe baza acestei activităţi el poate fi pe drept numit, ca unul din cei mai mari erudiţi ai secolului trecut. Personalitatea ei contradictorie, atitudinea ei arogantă a provocat apariţia şi a multor adversari. Prezenta colecţie de citate oferă o privire asupra prietenilor, adversarilor precum şi asupra personalităţii lui deosebit de colorate
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TELLER Edéről mondták – pályatársak, barátok, ellenségek véleménye: It was Told about Edward TELLER – Opinion of Career Couples, Friends, Enemies / Ce s-a spus despre Ede TELLER – părerile tovarăşilor de carieră, a prietenilor şi adversarilor lui
Erdélyi Magyar Műszaki Tudományos Társaság – EMT Hungarian Technical Scientific Society of Transylvania Societatea Maghiară Tehnico-Ştiinţifică din Tr, 2020Co-Authors: János VargaAbstract:Edward Teller, the Hungarian-born American theoretical physicist is known as „the father of the hydrogen bomb”; he liked to be called "the father of Reactor Safety" or "the father of two children". He made numerous contributions to nuclear physics, chemistry, quantum chemistry, Reactor physics, Reactor Safety, atomic and molecular physics, solid state physics, and space and plasma physics, statistical physics and many other disciplines. Based on his achievements he is rightly called one of the major scholars of the last century. His contradictory personality, arrogant attitude also gained him a lot of opponents. This collection of quotes gives you the views of his friends, enemies, and from those you get a glimpse of his colourful personality. Rezumat Ede Teller, fizicianul teoretic american de origine maghiară este cunoscut ca „tatăl bombei de hidrogen”; dar el poate fi denumit drept „tatăl securităţii reactoarelor nucleare” sau „tatăl a doi copii”. El a adus numeroase contribuţii la fizica nucleară, chimie, chimie cuantică, fizica reactoarelor, securitatea reactoarelor, la fizica atomică şi nucleară, fizica corpurilor solide, precum şi la fizica spaţială, fizica plasmei, fizica statistică şi la multe alte discipline. Pe baza acestei activităţi el poate fi pe drept numit, ca unul din cei mai mari erudiţi ai secolului trecut. Personalitatea ei contradictorie, atitudinea ei arogantă a provocat apariţia şi a multor adversari. Prezenta colecţie de citate oferă o privire asupra prietenilor, adversarilor precum şi asupra personalităţii lui deosebit de colorate
Staffan Qvist - One of the best experts on this subject based on the ideXlab platform.
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an autonomous reactivity control system for improved fast Reactor Safety
Progress in Nuclear Energy, 2014Co-Authors: Staffan Qvist, Ehud GreenspanAbstract:Abstract The Autonomous Reactivity Control (ARC) system is a new Safety device that can passively provide negative reactivity feedback in fast Reactors that is sufficient to compensate for the positive coolant density reactivity feedback even in large low-leakage cores. The ARC system is actuated by the inherent physical property of thermal expansion, and has a very small effect on core neutronics at standard operating conditions. Additionally, the ARC system does not have an identified failure mode that can introduce positive reactivity in to the core. An ARC system can be installed in conventional fuel assemblies by replacing a limited number of fuel rods with rods that fill a Safety function, providing negative reactivity to the core in the event of coolant temperature rise above nominal. These rods are of the same outer dimensions as the fuel rods, but contain smaller-diameter inner rods that are connected to liquid-filled reservoirs at the top and bottom of the assemblies. The reservoirs are filled with two separate liquids that stay liquid and immiscible throughout the applicable temperature range of fast Reactor operation. The lower reservoir contains a “neutron poison” liquid with a high neutron absorption cross-section. The upper reservoir is filled with a separate liquid with a small neutron absorption cross-section. As the temperature in the assembly increases, the liquids in the reservoirs thermally expand, effectively pushing the absorbing liquid up toward the active core region while compressing the inert gas that fills the volume above the liquid between the inner and outer tubes of the ARC rods. The ARC system can be installed, or retrofitted in to existing systems, in every fuel assembly in the core. Since ARC installations in individual fuel assemblies operate independently, the system has a high level of redundancy. ARC-systems respond to local transients as well as core-wide accident scenarios. After actuation, the system automatically returns to its initial state as temperatures decrease, without the need for intervention by Reactor operators. The ARC system concept and design considerations are described and illustrated.
William L Luyben - One of the best experts on this subject based on the ideXlab platform.
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use of dynamic simulation for Reactor Safety analysis
Computers & Chemical Engineering, 2012Co-Authors: William L LuybenAbstract:Abstract Dynamic simulations of chemical processes are widely used to develop effective plantwide control structures that provide stable regulatory control at some desired operating condition. This paper illustrates that they can also serve a very useful role in the analysis of Safety problems in the event of emergency situations. The dynamic response of the process when various failures occur is critical to the design of Safety systems for the process (alarms, overrides, interlocks, Safety valves and rupture disks). For example, a failure of the supply of cooling water will lead to rapid increases in pressures and temperatures in the process. Determining the rates of increase in these important variables and the time period to reach critical limits ( Safety response time ) permits the engineer to quantitatively design effective Safety systems. Chemical Reactors are typically the most sensitive and potentially the most dangerous units in many processes, particularly when exothermic reactions and low per-pass reactant conversions are involved. This paper illustrates how Aspen Dynamic simulation can be used for predicting the dynamic changes in critical variables. Dynamic emergency Safety simulations are presented for five processes with several types of cooled Reactors (CSTR and tubular) and residence times varying from 0.16 to 60 min. Safety response times vary from several seconds to several minutes, depending on both the Reactor, the system in which it is installed and the level of reactant conversion.
Ehud Greenspan - One of the best experts on this subject based on the ideXlab platform.
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an autonomous reactivity control system for improved fast Reactor Safety
Progress in Nuclear Energy, 2014Co-Authors: Staffan Qvist, Ehud GreenspanAbstract:Abstract The Autonomous Reactivity Control (ARC) system is a new Safety device that can passively provide negative reactivity feedback in fast Reactors that is sufficient to compensate for the positive coolant density reactivity feedback even in large low-leakage cores. The ARC system is actuated by the inherent physical property of thermal expansion, and has a very small effect on core neutronics at standard operating conditions. Additionally, the ARC system does not have an identified failure mode that can introduce positive reactivity in to the core. An ARC system can be installed in conventional fuel assemblies by replacing a limited number of fuel rods with rods that fill a Safety function, providing negative reactivity to the core in the event of coolant temperature rise above nominal. These rods are of the same outer dimensions as the fuel rods, but contain smaller-diameter inner rods that are connected to liquid-filled reservoirs at the top and bottom of the assemblies. The reservoirs are filled with two separate liquids that stay liquid and immiscible throughout the applicable temperature range of fast Reactor operation. The lower reservoir contains a “neutron poison” liquid with a high neutron absorption cross-section. The upper reservoir is filled with a separate liquid with a small neutron absorption cross-section. As the temperature in the assembly increases, the liquids in the reservoirs thermally expand, effectively pushing the absorbing liquid up toward the active core region while compressing the inert gas that fills the volume above the liquid between the inner and outer tubes of the ARC rods. The ARC system can be installed, or retrofitted in to existing systems, in every fuel assembly in the core. Since ARC installations in individual fuel assemblies operate independently, the system has a high level of redundancy. ARC-systems respond to local transients as well as core-wide accident scenarios. After actuation, the system automatically returns to its initial state as temperatures decrease, without the need for intervention by Reactor operators. The ARC system concept and design considerations are described and illustrated.
