122 research outputs found
Analyses of the OSU-MASLWR Experimental Test Facility
Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well
Analysis of an unmitigated 2-inch cold leg LOCA transient with ASTEC and MELCOR codes
The analyses of postulated severe accident sequences play a key role for the
international nuclear technical scientific community for the study of the effect of possible actions
to prevent significant core degradation and mitigate source term release. To simulate the
complexity of phenomena involved in a severe accident, computational tools, known as severe
accident codes, have been developed in the last decades. In the framework of NUGENIA TA-2
ASCOM project, the analysis of an unmitigated 2-inch cold leg LOCA transient, occurring in a
generic western three-loops PWR-900 MWe, has been carried out with the aim to give some
insights on the modelling capabilities of these tools and to characterize the differences in the
calculations results. The ASTEC V2.2b code (study carried out with ASTEC V2, IRSN all rights
reserved, [2021]), and MELCOR 2.2 code have been used in this code-to-code benchmark
exercise. In the postulated transient, the unavailability of all active injection coolant systems has
been considered and only the injection of accumulators has been assumed as accident mitigation
strategy
Analysis of Primary/Containment Coupling Phenomena Characterizing the MASLWR Design During a SBLOCA Scenario
Today considering the world energy demand increase, the use of advanced nuclear power
plants, have an important role in the environment and economic sustainability of country
energy strategy mix considering the capacity of nuclear reactors of producing energy in safe
and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World
Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al.,
2011d). According to the information’s provided by the “Power Reactor Information
System” of the International Atomic Energy Agency (IAEA), today 433 nuclear power
reactors are in operation in the world providing a total power installed capacity of 366.610
GWe, 5 nuclear reactors are in long term shutdown and 65 units are under construction
(IAEA PRIS, 2011).
In the last 20 years, the international community, taking into account the operational
experience of the nuclear reactors, starts the development of new advanced reactor designs,
to satisfy the demands of the people to improve the safety of nuclear power plants and the
demands of the utilities to improve the economic efficiency and reduce the capital costs
(D'Auria et al., 1993; Mascari et al., 2011c). Design simplifications and increased design
margins are included in the advanced Light Water Reactors (LWR) (Aksan, 2005). In this
framework, the project of some advanced reactors considers the use of emergency systems
based entirely on natural circulation for the removal of the decay power in transient
condition and in some reactors for the removal of core power during normal operating
conditions (IAEA-TECDOC-1624, 2009; Mascari et al., 2010a; Mascari et al., 2011d). For
example, if the normal heat sink is not available, the decay heat can be removed by using a
passive connection between the primary system and heat exchangers (Aksan, 2005; Mascari
et al., 2010a, Mascari, 2010b). The AP600/1000 (Advanced Plant 600/1000 MWe) design, for example, includes a Passive Residual Heat Removal (PRHR) system consisting of a C-Tube
type heat exchanger immersed in the In-containment Refueling Water Storage Tank
(IRWST) and connected to one of the Hot Legs (HL) (IAEA-TECDOC-1391, 2004; Reyes,
2005c; Gou et al., 2009; Mascari et al., 2010a). A PRHR from the core via Steam Generators
(SG) to the atmosphere, considered in the WWER-1000/V-392 (Water Moderated, Water
Cooled Energy Reactor) design, consists of heat exchangers cooled by atmospheric air, while
the PRHR via SGs, considered in the WWER-640/V-407 design, consists of heat exchangers
immersed in emergency heat removal tanks installed outside the containment (Kurakov et
al., 2002; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a). In the AC-600
(Advanced Chinese PWR) the PRHR heat exchangers are cooled by atmospheric air (IAEATECDOC
1281, 2002; Zejun et al., 2003; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari
et al., 2010a) and in the System Integrated Modular Advanced Reactor (SMART) the PRHR
heat exchangers are submerged in an in-containment refuelling water tank (IAEA-TECDOC-
1391, 2004; Lee & Kim, 2008; Gou et al., 2009; Mascari et al., 2010a). The International
Reactor Innovative and Secure (IRIS) design includes a passive Emergency Heat Removal
System (EHRS) consisting of an heat exchanger immersed in the Refueling Water Storage
Tank (RWST). The EHRS is connected to a separate SG feed and steam line and the RWST is
installed outside the containment structure (Carelli et al., 2004; Carelli et al., 2009; Mascari,
2010b; Chiovaro et al., 2011). In the advanced BWR designs the core water evaporates,
removing the core decay heat, and condenses in a heat exchanger placed in a pool. Then the
condensate comes back to the core (Hicken & Jaegers, 2002; Mascari et al., 2010a). For
example, the SWR-1000 (Siede Wasser Reaktor, 1000 MWe) design has emergency
condensers immersed in a core flooding pool and connected to the core, while the ESBWR
(Economic Simplified Boiling Water Reactor) design uses isolation condensers connected to
the Reactor Pressure Vessel (RPV) and immersed in external pools (IAEA-TECDOC-1391,
2004; Aksan, 2005; Mascari et al., 2010a).
The designs of some advanced reactors rely on natural circulation for the removing of the
core power during normal operation. Examples of these reactors are the MASLWR (Multi-
Application Small Light Water Reactor), the ESBWR, the SMART and the Natural
Circulation based PWR being developed in Argentina (CAREM)(IAEA-TECDOC-1391, 2004;
IAEA -TECDOC-1474, 2005; Mascari et al., 2010a). In particular the MASLWR (Modro et al.,
2003), figure 1, is a small modular integral Pressurized Water Reactor (PWR) relying on
natural circulation during both steady-state and transient operation.
In the development process of these advanced nuclear reactors, the analysis of single and
two-phase fluid natural circulation in complex systems (Zuber, 1991; Levy, 1999; Reyes &
King, 2003; IAEA-TECDOC-1474, 2005; Mascari et al., 2011e), under steady state and
transient conditions, is crucial for the understanding of the physical and operational
phenomena typical of these advanced designs. The use of experimental facilities is
fundamental in order to characterize the thermal hydraulics of these phenomena and to
develop an experimental database useful for the validation of the computational tools
necessary for the operation, design and safety analysis of nuclear reactors. In general it is
expensive to design a test facility to develop experimental data useful for the analyses of
complex system, therefore reduced scaled test facilities are, in general, used to characterize
them. Since the experimental data produced have to be applicable to the full-scale
prototype, the geometrical characteristics of the facility and the initial and boundary conditions of the selected tests have to be correctly scaled. Since possible scaling distortions
are present in the experimental facility design, the similitude of the main thermal hydraulic
phenomena of interest has to be assured permitting their accurate experimental simulation
(Zuber, 1991; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e).
Fig. 1. MASLWR conceptual design layout (Modro et al, 2003; Reyes et al., 2007; Mascari et
al., 2011a).
Different computer codes have been developed to characterize two-phase flow systems,
from a system and a local point of view. Accurate simulation of transient system behavior of
a nuclear power plant or of an experimental test facility is the goal of the best estimate
thermal hydraulic system code. The evaluation of a thermal hydraulic system code’s
calculation accuracy is accomplished by assessment and validation against appropriate
system thermal hydraulic data, developed either from a running system prototype or from a
scaled model test facility, and characterizing the thermal hydraulic phenomena during both
steady state and transient conditions. The identification and characterization of the relevant
thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic
systems codes, has been the objective of multiple international research programs (Mascari
et al., 2011a; Mascari et al., 2011c).
In this international framework, Oregon State University (OSU) has constructed, under a
U.S. Department of Energy grant, a system level test facility to examine natural circulation
phenomena of importance to the MASLWR design. The scaling analysis of the OSUMASLWR
experimental facility was performed in order to have an adequately simulation of
the single and two-phase natural circulation, reactor system depressurization during a
blowdown and the containment pressure response typical of the MASLWR prototype
(Zuber, 1991; Reyes & King, 2003; Reyes, 2005b). A previous testing program has been conducted in order to assess the operation of the prototypical MASLWR under normal full
pressure and full temperature conditions and to assess the passive safety systems under
transient conditions (Modro et al. 2003; Reyes & King, 2003; Reyes, 2005b; Reyes et al., 2007;
Mascari et al., 2011e). The experimental data developed are useful also for the assessment
and validation of the computational tools necessary for the operation, design and safety
analysis of nuclear reactors.
For many years, in order to analyze the LWR reactors, the USNRC has maintained four
thermal-hydraulic codes of similar, but not identical, capabilities, the RAMONA, RELAP5,
TRAC-B and TRAC-P. In the last years, the USNRC is developing an advanced best estimate
thermal hydraulic system code called TRAC/RELAP Advanced Computational Engine or
TRACE, by merging the capabilities of these previous codes, into a single code (Boyac &
Ward, 2000; TRACE V5.0, 2010; Reyes, 2005a; Mascari et al., 2011a). The validation and
assessment of the TRACE code against the MASLWR natural circulation database,
developed in the OSU-MASLWR test facility, is a novel effort.
This chapter illustrates an analysis of the primary/containment coupling phenomena
characterizing the MASLWR design mitigation strategy during a SBLOCA scenario and, in
the framework of the performance assessment and validation of thermal hydraulic system
codes, a qualitative analysis of the TRACE V5 code capability in reproducing it
Analysis of the OSU-MASLWR 001 and 002 tests by using the TRACE code
The Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a scaled integral test facility to examine natural circulation phenomena characterizing the Multi-Application Small Light Water Reactor (MASLWR) design. The MASLWR is a small modular PWR relying on natural circulation during both steady-state and transient operation, which includes an integrated helical coil steam generator within the reactor pressure vessel. Testing has been conducted in order to assess the operation of the prototypical MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems performance. The experimental data produced are useful also for the assessment of the computational tools necessary for the operation, design and safety analysis of nuclear reactors.
This report describes the assessment of TRACE code predictions, conducted under the NRC CAMP program, against the MASLWR tests OSU-MASLWR-001 and the OSU-MASLWR-002, respectively. This activity has been conducted in collaboration with the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA), the Department of Energy of the University of Palermo, the Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG) of University of Pisa, the Department of Nuclear Engineering and Radiation Health Physics at OSU and NuScale Power Inc. In particular the OSU-MASLWR-001 test, an inadvertent actuation of one submerged ADS valve, investigates the primary system to containment coupling under design basis accident conditions; the OSU-MASLWR-002 test, a natural circulation test, investigates the primary system flow rates and secondary side steam superheat for a variety of core power levels and feed water flow rates. The assessment against experimental data shows that the TRACE code predicts the main phenomena of interest of the selected tests reasonably well for most condition
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Analyses of the OSU-MASLWRExperimental Test Facility
Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well
CLORURO DE BENZALCONIO EN EL TRATAMIENTO DE LA DERMATOMICOSIS CAUSADA POR Trichophyton sp. EN EL CUY (Cavia cobayo)
Se evaluó la eficacia del cloruro de benzalconio en el tratamiento de la dermatomicosis causada por Trichophyton sp. en el cuy (Cavia cobayo). Se utilizó 200 cuyes de 2 meses de edad que presentaban signos clínicos de dermatomicosis. Los cuyes fueron divididos al azar en dos grupos (A y B) similares. Al grupo A se le aplicó una solución acuosa de cloruro de benzalconio al 0.1% como único tratamiento y en el grupo B se empleó agua como placebo. Previo al tratamiento, se tomaron muestras de pelos y escamas de piel para examen directo y cultivo en los medios Dermatophytes Test Medium (DTM) y Mycobiotic, así como biopsias de piel para estudio histológico. Todas las muestras resultaron positivas a hongos entre el 3º y 7º día, tipificándose la especie Trichophyton mentagrophytes con la ayuda de microcultivos y lectura en agar urea. El 60% de las muestras fueron positivas al examen directo. El estudio histológico mostró hiperqueratosis ligera. El 100% de los cuyes del grupo A mostró curación clínica con crecimiento de pelo al finalizar la 4a semana post-tratamiento. Se concluye que una sola aplicación de cloruro de benzalconio al 0.1% en solución es eficaz para el tratamiento topical de la dermatomicosis causada por Trichophyton mentagrophytes; lográndose una eficacia del 70% en la tercera semana post-tratamiento, sin que cause daño a la piel y sin ser afectada por la presencia de materia orgánica.The efficacy of benzalkonium chloride for the treatment of dermatomicosis caused by Trichophyton sp. in guinea pig (Cavia cobayo) was evaluated in 200 guinea pigs of two months old, that showed typical clinical signs of dermatomicosis. Hair samples and skin grudges were taken for direct exam and for in vitro culture in Dermatophytes Test Medium and Mycobiotic. Skin biopsies for pathobiology were prepared. The guinea pig were divided in groups A and B of 100 animals each. Group A was treated with 0.1% aqueous benzalkonium chloride solution, whereas group B was treated with water as placebo. All samples were positive to media culture by 3rd to 7th day and Trichophyton mentagrophytes was isolated. Sixty percent of the samples were positive to direct exam. Histological study showed mild hyperkeratosis. Healing of lesions and hair growth were observed in all treated guinea pigs by the end of 4th week. The study showed that a single application of 0.1% aqueous benzalkonium chloride solution was effective for topical treatment on guinea pig dermatomycosis
Estimation of the length of interactions in arena game semantics
We estimate the maximal length of interactions between strategies in HO/N
game semantics, in the spirit of the work by Schwichtenberg and Beckmann for
the length of reduction in simply typed lambdacalculus. Because of the
operational content of game semantics, the bounds presented here also apply to
head linear reduction on lambda-terms and to the execution of programs by
abstract machines (PAM/KAM), including in presence of computational effects
such as non-determinism or ground type references. The proof proceeds by
extracting from the games model a combinatorial rewriting rule on trees of
natural numbers, which can then be analyzed independently of game semantics or
lambda-calculus.Comment: Foundations of Software Science and Computational Structures 14th
International Conference, FOSSACS 2011, Saarbr\"ucken : Germany (2011
Current status of MELCOR 2.2 for fusion safety analyses
MELCOR is an integral code developed by Sandia National Laboratories (SNL) for the US Nuclear Regulatory Commission (USNRC) to perform severe accident analyses of Light Water Reactors (LWR). More recently, MELCOR capabilities are being extended also to analyze non-LWR fission technologies. Within the European MELCOR User Group (EMUG), organized in the framework of USNRC Cooperative Severe Accident Research Program (CSARP), an activity on the evaluation of the applicability of MELCOR 2.2 for fusion safety analyses has been launched and it has been coordinated by ENEA. The aim of the activity was to identify the physical models to be possibly implemented in MELCOR 2.2 necessary for fusion safety analyses, and to check if those models are already available in MELCOR 1.8.6 for fusion version, developed by Idaho National Laboratory (INL). From this activity, a list of modeling needs emerged from the safety analyses of fusion-related installations have been identified and described. Then, the importance of the various needs, intended as the priority for model implementation in the MELCOR 2.2 code, has been evaluated according to the technical expert judgement of the authors. In the present paper, the identified modeling needs are discussed. The ultimate goal would be to propose to have a single integrated MELCOR 2.2 code release capable to cover both fission and fusion applications
Current status of Melcor 2.2 for fusion safety analyses
MELCOR is an integral code developed by Sandia National Laboratories (SNL) for the US Nuclear Regulatory Commission (USNRC) to perform severe accident analyses of Light Water Reactors (LWR). More recently, MELCOR capabilities are being extended also to analyze non-LWR fission technologies. Within the European MELCOR User Group (EMUG), organized in the framework of the USNRC Cooperative Severe Accident Research Program (CSARP), an activity on the evaluation of the applicability of MELCOR 2.2 for fusion safety analyses has been launched and it has been coordinated by ENEA. The aim of the activity was to identify the physical models to be possibly implemented in MELCOR 2.2 necessary for fusion safety analyses, and to check if those models are already available in MELCOR 1.8.6 fusion version, developed by Idaho National Laboratory (INL). From this activity, a list of modeling needs that emerged from the safety analyses of fusion-related installations has been
identified and described. Then, the importance of the various needs, intended as the priority for model implementation in the MELCOR 2.2 code, has been evaluated according to the technical expert judgment of the authors. In the present paper, the identified modeling needs are discussed. The ultimate goal would be to propose to have a single integrated MELCOR 2.2 code release capable to cover both fission and fusion applications
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