27 research outputs found
Validation and uncertainty analysis of ASTEC in early degradation phase against QUENCH-06 experiment
Two pyrene-tetrazole conjugates were synthesized as photoreactive chromophores that allow for the first time the combination of metabolic labelling of DNA in cells and subsequent bioorthogonal “photoclick” modification triggered by visible light. Two strained alkenes and three alkene-modified nucleosides were used as reactive counterparts and revealed no major differences in their “photoclick” reactivity. This is a significant advantage because it allows 5-vinyl-2′-deoxyuridine to be applied as the smallest possible alkene-modified nucleoside for metabolic labelling of DNA in cells. Both pyrene-tetrazole conjugates show fluorogenicity during the “photoclick” reactions, which is a second advantage for cellular imaging. Living HeLa cells were incubated with 5-vinyl-2′-deoxyuridine for 48 h to ensure one cell division. After fixation, the newly synthesized genomic DNA was successfully labelled by irradiation with visible light at 405 nm and 450 nm. This method is an attractive tool for the visualization of genomic DNA in cells with full spatiotemporal control by the use of visible light as a reaction trigger
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
The EC MUSA project on management and uncertainty of severe accidents: Main pillars and status
In the current state of maturity of severe accident codes, the time has come to foster the systematic application of Best Estimate Plus Uncertainties (BEPU) in this domain. The overall objective of the HORIZON-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” is to quantify the uncertainties of severe accident codes (e.g., ASTEC, MAAP, MELCOR, and AC2) when modeling reactor and spent fuel pools accident scenarios of Gen II and Gen III reactor designs for the prediction of the radiological source term. To do so, different Uncertainty Quantification (UQ) methodologies are to be used for the uncertainty and sensitivity analysis. Innovative AM measures will be considered in performing these UQ analyses, in addition to initial/boundary conditions and model parameters, to assess their impact on the source term prediction. This paper synthesizes the major pillars and the overall structure of the MUSA project, as well as the expectations and the progress made over the first year and a half of operation
Evaluation of a Double-Ended Guillotine Break Transient in a Three-Loop PWR-900 like with TRACE Code Coupled with DAKOTA Uncertainty Analysis
In the present study, the model of a reference generic three-loop PWR-900 like western type reactor has been developed and a double-ended guillotine break on the cold leg has been simulated by the TRACE code. Through the SNAP graphical interface, a DAKOTA uncertainty analysis, based on the probabilistic method to propagate input uncertainty, has been performed by selecting uncertain parameters related to the safety injection system and to the initial plant status. In particular, six uncertain input parameters have been considered: the accumulators’ initial pressure and temperature, the safety injection system temperature and flow rate, the reactor initial power and the containment initial pressure. The main figure of merit selected for the application of regression correlation is the hot rod cladding temperature. Both Pearson and Spearman’s correlation coefficients have been computed for the cladding temperature of the hot rod to characterize its correlation with the input parameters in the different phases of the transient. In addition, the dispersion of the calculated data have been evaluated for selected relevant thermal- hydraulic parameters, such as the primary pressure, the core mass flow rate and the water level in the vessel
Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis
In the present study, the model of a generic three-loops PWR-900 western type reactor has been developed and a double-ended guillotine break on the cold leg has been simulated by TRACE code. Through the SNAP graphical interface, a DAKOTA uncertainty analysis, based on the probabilistic method to propagate input uncertainty, has been performed by selecting uncertain parameters related to the safety injection system and to the initial plant status. In particular, six uncertain input parameters have been considered: the accumulators’ initial pressure and temperature, the safety injection system temperature and flow rate, the reactor initial power and the containment initial pressure. The main figure of merit selected for the application of regression correlation is the hot rod cladding temperature. Both Pearson and Spearman’s correlation coefficients have been computed for the cladding temperature of the hot rod to characterize its correlation with the input uncertain parameters in the different phases of the transient. In addition, the dispersion of the calculated data have been discussed for selected relevant thermal-hydraulic parameters, such as the primary pressure, the core mass flow rate and the water collapsed level in the vessel