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The International Reactor Innovative and Secure (IRIS) is modular, economic, medium size (1000 MWth), grid appropriate and based on the safety-by-designTM philosophy. The integral configuration of the primary system, the strong coupling of reactor vessel and containment, during accident conditions, and the passive safety systems guarantee a very high degree of inherent safety. The IRIS International Consortium, that led by Westinghouse has developed the plant design, is strongly pursuing the objective to obtain the Final Design Approval (FDA) by the U.S. Nuclear Regulatory Commission (NRC), by 2013.The relative licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is being designed and will be built at SIET laboratories. In the early 90s, the SIET company upgraded the SPES facility (simulating a three loop PWR for the italian PUN - Progetto Unificato Nazionale) in SPES2, providing the experimental data that allowed the licensing of the Westinghouse AP-600 reactor. On the basis of the lessons learned and relying on existing auxiliary systems, the simulation of an accidental sequences matrix on the SPES3 facility will provide the needed experimental results to verify the general behaviour of the system and allow a code assessment process suitable to produce a reliable tool for the IRIS plant analyses.The main scaling parameters for the facility design have been determined through a Phenomena Identification and Ranking Table (PIRT) and a Hierarchical Two-Tired Scaling Analysis (H2TS) process resulting in 1:100 volume and power scale, 1:1 elevation scale, prototypical fluid at plant pressure and temperature nominal conditions. The detailed scaling of all plant components is the result of an iterative process between a Top-Down system scaling and Bottom-Up process scaling, i.e. a continuous verification of the experimental facility component design adequacy to appropriately represent what expected in the plant.The SPES3 facility simulates the primary, secondary and containment systems of the IRIS reactor. The Reactor Vessel (RV) includes the fuel bundle with electrically heated rods, the inverted hat pressurizer (PRZ) and three helical coil steam generators (SG) simulating the eight IRIS SGs. A single outer pump simulates the eight IRIS primary internal circulation pumps. Two Emergency Boration Tanks (EBT), the two Direct Vessel Injection (DVI) lines and the Automatic Depressurization System (ADS) are also simulated.The secondary side feed lines and steam lines are simulated up to the main isolation valves and the three Emergency Heat Removal Systems (EHRS) represent the four trains of the IRIS primary safety system. The facility configuration is suitable to investigate the natural circulation loops that allow removing the decay heat during the long term accidental transients. The IRIS containment compartments are simulated in SPES3 by appropriately connected separated tanks, representing the Dry-Well (DW), two Pressure Suppression Systems (PSS), two Long Term Gravity Make-up Systems (LGMS), the Reactor Cavity (RC) and the ADS Quench Tank (QT). Shape and dimensions are fixed in order to reproduce the trend of IRIS compartment volumes versus height. Both the primary and secondary loops are designed for 17.25 MPa pressure and the corresponding saturation temperature. The containment tanks are designed for 2 MPa pressure and the corresponding saturation temperature.In order to fit the new facility components and pipes on the existing steel structure of SPES2, an appropriate design of piping layout has been defined.A large set of instruments (about 600) will be installed on SPES3 to provide data both for the test run and analysis. It will consist of conventional (relative and absolute pressure transmitters and temperature sensors) and special (for two-phase flow measurement) instrumentation.The planned test matrix consists of 13 integral tests and 2 separate effect tests. The integral tests, which will simulate Design Basis Accidents (DBA) and Beyond Design Basis Accidents (BDBA) scenarios, are aimed at assessing the plant response and verifying the safety system effectiveness. The considered accidental events consist of a series of split and Double Ended Guillotine (DEG) breaks, ranging from 2 to 6 inch equivalent, on the main piping of the primary and secondary sides (DVI, EBT balance line and ADS, Feed Line and Steam line). The separate effect tests are devoted to investigate the interaction and characterize the heat transfer of innovative design components like the helical coil steam generators and the EHRS for the long term removal of the decay heat.The Relap5 code is used for the complete simulation of the facility through the three steps planned to achieve the foreseen objectives: a) Supporting design analyses aimed at obtaining feedback information on the facility design (in particular the comparison between the SPES3 facility and the IRIS reactor simulations through the Relap5+Gothic codes will provide information on the appropriateness of the performed scaling choices); b) Pre-test analyses aimed at the test design and test procedure set-up; c) Post-test analyses and code assessment on a set of qualified data to be used to tune the IRIS-specific analyses.Thanks to the iteration between facility design and analysis, the SPES3 facility will provide suitable experimental data of the main tests required by the NRC. The code assessment on such data will guarantee a reliable tool to perform the IRIS reactor safety analyses necessary for the Final Design Approval.

The mission of the SIET company, since it was founded in 1983, is to carry out experimental studies on the thermal-hydraulics of nuclear power plants components and systems. The general goal of the SIET testing activities is to provide data for design certification required in the NPP licensing process. The design development and the experimental qualification of specific plant components, as well as the assessment of thermal-hydraulic codes to be used for NPP safety analysis contribute to reach this goal. The SIET overall capabilities in terms of thermal power and fluid flow-rates are big enough to allow full scale or large scale testing of both single components and systems in a wide range of fluid conditions as shown below. - High pressure super-heated steam flow-rate 8 Kg/s (10 MPa, 500 °C); - High pressure saturated steam flow-rate 40 Kg/s (10 MPa); - High pressure water flow-rate 200 Kg/s (16 MPa, 330 °C); - DC power for Test Section heating 7 MW (70 kA); - AC power for plants 9 MWSince the è90s the large experimental facilities designed, built and operated by SIET to investigate the thermal-hydraulic behaviour of components (performance and/or concept demonstration), systems and plants (integral testing), have been mainly oriented to the advanced LWR and, in particular, to the Passive Safety Systems, the focal point of the SIET business. Four SIET experimental facilities are involved in the Passive Safety Systems testing: PANTHERS, GEST, SPES and IETI. PANTHERS (Performance ANalysis and Testing of HEat Removal Systems ) is a high power facility (20 MW) that allows testing of full scale components at prototypical steady-state and transient conditions. In particular for in-pool heat exchangers, a complete experimental assessment of heat transfer performances can be done in a wide range of operating conditions and with different primary fluids (i.e.: steam only, steam-air and steam-helium mixtures) during steady-state tests. Typical transient tests concern system start-up under full normal operating and ATWS conditions, investigation on gas build-up and pool water level effects.As far as the past tests are concerned, an experimental campaign in support of the licensing of GE-SBWR (Small, Simplified Boiling Water Reactor) was performed in 1994-è95 for two prototypes of in-pool heat exchangers, IC (Isolation condenser) and PCC (Passive Containment Condenser). Both thermal-hydraulic and structural behaviour were investigated. In 2003 the facility was used to successfully test an innovative emergency heat removal system designed by SIET within an ENEA funded research program. The system, called PERSEO (in-Pool Energy Removal System for Emergency Operation), consists of a heat exchanger and two pools connected to each other at the bottom and the top, being the heat exchanger pool empty and the other one full of cold water during stand-by condition. By opening a valve installed on the pool bottom connecting pipe the system enables the heat removal by flooding the heat exchanger. The Relap-5 code was extensively validated against heat transfer phenomena in vertical bundles located in water pools.The GEST (GEnerator Separator Tests) facility plant was originally designed for experiments on the PWR Steam Generators at full conditions. The facility was modified in 1996 for testing injection systems as the PIDS (Passive Injection and Depressurisation System) conceived by ANSALDO Nuclear Division for ALWR applications and the SIP-1 (Sistema di Iniezione Passiva) conceived by SIET. In particular, the SIP-1 passive injection system tests demonstrated its capability of water injection into a pressurised vessel (e.g.: reactor vessel or PWR steam generator vessel), simply unbalancing the passive system circuit by a low water level occurrence due, for example, to a Loss of Coolant Accident. The SPES (Simulatore Pressurizzato per Esperienze di Sicurezza - pressurized simulator for safety experiences) facility, initially simulating a three loop PWR and operated for integral tests on the italian PUN (Progetto Unificato Nazionale - National Unified Project), was upgraded in the è90s in order to carry out an integral test program devoted to the Westinghouse AP-600 design certification, with particular focus on the performances of the passive safety systems of such a reactor. The IETI (Impianto per Esperienze Termo-Idrauliche - facility for thermal-hydraulic tests) loop is a multipurpose-high performance test facility (20 MPa, 430°C, 15 kg/s water and 4 kg/s steam flow-rates). It was initially operated for studying the thermal-hydraulic behaviour of LWR power channels. In 1992 the IETI loop was modified to perform tests on Steam Jet Pumps (SJP) for LWR Passive Safety Systems. The SJP (or Steam Injector) is a static device where steam and subcooled water are mixed together and the delivered sub-cooled water can reach a pressure significantly higher than the inlet steam pressure. Therefore the SJP were studied in the frame of the emergency systems for ALWRs as injection and make-up in the LWRs primary loops as well as emergency feed water in the PWRs secondary loops. Two experimental campaigns, sponsored by ENEL, were successfully completed at SIET on a SJP designed by CISE in the è90s, aimed at measuring the SJP performance at the LWR conditions both in steady-state and transient situations (start-up and shutdown). Subsequently different type of SJP were tested at SIET in the frame of European Projects (SYNTHESIS-1996/1998 and DEEPSSI-2001/2003). More recently (2005-2006) a multi-stage Steam Injector, designed by Toshiba Corporation for the ABWR, was successfully tested at SIET. Thanks to the large number of experimental campaigns run on different facilities and devices, SIET has gathered a significant experience on the Passive Safety Systems for Light Water Reactors which represents an unique asset of this company, even in the context of a forthcoming worldwide renaissance of nuclear power.

The International Reactor Innovative and Secure (IRIS) is a modular, economic, medium size (1000 MWth), grid appropriate reactor belonging to the Innovative Nuclear Power Plants (NPP), in which a new plant configuration is obtained through the evolution of the mature PWR technology with a safety-by-designTM approach. The IRIS International Consortium, led by Westinghouse, is finalizing the design with the objective of obtaining the Final Design Approval (FDA) by the U.S. Nuclear Regulatory Commission (NRC), by 2013. The main feature of the IRIS is an integral primary system configuration with all main components contained in the Reactor Vessel (RV): pressurizer, primary circulation pumps and once-through helical coil steam generators. A compact spherical steel containment is part of the IRIS safety systems and is involved in a coupled dynamic behaviour in the passive mitigation strategy which enhances the safety and reliability of the IRIS.One of the steps required by the licensing process is the execution of integral and separate effect tests on a properly scaled reactor simulator of the IRIS reactor. The SPES3 facility is being designed and will be built at SIET laboratories in the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development. The simulation of a series of accidental sequences will provide data to verify the general behaviour of the system and allow a wide code assessment guaranteeing a reliable tool for the IRIS plant analyses.In the early 90s, the SIET company upgraded the SPES facility (simulating a three loop PWR for the italian PUN - Progetto Unificato Nazionale) in SPES2, providing the experimental data that allowed the licensing of the W AP-600 reactor. On the basis of the gained experience and relying on the same auxiliary systems as SPES2, the SPES3 will provide the experimental results in sight of the IRIS Final Design Approval (FDA) by 2013. The main scaling parameters were defined on the basis of a Phenomena Identification and Ranking Table (PIRT) and of a Hierarchical Two-Tired Scaling Analysis (H2TS) methodology. The IRIS System Decomposition and the Scale Identification (first two steps of H2TS) led to choose 1:100 volume and power scale, 1:1 elevation scale, prototypical fluid at plant pressure and temperature nominal conditions. The detailed scaling of all components of the plant is the result of an iterative process between a Top-Down system scaling and Bottom-Up process scaling, i.e. a continuous verification if the experimental facility component design choices are suitable to represent what expected in the plant. The SPES3 facility simulates the primary, secondary and containment systems of the IRIS reactor. The Reactor Vessel (RV) includes the fuel bundle with electrically heated rods, the inverted hat pressurizer (PRZ) and three helical coil steam generators (SG) simulating the eight IRIS SG modules. A single outer pump simulates the eight IRIS internal primary circulation pumps. Two Emergency Boration Tanks (EBT), the Direct Vessel Injection (DVI) lines and the Automatic Depressurization System (ADS) are also simulated.The secondary side feed lines and steam lines are simulated up to the main isolation valves and the three Emergency Heat Removal Systems (EHRS), represent the four trains of the IRIS primary safety system. The facility configuration is suitable to investigate the natural circulation loops that allow the decay heat removal during the long term accidental transients. Both primary and secondary side are designed for 17.25 MPa pressure and the corresponding saturation temperature.The IRIS containment compartments are simulated in SPES3 by separate tanks, connected by pipes. The tanks, in particular, represent the Dry-Well (DW), two Pressure Suppression Systems (PSS), two Long Term Gravity Make-up Systems (LGMS), the Reactor Cavity (RC) and the ADS Quench Tank (QT). In some cases, the shape of the tanks differs from cylindrical in order to reproduce the trend of IRIS plena volumes versus height. The containment tanks are designed for 2 MPa pressure and the corresponding saturation temperature.A complex design of piping layout allows to fit the new facility components and pipes on the existing steel structure that housed the SPES2. The SPES3 facility will be equipped with about 600 instruments, mainly of conventional type, like relative and absolute pressure transmitters and temperature sensors. Special instrumentation for two-phase flow measurement will be installed too. A matrix, including 15 tests: 13 for integral effects and 2 for separate effects, is foreseen on SPES3. The integral tests are devoted to verify the facility response and safety system effectiveness to take the plant in safe conditions after Design Basis Accidents (DBA) and Beyond Design Basis Accidents (BDBA). Split and Double Ended Guillotine (DEG) Small Break Loca Accidents (LOCA) are foreseen both on the primary side (DVI, EBT balance line and ADS) and the secondary side (Feed Line and Steam line) ranging from 2 to 6 inch equivalent.The separate effect tests are devoted to investigate and characterize the behaviour of innovative design components like the helical coil steam generators and their interaction/heat transfer capabilities with the EHRS.The Relap5 code is used to simulate the SPES3 facility in all its parts. Three steps are foreseen: a) Supporting design analyses aimed at obtaining feedback information on the facility design. In particular the comparison between the SPES3 facility and the IRIS reactor simulations (Relap5+Gothic codes) will provide information on the correctness of the performed choices; b) Pre-test analyses aimed at the test design and test procedure set-up; c) Post-test analyses and code assessment on a set of qualified data.The SPES3 facility will provide experimental data of the main tests required by the NRC. The code assessment on such data will guarantee a reliable tool to perform the IRIS reactor safety analyses in sight of the Final Design Approval.

The synthesis and microbiological activities of 1-arylideneaminoimidazole derivatives are reported. Antimicrobial data show that some of the tested imidazoles exhibited an interesting activity on Candida albicans.