Pressurized Water Small Modular Reactor (SMR), Design Basis Accident Analysis using the ASTEC code

According to classifications adopted by the IAEA, small reactors are characterized by an equivalent electric output of less than 300 MW while medium-sized reactors by an equivalent electric power between 300 and 700 MW. Pressurized water small and medium sized reactors (SMR) generally adopt an integral layout of the primary circuit with in-vessel location of steam generators and control rod drives; one compact modular loop-type design features reduced length piping, an integral reactor cooling system accommodating all main and auxiliary systems within a leak-tight pressure boundary, and leak restriction devices. In this paper, a description is given of the development of the modelling and noding of the primary loop, secondary loop passive core cooling system and containment for a SMR, based on the available data of the SPES3-IRIS integral test facility. SPES3-IRIS is under construction at SIET laboratories in Piacenza (Italy), simulating with 1:100 volume scale and 1:1 height scale, the primary, secondary, containment and safety systems typical of the IRIS small modular reactor. Three ASTEC code modules were adopted: the ICARE module to predict the in-vessel phenomena, the CESAR module to compute two-phase thermal\u2013hydraulics in the Reactor Cooling System (RCS) and for the control and safety systems, and the CPA module to evaluate thermal\u2013hydraulic and aerosol behaviour in the reactor containment. The SMRs as well as the advanced nuclear water-cooled reactors rely on containment behaviour to achieve some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts. Thus, to simulate correctly the main phenomena involved during an accident scenario, the coupling between primary circuit and containment has to be reproduced accurately. Furthermore, given that the containment plays a fundamental role during every accident scenario, it has to be taken into account just as a real safety system. The worst design basis event for the SMR was analysed, and the calculated results were compared with those obtained by the University of Zagreb in collaboration with Westinghouse using the coupled codes RELAP-GOTHIC. The aim of this work is to evaluate the applicability of ASTEC coupled modules in the safety analyses of the new reactor systems with strong interaction between primary system and containment

Exploratory Studies of Small Modular Reactors Using the ASTEC Code

Nuclear safety has been one of the major issues studied since the inception of the nuclear industry. Establishing and maintaining core cooling and ensuring containment integrity are two main goals that nuclear safety must guarantee. Improvement in these safety systems has generally involved the development of suitable Passive Containment Cooling Systems (PCCSs). This kind of safety approach poses significant issues for computational and analysis methods since the vessel and containment are strongly coupled and the system response is based on the interaction between the two. This is the case of Small Modular Reactors (SMRs), which adopt a completely passive safety approach, and the integral design eliminates the large coolant loop piping, which in turn eliminates large Loss-Of-Coolant-Accidents (LOCAs) as well as the individual component pressure vessels and supports. For these reasons, no severe accident analyses have yet been conducted on this type of plant. Nevertheless, it is useful to investigate the possible consequences of a multiple failure scenario in these advanced systems. In order to perform these analyses, starting from the available data of the SPES3 experimental facility, a SMR model was developed for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. The facility based on the IRIS reactor (International Reactor Innovative and Secure) design reproduces the primary, secondary and containment systems with a 1:100 volume scale, full elevation and prototypic fluid and thermal hydraulic conditions. The IRIS reactor is a SMR developed by an international consortium led by Westinghouse/BNFL, which includes universities, national laboratories, commercial companies and utilities. The Design Basis Accident (DBA) Direct Vessel Injection (DVI) line double-ended guillotine break was the reference transient that allowed matching the trend of the main physical parameters predicted by the ASTEC code model with those computed by the well-established best estimate coupled codes RELAP-GOTHIC. In this paper, a multiple system failure scenario was reproduced, to investigate and evaluate the in-vessel phase phenomena and the effectiveness of the passive mitigation measures. The results of the calculations confirmed the good performance of the IRIS system during the DBA accident, and showed for the first time how the ASTEC code can reproduce well the behaviour of this non- prototypic system

Post-test of the sgtr experimental tests carried out in the LIFUS facility

The steam generator tube rupture (SGTR) accident is considered by the International Regulatory Bodies a design basis event in HLM technology because of its safety concerns in terms of impulsive dynamic load caused by the thermal interaction between the lead/lead-bismuth-water. Aim of this study is to investigate, through numerical analyses, the dynamic response induced in LIFUS5/Mod2 facility during a SGTR. In particular it is intended to furnish a post-test evaluation of the structural effects caused by the interaction between water and lead-bismuth (during a SGTR accident event) with reference to the experimental tests carried out at the ENEA Brasimone Research Centre. The interacting fluids are Lead Bismuth Eutectic (LBE) at 0.1 MPa and 400°C and water at 4 MPa and 200°C. After a briefly description of Test A1.3 and A1.4, executed at the ENEA Brasimone Research Centre in the LIFUS 5/Mod 2 facility, the result of the numerical FEM analyses are presented and discussed. The propagation of the pressure wave was simulated by 3-dimensional Finite Element Method (FEM) models implemented in suitable FEM structural codes. The obtained results, validated by a comparison with the experimental data results, indicate that no significant dangerous effects on the LIFUS test vessel occur - as stress levels are below the allowable limit value - with the assumed geometry and in the testing conditions. The present work may also provide an useful contribution to the development of lead reactor technology by means of the test programme to be carried out in the LIFUS experimental facility (work performed in the framework of the 7th FP THINS)

Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

The development of a conceptual design of an industrial-scale transmutation facility (EFIT) of several 100 MW thermal power based on accelerator-driven system (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios

Ability of Current Advanced Codes to Predict Core Degradation, Melt Progression and Reflooding

The experimental database on core degradation and melt relocation (and their consequences on hydrogen production, vessel rupture) is limited to small-scale experiments which are only partially representative of what could occur in a reactor. As a consequence, there is uncertainty in the capability of codes to predict core degradation in postulated severe accident transients of nuclear power plants. The GAMA has launched an action in order to determine the ability of current advanced codes to predict core degradation in nuclear reactors. The TMI-2 scenario was selected as the case to analyze since it concerns the only full scale Pressurized Water Reactor to have experienced core degradation. Data from the code calculations were compared to the TMI-2 end-state to determine the codes‟ predictive capability. The study was completed in 2004, and is documented in ref. 2. One conclusion of the study is that variability in the codes predictions existed in part because initial conditions of the tMI-2 scenario were not well defined. It was concluded that code variability could be better evaluated if these conditions were better defined. Therefore, an additional task was proposed to benchmark the codes. This phase II study evaluates the variability in the codes‟ results using a postulated core degradation scenario of the TMI-2 reactor. The scenario was specified with simple initial and boundary conditions so that the influence of uncertainty of these conditions was minimized and the variability in the codes‟ results is more readily determine

Investigation of core degradation (COBE)

International audienceThe COBE project started in February 1996 and finished at the end of January 1999. The main objective was to improve understanding of core degradation behaviour during severe accidents through the development of computer codes, the carrying out of experiments and the assessment of the computer codes’ ability to reproduce experimental behaviour. A major effort was devoted to quenching behaviour and a substantial achievement of the project was the design and commissioning of a new facility for the simulation of quenching of intact fuel rods. Two tests, carefully scaled to represent realistic reactor conditions, were carried out in this facility and the hydrogen generated during the quenching process was measured using two independent measuring systems. The codes were able to reproduce the results in the first test, where little hydrogen was generated but not the second test, where the extra steam produced during quenching caused an invigorated Zircaloy oxidation and a substantial hydrogen generation. A number of smaller parametric experiments allowed detailed models to be developed for the absorption of hydrogen and the cracking of cladding during quenching. COBE also investigated other areas concerned with late-phase phenomena

Final synthesis report on ASTEC

This work provides a synthesis of the ASTEC activities which were carried out within the SARNET FP7 Work-Package number 4 (WP4). These WP4-ASTEC activities were distributed over 3 sub-Work-Packages: USTI (Users’ Support, Training and Integration), ACAS (ASTEC Code ASsessment) and AMEX (ASTEC Model EXtension). · The USTI activity aimed at providing a support to ASTEC code users through elaboration, improvement, delivery and maintenance of code versions and revisions, assistance in code use, organization of training courses … as well as at enhancing code use experience through organization of users’ club meetings. Its other goal was to allow capitalising the R&D knowledge in the field of severe accidents by integration of models proposed in the WP5, WP6, WP7 and WP8 work-packages. · The ACAS activity aims at covering a broad matrix of ASTEC reactor applications, aiming at the most important accident scenarios for 4 types of reactors (PWR, VVER, BWR, CANDU). As to Gen.III NPPs, while the global applicability of ASTEC V2 to the EPR design has been yet demontrated by IRSN, it was planned to also start examining the applicability of ASTEC V2 to few other designs (such as for instance AP1000, APR1400, …) in the frame of the ACAS workpackage. In complement to these reactor applications, the second main objective of the ACAS activity was to continue the ASTEC assessment against experimental data. · The AMEX activity aims at extending the current ASTEC applicability to BWR and CANDU reactors