The ASTEC (Accident Source Term Evaluation Code) software system makes it possible to simulate all phenomena that take place during a water-cooled reactor meltdown accident, from the initiating event to the discharge of radioactive materials (called the "source term") out from the containment . This particularly covers Western-designed pressurized water reactors (PWRs) such as those used for France's nuclear power supply, Russian-designed PWRs (VVERs), boiling water reactors (BWRs) and CANDU heavy water reactors. ASTEC is currently maintained and developed by the IRSN.
The main applications of the software package are safety analyses for nuclear reactors (e.g. the European Pressurised Reactor - EPR), source term evaluations in accident situations and the development of severe accident management guidelines . ASTEC is widely used in IRSN level 2 probabilistic safety assessments (PSA2) for French reactors. It is also used to prepare and interpret the experimental programs with respect to either in-vessel or ex-vessel phenomena which could occur during meltdown accidents. Furthermore, ASTEC is also occasionally used for preparing certain crisis exercises (nuclear or radiological accident simulation exercises for training various players and the organization itself) at the IRSN Crisis Technical Centre.
ASTEC is the European reference software within the European Commission
SARNET (Severe accident research network) network of excellence. It is also used by several organisations outside Europe (Canada, Russian Federation, India, Singapore, Ukraine, etc.).
ASTEC covers the entire phenomenology of severe accidents except steam explosion (for which the IRSN uses the
MC3D software) and the mechanical integrity of the containment (for which the IRSN uses the CEA CAST3M software package). Its modular structure
(see Figure 1) simplifies qualification of the physical models implemented in ASTEC by comparing the simulated results with those obtained experimentally.
Each module (described in details in  and ) handles phenomena that occur in a part of the reactor or phase of the accident, and in particular:
The degradation of materials within the vessel, when the temperatures reached under the effect of the core's residual power exceed a threshold leading to significant oxidization of the fuel rod claddings due to water vapor as well as various chemical interactions between the materials that make up either the fuel rods or the control rods. This may go as far as the materials melting, leading to the formation of a mixture of melted materials known as "corium". Another possibility is the formation of solid debris from the premature fragmentation of fuel pellets combined with the embrittlement of fuel rod claddings. The behavior of corium, once located in the vessel bottom after its slumping from the core, is modeled accounting dynamically for the possible stratification of materials to form several metal and oxide layers. This behavior is modeled until the vessel bottom barrier is either breached or the corium layers achieve to be stabilized on the inside if the vessel can be cooled from the outside (see IVMR below).
The release of fission products (FP), particularly iodine, from any fuel material in the core, i.e. release from both standing pellets or relocated fuel compounds (either particulate debris or corium).
The transport of FPs and aerosols as well as their physical and chemical behavior in the primary and secondary cooling systems, then in the containment (see Figures 3 and 4). Special attention is particularly paid to the behavior of the many iodized species in its various forms (molecular iodine, gaseous organic iodides, iodine oxides aerosols…).
The thermal-hydraulics within the containment using a 0D volumes approach, classically called "lumped-parameter code"
(see Figure 5).
The erosion of the vessel shaft's raft by corium located there in the event that the vessel is breached, taking into account possible subsequent arrival of further materials (linked to delayed slumping of corium into the cavity since the degradation processes in the core can extend beyond the vessel rupture time). This basemat erosion, called "corium-concrete interaction" (CCI) is modeled using a volume-based approach or single-dimensional layers (see figure 6). The model makes it possible to process the conditions of a dry CCI or a CCI under water, with the second situation including the cooling of corium by submersion in water based on the procedures for managing severe accidents.
Other models and features
ASTEC also simulates other phenomena, associated with the possible stages of the severe accident, notably including:
- direct containment heating (DCH) by the transfer of hot gases and corium droplets from the reactor cavity, following the rupture of the vessel;
- the combustion of hydrogen or carbon monoxide accumulated within the containment and the associated risk of explosion.
Furthermore ASTEC evaluates the radioactivity of the isotopes and the associated residual power in all parts of the reactor, as well as dose rates in the containment.
In ASTEC's most recent version (V2.1), the models are state-of-the-art. With respect more particularly to the behavior of fission products, much of the knowledge was gained when interpreting tests conducted as part of the Phébus and ISTP programs, then the OECD/NEA STEM/STEM2 projects. The scientific accuracy of the corresponding models is one of the ASTEC's major benefits over other international softwares. With respect to Generation III & III+ nuclear plants, specific models simulate the EPR corium catcher. Furthermore, new features have also been incorporated into version V2.1 in order to be able to cover the issues related to the strategy of retaining the corium within the vessel; these developments were carried out by IRSN as part of the European Project H2020 IVMR ("In-Vessel Melt Retention") (see ).
The core degradation models, an essential part of ASTEC, were greatly improved in the version V2 family, using those of the mechanistic software ICARE2, to which IRSN has devoted considerable effort since the early 1990s (see illustrations in Figure 7), particularly in connection with the progress of the experimental programs Phébus SFD and Phébus FP  and .
The ASTEC package has been validated by over 160 tests, including:
Analytical tests with separated or combined effects. For example, VERDON tests (CEA) relate to the release and transportation of fission products, CHIP+ (IRSN) the chemistry of aerosols in the cooling system, OLHF (SNL, U.S.A.) the mechanical breach conditions of the vessel bottom, and BIP (CNL, Canada) the behavior of iodine in the containment.
Integral tests, including the
Phébus FP in-pile tests (IRSN) simulating an entire accident with real materials up to the source term in the containment, or QUENCH tests (KIT, Germany) representing a bundle of electrically heated fuel rods in the core, using simulant materials or CCI tests (ANL, USA) dealing with corium-concrete interaction.
Among these 160 tests, the OECD/NEA ISP exercises (International Standard Problem) have been notably selected as these constitute an international reference by virtue of the high quality of the measurements and their use as benchmarks when comparing software packages (e.g. PACTEL, VANAM, BETHSY, LOFT tests, etc.). The validation matrix is continually being expanded by the results of international programs, including: CCI-OECD (Argonne National Laboratories, USA), OECD STEM/STEM2 at the IRSN, OECD BIP2/BIP3 (Canadian Nuclear Laboratories, Canada), OECD THAI-2/THAI-3 (Becker Technologies, Germany), etc.
For each new version of ASTEC, the IRSN itself conducts only some of the abundant validation work. Most of IRSN's international partners in the field of severe accidents have contributed to this vast software validation task, most frequently in the area of collaborative projects, and have for years. This was, for instance, the case for versions V2.1.0 and V2.1.1 as part of the CESAM project (see ) of European Commission FP7.
Finally, the software is regularly applied to light-water nuclear reactor real accidents that have taken place around the world in order to consolidate its results before applying it to reactor configurations in production or under construction worldwide. These reactor-scale validation tasks particularly relate to unit 2 of the American Three Mile Island reactor (TMI-2) that had an accident in 1979, and the three that suffered accidents at Fukushima Daiichi in Japan in 2011.
Its ability to appropriately simulate any core meltdown accident scenario for reactors currently in operation greatly benefited from the IRSN conducting level 2 probabilistic safety assessment (PSA2) on the 900 and 1300 MWe PWRs and the EPR. ASTEC can simulate the majority of safety systems and actions or procedures taken by operators working in current reactors (primary coolant system depressurization, containment spraying, discharging water into a rather heavily degraded core, hydrogen recombination in the containment, etc.).
Additionally, comparisons are regularly made with mechanistic softwares in order to verify that the thermalhydraulics of the cooling systems during the front end phase of a severe accident sequence is correctly assessed by ASTEC. The purpose of this approach is to ensure that the initial conditions of the in-vessel degradation phase as estimated by ASTEC can be considered sufficiently consolidated to be able to conduct a relevant analysis of the remainder of the accident sequence.
Since 2004, about 30 members of the SARNET network have been continually evaluating ASTEC, either through validation against the results of experimental programs (see above), or by means of benchmark comparisons with other software packages for accident scenarios in various types of reactor (900 MWe PWR Framatome, PWR Konvoi 1300 MWe, PWR Westinghouse 1000 MWe, VVER 440 MWe and VVER 1000 MWe). For many years, the IRSN has also been working in close collaboration on similar work with a number of organisations outside Europe, including the Kurchatov Institute (Russia), Atomic Energy of Canada Limited (AECL, now CNL, Canada), Bhabba Atomic Research Centre (BARC, India) and Atomic Energy Regulatory Board (AERB, Indian safety authority). Then, other non-European organizations have also joined the ASTEC project, notably including TSOs (Technical Safety and Support Organisations) such as SEC-NRS (Russia) and NSC (China), but also other partners in Asia such as CNPRI (China), NPCIL (India), and the national university of Singapore. More recently, few other organizations have joined the ASTEC community, including e.g. Energorisk (Ukraine) and Türkiye Atom Enerjisi Kurumu (TAEK, Turkey) near the EU, and Egyptian Nuclear and Radiological Regulatory Authority (ENRRA, Egypt) in the Middle East.
The work some years ago done in SARNET has shown that ASTEC V2.0 is able to simulate a large part of severe accident scenarios in boiling water reactors (BWR) and heavy water CANDU reactors, except for the stage of core degradation. The work later conducted as part of the CESAM project (2013-2017) showed that version V2.1, as expected, was able to eliminate this restriction, due to the newfound ability to adequately model the actual geometry of BWR and CANDU cores.
Furthermore, the IRSN participated with ASTEC in international benchmarking for the simulation and interpretation of the Fukushima Daiichi accidents coordinated by the OECD as part of the projects BSAF and BSAF2 (2012-2018) (cf.  and ).
Finally, intercomparison exercises with other severe accident software, such as MELCOR, developed by US-NRC, are also regularly conducted in order to quantify the discrepancies attributable to modeling choices in particular. These exercises supplement the various benchmarks carried out in the course of European projects (see above): One example is the "crosswalk" exercise (see ) where the target here is a BWR similar to the one in Fukushima.
ASCOM collaborative project
The ASCOM project was initiated by IRSN in the second half of 2017 as part of Technical Area #2 of the association NUGENIA to strengthen the existing link between the various users of the ASTEC software and their partnership. The ASCOM project thereby received the NUGENIA label in late 2017, to begin in mid-2018 for a period of 4 years. In concrete terms, ASCOM aims both to consolidate and preserve the knowledge gained from the earlier SARNET and CESAM projects and to collectively move towards the final goal of the entire ASTEC community, which is to have a computing tool that is both reliable and based on the latest knowledge in order to analyze the safety of various types of nuclear facilities, evaluate the source term, and evaluate the core meltdown accident mitigation management procedures.
Furthermore, in the continuation of the European CESAM project, the ASCOM project also seeks to increase the software's scope by collaboratively developing "generic" datasets in areas that have been largely undercovered by ASTEC (SMR, Gen.III and III+ reactors other than EPR, fuel storage pools, etc.).
Maintenance and user support
With almost a hundred users of ASTEC throughout the world, several engineers are employed to provide an efficient user support ensuring a rapid response to user requests.
Users can easily download any new code version or revision as well as any update of the code documentation through the ASTEC web portal (cf. Figure 8). Besides, a dedicated internet system has been established to manage the exchange of information between the developers of the software and the users, particularly bug reports and fixes (see Figure 9).
Yearly, IRSN proposes training courses to use ASTEC as part of ENSTTI (European nuclear safety training and tutoring institute).
ASTEC modelling perspectives
ASTEC is constantly incorporating model improvements that reflect advances in R&D in the field of severe accidents. ASTEC's validation will intensively continue in the OECD's standard exercises, as well as programs underway abroad (OECD BIP-2/BIP-3 on iodine chemistry, OECD THAI-2/THAI-3, on hydrogen risk in the containment covering the distribution, combustion, and recombination aspects, OECD CCI on corium-concrete interaction, CORDEB/CORDEB2 on the behavior of corium and debris in the vessel bottom, IPRESCA on pool-scrubbing, etc.) or at IRSN (PROGRES on reflooding debris beds, OECD STEM/STEM2 on the chemistry of iodine and ruthenium, etc.). IRSN plans to assign the conducting of some of the validation work to several partners, particularly as part of the ASCOM project.
Recalculations with ASTEC of accidents in the Japanese Fukushima Daiichi reactor continue, primarily as part of the successive projects coordinated by the OECD in the matter (see ,  and ).
Work has already been conducted to use some ASTEC modules in accident crisis tools in reactors, and this will continue; it should be emphasized that besides continually improving the relevance of ASTEC's physical models, which remains a permanent goal for IRSN, the other issue is to continue in parallel the digital optimization and software acceleration efforts that have already been begun within this framework. An expansion of the scope to core meltdown accident simulators is also possible.
Furthermore, still in connection to the crisis management activities, specific tasks aiming at progressively providing ASTEC with diagnosis capabilities (in supplement to its usual prognosis goal) was started in the CESAM frame through two main lines : by interfacing ASTEC with atmospheric dispersion tools (such as the pX software belonging to the IRSN C3X platform) in order to enhance capabilities of direct comparison with on-site measurement; by developing a methodology based on Bayesian networks for evaluating the probability of the different possible accident scenarios from the uncertain information provided by the available plant instrumentation . These works are being now continued in the frame notably of the H2020 FASTNET project coordinated by IRSN.
Besides ASTEC's many applications to Gen.II and Gen.III & III+ reactors, work to adapt the models has already been carried out for the safety analysis of the ITER fusion facility. Efforts in this direction will continue in the years ahead.
ASTEC's scope has also gradually expanded in recent years to Generation IV reactors, particularly sodium-cooled fast reactors as part of the European JASMIN project  (2012-2015). This work now continues as part of European project H2020 SMART-SFR  (2017-2021).
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