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ASTEC - RAVEN coupling for uncertainty analysis of an ingress of coolant event in fusion plants
Abstract
The integrated ICE (Ingress-of-Coolant Event) facility, scaled 1/1600 with respect to the ITER-FEAT design, was built at JAERI with the aim of reproducing the phenomenology occurring in an ICE accident. An ICE occurs when a rupture in the coolant pipes causes the pressurized coolant to enter into the Plasma Chamber, which is held under high vacuum condition. A suppression system is used to mitigate the overpressurization and to prevent mechanical damages to the structures. The CPA module of the ASTEC severe accident code (Study carried out with ASTEC V2, IRSN all rights reserved, [2020]), has been adopted for the modelling and the simulation of a test conducted in the ICE facility. The experimental results of the main thermal-hydraulic parameters have been compared to the code results to characterize the ASTEC capability to predict the phenomenology of a low-pressure two-phase flow transient occurring in a fusion reactor. By coupling the ASTEC code with the uncertainty tool RAVEN, developed by Idaho National Laboratory, an uncertainty analysis has been conducted on the transient. The aim of the present activity is to investigate the dispersion and the sensitivity of the code response to the variation of selected uncertain input parameters, which could influence the simulation of an ICE. The activity also provides a first application of uncertainty analysis through the RAVEN-ASTEC coupling.
... Since RAVEN does not have a dedicated coupling interface for ASTEC, the coupling was realized by developing a specific Python interface that was embedded in the RAVEN source code. [27] The new interface has the same features of the "generic" interface of RAVEN, with the addition that the software is able to locate the output file of ASTEC and inspect its content to understand if a simulation has successfully ended or it failed. This is a key advantage in the case of UQ studies, where failed calculation results must be identified and discarded from the statistics. ...
... The ASTEC-RAVEN coupling workflow for UQ analysis is schematized in Fig. 4. More details are available in Ref. [27]. ...
... Fig. 4. Scheme of ASTEC-RAVEN coupling workflow for UQ analysis. [27] NUCLEAR TECHNOLOGY · VOLUME 00 · XXXX 2025 ...
Severe accident (SA) codes and their core degradation models have to deal with strongly nonlinear and discontinuous phenomena. In the application of uncertainty quantification to SA simulations, the combination of such phenomena may lead to a strong increase in the uncertainty propagated through the simulation, as well as to the chaotic behavior of the output variables. In this framework, the application of the limit surface search method of the RAVEN tool is proposed for a case where cliff-edge effects of SA phenomena determine a bifurcation of an output figure of merit. The algorithm is based on a predictive method making use of a support vector machine model, and it is applied with the aim of separating those input values that lead to different phenomenologies among the uncertainty calculations. The case study is in regard to the uncertainty analysis of the ASTEC code simulation of the QUENCH6 experimental test conducted in the framework of the International Atomic Energy Agency Coordinated Research Project I31033.
... With the aim to evaluate the code uncertainty in the simulation deriving from specific uncertainty sources, the probabilistic propagation of input uncertainties method was applied by means of ASTEC code coupling with the UT RAVEN and the implementation of the codes coupling on a HPC platform [31,32]. Ranges and PDFs of the 8 selected input uncertain parameters were retrieved by previous UA. ...
... /RAVEN, dispersion of the calculated data of the PC pressure vs experimental data[31] Figure 17 ASTEC/RAVEN Spearman's coefficient for the correlation of the uncertain input parameters with the PC pressure[31] ...
... /RAVEN, dispersion of the calculated data of the PC pressure vs experimental data[31] Figure 17 ASTEC/RAVEN Spearman's coefficient for the correlation of the uncertain input parameters with the PC pressure[31] ...
Deterministic safety analyses are employed to characterize the behavior of Nuclear Power Plants (NPP) due to postulated initiatory events. In fact, they give the necessary information to judge if selected safety requirements are fulfilled by NPPs in steady and transient conditions. In general, high challenging scenarios (e.g. DBA and BDBA) are selected for deterministic safety analyses. Despite the high level of maturity reached by state-of-art safety analyses codes, in their application there are still some sources of uncertainty affecting the calculation results (e.g. code uncertainty, nodalization effect, scaling issue, plant uncertainty, user effect). For this reason, it is important to quantify the uncertainty of safety analyses calculations. Several methodologies to characterize the uncertainty of the results have been developed in the past, and they can be divided in methods to propagate the input uncertainty (probabilistic and deterministic methods) and method to extrapolate the output uncertainty. In this framework, the target of this paper is to present the probabilistic method to propagate input uncertainty in the context of deterministic safety analyses. The presentation will be supported by some exemplificative applications regarding both thermal-hydraulic and severe accidents analyses in fission plant and applications for fusion plant.
... For other codes, the coupling can be done through a generic interface or, as an alternative, users can develop their own specific Python interface to be included in the source-code. By choosing this last option, a dedicated RAVEN-ASTEC coupling interface was developed by ENEA (Maccari et al., 2021). The ASTEC input-deck was also properly modified for the codes coupling, i.e. to allow RAVEN to retrieve the information needed to modify the inputparameters. ...
... 23 input uncertain parameters were selected by KIT. These include Fig. 12. Scheme of ASTEC -RAVEN coupling workflow for UQ analysis (Maccari, 2021). geometric data parameters, initial and boundary conditions, integrity criteria and heat transfer models parameters. ...
Severe Accident (SA) integral codes, such as the Accident Source Term Evaluation Code (ASTEC) developed by IRSN, are used to simulate the phenomena occurring during accident progression in Nuclear Power Plants (NPPs) up to the source term evaluation. Code validation against experimental data is fundamental to carry out deterministic safety analysis and apply these codes to NPPs. In addition, in the Best Estimate Plus Uncertainty (BEPU) framework, the quantification of the results uncertainty is needed. In the framework of the IAEA CRP I31033 “Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors”, the QUENCH test-6 experiment, conducted at KIT, has been selected to develop an uncertainty analysis using the ASTEC v2.2b code. The accuracy of the best-estimate ASTEC simulation was evaluated with the Fast Fourier Transform Based Method (FFTBM) against the experimental data. Then, the uncertainty of the code results was quantified by using the probabilistic propagation of input uncertainties method, through the coupling of ASTEC with RAVEN (Risk Analysis and Virtual Environment). Beyond identifying the main sources of uncertainty affecting the simulated test, the outcomes of the work also include some general discussion on the uncertainty propagation in a SA sequence.
As the horizon of nuclear energy expands with the advent of small modular reactors, Generation IV reactors, and fusion reactors, there is a growing perspective that the licensing process could benefit from a more comprehensive approach. Moving beyond traditional deterministic and probabilistic risk assessment analyses might pave the way for a novel safety analysis paradigm propelled by the increasing computational power at our disposal. This paper explores different methodologies that can improve the outcomes of nuclear safety analysis. These range from uncertainty quantification techniques, aimed at enhancing the precision of safety margins, to deploying dynamic event trees by driving system code simulations, capturing the potential evolutions of severe accidents. These methodologies offer a better understanding of the management and consequences of nuclear accident scenarios, significantly improving the accuracy and efficiency of safety predictions compared to traditional methods. Specific case studies illustrate the practical application of these advanced techniques, demonstrating substantial improvements in predicting and managing the dynamics of severe accidents. These findings underscore the effectiveness of these methodologies in enhancing risk assessment capabilities and informing decision-making processes for nuclear safety management. The paper also emphasizes the importance of adaptability and continuous evolution, a call for action to address emerging nuclear safety concerns and highlights the utility of advanced tools like RAVEN.
Russian VVER-type reactors are one of the most common used commercial reactors in the world. The validation of severe accident codes using experimental data is focused in core degradation both in-vessel and ex-vessel phenomena such as the QUENCH-12 tests performed at the Karlsruhe Institute of Technology (KIT). This step is needed before integral codes are applied to evaluate the behavior of VVER-plants under severe accident conditions. The main goal of the QUENCH-12 test was to evaluate the hydrogen generation resulting from the injection of cold quench water into an overheated and oxidized bundle of fuel rod simulators representing the VVER-fuel rods. Such phenomena are expected to occur during a severe accident sequence e.g. LOCA or as part of an accident management measure. This paper describes the investigations done to simulate the QUENCH-12 test using the ASTEC V2, which is developed by IRSN to simulate all physical and chemical phenomena during severe accident condition. A model of the QUENCH-12 test facility was developed for ASTEC first time taking into account the specific material data, geometry and boundary conditions. The evaluation of the results have shown, that ASTEC V2 is able to predict the main trends of key-parameters during the all test phases in good agreement with the measured data. The integral hydrogen generation is only over-predicted during the quench phase. As next steps, the model will be improved by taking into account the real material properties of the ZrNb-cladding typical of Russian reactors instead of the ones of Zr. Also the sensitivity studies regarding the electrical heater resistance will be carried out.
SciPy is an open-source scientific computing library for the Python programming language. Since its initial release in 2001, SciPy has become a de facto standard for leveraging scientific algorithms in Python, with over 600 unique code contributors, thousands of dependent packages, over 100,000 dependent repositories and millions of downloads per year. In this work, we provide an overview of the capabilities and development practices of SciPy 1.0 and highlight some recent technical developments. This Perspective describes the development and capabilities of SciPy 1.0, an open source scientific computing library for the Python programming language.
Safety studies are performed in the frame of the conceptual design studies for the European Demonstration Fusion Power Plant (DEMO) to assess the safety and environmental impact of design options. An exhaustive set of reference accident sequences are defined in order to evaluate plant response in the most challenging events and compliance with safety requirements.
The Functional Failure Mode and Effect Analysis (FFMEA) has been chosen as analytical tool, as it is a suitable methodology to define possible accident initiators when insufficient design detail is available to allow for more specific evaluation at component level. The main process, safety and protection functions related to the DEMO plant are defined through a functional breakdown structure (FBS). Then, an exhaustive set of high level accident initiators is defined referring to loss of functions, rather than to specific failures of systems and components, overcoming the lack of detailed design information. Nonetheless reference to systems or main components is always highlighted, as much as possible, in order to point out causes and safety consequences.
Through the FFMEA a complete list of postulated initiating events (PIEs) is selected as the most representative events in terms of challenging conditions for the plant safety.
All the four blanket concepts of the European DEMO reactor have been analysed.
A new major version of the European severe accident integral code ASTEC, developed by IRSN with some GRS support, was delivered in November 2015 to the ASTEC worldwide community.Main modelling features of this V2.1 version are summarised in this paper. In particular, the in-vessel coupling technique between the reactor coolant system thermal-hydraulics module and the core degradation module has been strongly re-engineered to remove some well-known weaknesses of the former V2.0 series. The V2.1 version also includes new core degradation models specifically addressing BWR and PHWR reactor types, as well as several other physical modelling improvements, notably on reflooding of severely damaged cores, Zircaloy oxidation under air atmosphere, corium coolability during corium concrete interaction and source term evaluation.Moreover, this V2.1 version constitutes the back-bone of the CESAM FP7 project, which final objective is to further improve ASTEC for use in Severe Accident Management analysis of the Gen.II-III nuclear power plants presently under operation or foreseen in near future in Europe. As part of this European project, IRSN efforts to continuously improve both code numerical robustness and computing performances at plant scale as well as users' tools are being intensified.Besides, ASTEC will continue capitalising the whole knowledge on severe accidents phenomenology by progressively keeping physical models at the state of the art through a regular feed-back from the interpretation of the current and future experimental programs performed in the international frame.
During the recent years, an increasing interest in computational reactor safety analysis is to replace the conservative evaluation model calculations by best estimate calculations supplemented by uncertainty analysis of the code results. The evaluation of the margin to acceptance criteria, for example, the maximum fuel rod clad temperature, should be based on the upper limit of the calculated uncertainty range. Uncertainty analysis is needed if useful conclusions are to be obtained from “best estimate†thermal-hydraulic code calculations, otherwise single values of unknown accuracy would be presented for comparison with regulatory acceptance limits. Methods have been developed and presented to quantify the uncertainty of computer code results. The basic techniques proposed by GRS are presented together with applications to a large break loss of coolant accident on a reference reactor as well as on an experiment simulating containment behaviour.
Among the Postulated Initiating Events in nuclear fusion plants, the Ingress of Coolant Event (ICE) in the Plasma Chamber is one of the main safety issues. In the present paper, the best estimate thermal-hydraulic system code TRACE, developed by USNRC, has been adopted to study the ICE, and it has been qualified based on experimental results obtained in the Integrated ICE facility at JAERI. A nodalization has been developed in the SNAP environment/architecture, using also the TRACE 3D Vessel component where multidimensional phenomena could occur. The accuracy of the code calculation has been assessed both from a qualitative and quantitative point of view. In addition, an Uncertainty Analysis (UA), with the probabilistic method to propagate the input uncertainties, has been performed to characterize the dispersion of the results. The analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in the SNAP environment/architecture. Results show the adequacy of the 3D nodalization and the capability of the code to follow the transient evolution also at a very low pressure. Response correlations have been computed to characterize the correlation between the selected uncertain input parameters and the Plasma Chamber pressure.
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.
An event of water coolant ingress into the vacuum vessel (VV) is one of the most important events leading to severe consequences in nuclear fusion reactors. The ingress of coolant to the VV could appear due to coolant pipe rupture of in-vessel components. Any damage of in-vessel components could lead to water ingress and may lead to pressure increase and possible damage of the VV. Therefore, it is important to understand thermohydraulic processes in the VV during the ingress of coolant event (ICE) to prevent overpressurization of the VV. This technical note updates the developed Wendelstein 7-X (W7-X) model in accordance with the experience gained from the modeling of ICE experiments. Calculation results using the updated model are compared with the results obtained using an older model and the results of other researchers. The calculation results of the updated W7-X model show a much smaller pressure increase rate in the VV compared to the old model. In order to find the maximal area of partial break, which increases pressure in the VV but does not reach burst disk activation pressure (no steam release from the VV to the environment), the best-estimate approach is provided. The results of the analysis reveal that partial break using the updated W7-X model could be much bigger than what was considered before.
Accident analysis of fusion facilities is important part of safety assessment. Validated computer codes are required to perform such analysis. At present ASTEC code is being actively developed by IRSN to include features required for safety analysis of fusion facilities. This paper presents detailed analysis of P1 test performed at ICE test facility using CPA module of ASTEC code. Results of performed analysis confirmed that ASTEC could be successfully applied for analysis of loss of coolant accidents in fusion facilities like ITER or W7-X. Modelling of the pressure suppression system included in ASTEC is adequate as well. Peculiar gas compression effect was observed in the drain tank. This compression effect and gas temperature behaviour could be further investigated to resolve remaining issues.
An integrated ingress of coolant event (ICE) test facility was constructed to demonstrate
that the ITER safety design approach and design parameters for ICEs are adequate. The major objectives
of the integrated ICE test facility are: to estimate the performance of an integrated pressure
suppression system; to obtain validation data for safety analysis codes; and to clarify the effects
of two phase pressure drops at the divertor and condensation in the suppression tank.
The integrated ICE test facility simulates the current ITER components with a scaling factor of
1/1600. The TRAC-PF1 code is used to verify the integrated ICE experimental results. From
the present study the effectiveness of the ITER pressure suppression system was verified
experimentally and analytically, and then it was clarified quantitatively that the
prediction accuracy of the TRAC-PF1 code is very high.
An integrated ICE (Ingress-of-Coolant Event) test facility was constructed in order to demonstrate the adequacy of the ITER (International Thermonuclear Experimental Reactor) safety design approach and investigate two-phase flow behavior during an ICE event. The integrated ICE test facility corresponds to approximately 1/1600 of the ITER components. In the experiments the fluid flow configurations inside the vacuum vessel at the ICE events were observed visually and pressure transients were measured quantitatively. Two-phase flow analyses were carried out with the TRAC code and the experimental results were validated numerically. From the present study it was clarified that the ITER pressure suppression system is very effective in reducing the pressure rise in case of the ICE event and the water–vapor two-phase flow characteristics in ITER can be predicted numerically with sufficient accuracy.
For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA–INTRA codes and presents their modeling results for the following breaches of a water cooling line into the tokamak vacuum vessel, (1) water injection in vacuum with condensation of the flashed steam and (2) water injection in vacuum. The modeling predictions are compared with experimental results from the ingress-of-coolant event (ICE) facility in Japan. It is observed that the codes’ predictions exhibit good agreement with the experiment data, which suggests that the codes include the proper physical mechanisms associated with the chosen accident scenarios. It is thereby concluded that either of the codes can be used to model the defined transient.
The long-term performance of a nuclear fuel waste disposal system is typically studied by modelling potential releases of contaminants to the environment and the consequent health risk to humans. To deal with uncertainty in the estimated consequences of the releases, the behaviour of the disposal system may be repeatedly simulated under the direction of a probabilistic assessment code. In each simulation, parameters in the system model take different values to reflect the uncertainty about their values in the real system. The selection of the possible values of the parameters is governed by a probability density function (PDF) for each parameter.This paper examines the question of how to derive parameter PDFs for such assessments. The paper includes: 1.1. A definition of the concept of probability, and the mathematical properties of PDFs and the related cumulative distribution functions (CDFs).2.2. A discussion of the influence of the context in which PDFs are derived on their form and interpretation. Key elements of the context include what is known about the disposal system, the structure and extent of the model description of the system, and how the results of the assessment will be used.3.3. A description of approaches to deriving parameter PDFs. In cases where data are unavailable or incomplete, reliance must be placed on the judgement of experts to generate subjective probability distributions.Examples are given of PDFs defined for recent Canadian assessments.The paper concludes with remarks on managing reliably the large body of PDF data for an assessment, how values are sampled from PDFs, and desirable future developments.
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