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Methodology for the improvement of the AINA Code wall-model applied to DEMO WCPB blanket
Abstract
The present work describes and supports the methodology for the improvement of the wall model developed in AINA and its specific application to the Japanese DEMO Water Cooled Pebbled Bed. The set-up and application of this approach aims to obtain robust models by estimating the behavior of the studied systems as accurately as possible. These systems are represented in a simplified way. This requires the computation of a 3D radiation transport which has been carried out by means of MCNP6.1, ADVANTG and thermal-hydraulic calculations using ANSYS® Fluent®. Several CFD mesh typologies and discretizations have also been employed to test the Richardson theorem. In addition, 1D simplified models have also been created and optimized for their usage in AINA code. The temperature distribution also shows good agreement (within 7%). In some cases the simplified models have not behaved in a conservative manner compared with the outcomes obtained for the 3D models. This observed absence of conservatism is intrinsic to the 1D approach. To cope with these effects, scaling functions have been determined as a ratio between the most conservative radial temperature distribution – computed by fully detailed 3D CFD – and the 1D simplified model. The scaling functions will be applied to the AINA computed wall temperature distribution. To conclude, the determination and coherence of the result obtained using independent tools and approaches, ANSYS® Fluent® vs AINA thermal-hydraulic routines, lead us to recommend the proposed methodology.
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This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space-and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.
Monte Carlo (MC) method has distinct advantages to simulate complicated nuclear systems and is envisioned as a routine method for nuclear design and analysis in the future. High-fidelity simulation with MC method coupled with multi-physics phenomena simulation has significant impact on safety, economy and sustainability of nuclear systems. However, great challenges to current MC methods and codes prevent its application in real engineering projects. SuperMC, developed by the FDS Team in China, is a CAD-based Monte Carlo program for integrated simulation of nuclear systems by making use of hybrid MC and deterministic methods and advanced computer technologies. The design objective, architecture and main methodology of SuperMC are presented in this paper. SuperMC2.1, the latest version, can perform neutron, photon and coupled neutron and photon transport calculation, geometry and physics modeling, results and process visualization. It has been developed and verified by using a series of benchmarking cases such as the fusion reactor ITER model and the fast reactor BN-600 model. SuperMC is still in its evolution process toward a general and routine tool for the simulation of nuclear systems.
In this contribution, the work done in AINA code during 2014 and 2015 at IFERC is presented. The main motivation of this work was to adapt the code and to perform safety studies for a Japanese DEMO design. Related to AINA code, the work has supposed major changes in plasma models. Significant is the addition of an integrated SOL-pedestal model that allows the estimation of heat loads at divertor. Also, a thermal model for a WCPB (water cooled pebble bed) breeding blanket has been developed based in parametric input data from neutronics calculations. Related to safety studies, a major breakthrough in the study of LOPC (loss of plasma control) transients has been the use of an optimization method to determine the most severe transients in terms of the shortest melting times.
The results of the safety study show that LOPC transients are not likely to be severe for breeding blanket, but for the case of divertor can induce severe melting. For ex-vessel LOCA (loss of coolant accident) analysis, it is severe for both blanket and divertor, but in the first case the transient time until melting is nearly two orders of magnitude higher. The results point out that the recovery time for plasma control system should be at least one order of magnitude lower than confinement time to avoid melting of divertor targets.
In this contribution, two ITER loss of plasma control events are investigated. The first one is a failure of pumps and/or pellet injection system which fuel the plasma, producing an overfuelling event. The second examines an increase of external heating during the steady state operation of ITER. For both events, initial simulation parameters are scanned in order to find out which could lead to highest fusion power.
Extensive simulation work has been done with AINA-1.0 (July-2007), the safety code developed by Fusion Energy Engineering Laboratory (FEEL-UPC) on the basis of SAFALY. AINA is a hybrid code comprising a zero-dimensional plasma dynamics and radial and poloidal thermal analyses of in-vessel components, and is intended to the quantitative investigation of plasma events in nuclear fusion reactors such as ITER.
FEEL-UPC research group has been developing AINA code since 2004. AINA code brings significant improvements in relation to the original SAFALY code, and FEEL-UPC aims to continue improving it in the near future, with new models for plasma wall interactions, blanket thermal analysis, and plasma edge and core.
The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. In the past, conservative analyses based on simple OD models for plasma balance equations and 1D models for wall heat transfer showed that a hypothetical scenario of first wall coolant tubes melting and subsequent entering of water in the vacuum vessel could not be totally excluded. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering Laboratory (FEEL) in Barcelona. It uses a OD-1D architecture, similar to that used for previous analyses of ITER loss of plasma control events. The results of this study show the simultaneous effect of two perturbations (overfuelling and overheating) over a plasma transient, and compare it with the isolated effects of each perturbation. It is shown that the combined effect can be more severe, and a method is outlined to locate the most dangerous transients over a nT diagram.
PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system, from the basic plasma physics to the generation of electricity. It has been used for many years, but details of its operation have not been previously published. This paper describes some of its capabilities. PROCESS is usually used in optimisation mode, in which it finds a set of parameters that maximise (or minimise) a figure of merit chosen by the user, while being consistent with the inputs and the specified constraints. Because the user can apply all the physically relevant constraints, while allowing a large number of parameters to vary, it is in principle only necessary to run the code once to produce a self-consistent, physically plausible reactor model. The scope of PROCESS is very wide and goes well beyond reactor physics, including conversion of heat to electricity, buildings, and costs, but this paper describes only the plasma physics and magnetic field calculations.
The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. Conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer are used to determine the effects of such transients on wall integrity from a thermal point of view.
In this contribution, progress in a “two simultaneous perturbations over plasma” approach to the analysis of the loss of plasma control transients in ITER is presented. The effect of variation in confinement time is now considered, and the consequences of this variation are shown over a n–T diagram. The study has been done with the aid of AINA 3.0 code. This code implements the same 0D plasma-1D wall scheme used in previous LOPC studies.
The rationale of this study is that, once the occurrence of a loss of plasma transient has been assumed, and due to the uncertainties in plasma physics, it does not seem so unlikely to assume the possibility of finding a new confinement mode during the transient.
The cases selected are intended to answer to the question “what would happen if an unexpected change in plasma confinement conditions takes place during a loss of plasma control transient due to a simultaneous malfunction of heating, or fuelling systems?”
Even taking into account the simple models used and the uncertainties in plasma physics and design data, the results obtained show that the methodology used in previous analyses could probably be improved from the point of view of safety.
Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.
Safety studies of plasma-wall events with {AINA} code for Japanese {DEMO}, Proceedings of the 12th International Symposium on Fusion Nuclear Technology-12 (ISFNT-12)
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ITER safety studies: the effect of two simultaneous perturbations during a loss of plasma control transient, Proceedings of the 11th International Symposium on Fusion Nuclear Technology-11 (ISFNT-11) Barcelona, Spain, 15–20 September, 2013
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R&D status on water cooled ceramic breeder blanket technology, Proceedings of the 11th International Symposium on Fusion Nuclear Technology-11 (ISFNT-11) Barcelona, Spain, 15–20 September, 2013
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