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[en] Very small special purpose reactors (SPRs) are under consideration by various entities for remote installations and communities. This type of plug-and-go resource requires inherently positive safety attributes and sound design. Such reactors may be transportable with fresh fuel to serve as a complete power “battery” and they may even be transportable on short notice after initial use. SPRs could eliminate costly and in some cases dangerous fossil fuel deliveries. Small passively safe reactors could be operated remotely by an offsite operations crew. While use cases for such reactors are currently under development, advanced manufacturing processes should be considered to support reactor economics.
[en] The first time I met George was in March 1998 for dinner. And the last time I had dinner with him was in March of last year, or two decades later exactly. During the first evening, we discussed and confronted our strategies for the short period of time left then until his retirement in 2004; was it sufficient time to contribute to the discipline with something risky and innovative? During the last four-hour dinner, after having evoked life and death rather joyfully and serenely, we looked back and analysed what came out of our two decades of collaboration. This paper, which was prepared with the aim to portray George’s achievements in thermal-hydraulics during the last 20 years of our partnership (1998–2018), is written in the spirit of narrating our epilogue culinary-science-talk, while trying to be faithful to his thoughts and ideas about the developments of the discipline and its perspectives. The paper introduces first the cascade of computational tools and discusses trends related to reactor safety problems and developments needed, as well as the need for new kinds of refined experimental data. Although George was not directly involved in all the examples presented here, he felt so concerned about the success of each project/case presented in this paper that he was virtually part of it; and as he wrote in our last paper (Yadigaroglu and Lakehal, 2016): this is after all his near-home work. Finally, in memory of the two other great scientists who have left us recently, Geoff Hewitt and Sam Martin, the content of the paper includes two cases in which both of them had collaborated directly or indirectly.
[en] The paper focuses on the performance of reflector trap passive safety system incorporated in the GFR 2400 reactor design. Multiple studies are raising question about the technical feasibility of the GFR 2400 concept, thus the new neutron trap passive system is introduced in the paper to cope with abnormal operation of this reactor. The coupled calculation scheme is used in the paper, where NESTLE code is used for coupled steady state and transient simulations of the reactor performance. The NESTLE code system solves multi-group transient diffusion equation utilizing nodal expansion method and is internally coupled with thermal-hydraulic sub-channel code. The SCALE6 software package is used for processing of macroscopic multi-group cross-sections that are used in the NESTLE code. The performance of the neutron trap passive system is simulated by the developed model for typical abnormal transients, such as control rod withdrawal or primary blower shutdown during the full power operation. Curie point latch acts as the actuation mechanism of the passive system. Temperature distributions are studied and the applicability of the neutron trap passive safety system is discussed in the paper. (author)
[en] The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory (Argonne) for advanced non-LWR safety analysis. To assist NRC to assess SAM capabilities for advanced reactor safety analysis and licensing at the NRC, a series of verification and other standard tests (OSTs) are modeled in SAM and code simulation results are compared with available analytical results. This report documents the preliminary SAM assessment using a matrix of six test problems. Each test problem is further examined with different boundary conditions, system configurations, or modeling options. Although relatively simple, these tests cover the basic equation models, basic component models, and basic system level processes and phenomena that must be modeled for advanced reactor safety analyses.
[en] Fast reactors (FR) have a unique potential to insure sustainability to the nuclear power option and to keep a large range of fuel cycle options open without leaving a legacy of highly radiotoxic and radioactive material to future generations, in case e.g. of an hypothetical drastic change of energy mix components in future and a reduced role of nuclear energy. To implement the favorable and flexible features of fast reactors it is however necessary to carefully analyze, beside core safety characteristics, all fuel cycle issues in order to assess the feasibility (also from the economy point of view) of the different strategies that can be based upon the use of fast reactors. (orig.)
[en] Highlights: •We describe the multi-scale approach as applied to nuclear fuel studies. •We present the perspective of multiscale fuel behaviour modelling at IRSN and PSI. •We describe fuel behaviour code FRAPCON, FALCON, SCANAIR and DRACCAR. •We present mesoscopic approach to modelling effects of fuel re-structuring. •We present MECOX and DIFFOX fuel cladding oxidation modelling. -- Abstract: The activities on fuel behaviour modelling using the multi-scale approach, carried out at IRSN and PSI, are described. The perspective of the two organizations on already acquired and potential future advantages from mutual application of the micro-, meso- and macroscopic simulations for fuel reliability and safety is presented. Finally, the conclusion is put forward regarding the merit to further develop the multi-scale approach to fuel behaviour modelling at IRSN and PSI.
[en] In response to the new regulatory requirements issued by the U.S. Nuclear regulatory Commission (NRC), the AP1000© Shield Building Design addresses the beyond design basis aircraft impact events where the unprotected structure was changed from a Reinforced Concrete structure (RC) to an enhanced Steel Concrete Composite (SC) structure. This paper summarizes the considerations, conclusions and lessons learned from the experiences on the development of the design process for the AP1000© Shield Building SC structure and its connections to the RC portion of the Building, with a specific overview to the existing code requirements for those structures. An overview of the SC design methodology of the AP1000 Shield Building is presented herein. It relies on the conservative application of the ACI349 code provisions for reinforcement concrete nuclear structures. This methodology is supported by an extensive international testing and other international building codes for SC structures. In line with the SC structure design, a discussion on the methodology and considerations for the analysis of the RC/SC connections is also developed, with a special insight to the ductility aspects of the connection during the design seismic events.
[en] Along with redundancy, physical separation and functional independence concepts, diversity shall be considered wherever a safety function requires a high level of reliability in an advanced reactor design. The paper presents the main stages of an innovative approach for a diversity analysis identifying the needs for additional diverse line to back-up main prevention line of safety functions involved in several Defence-in-Depth levels as well as its minimal reliability. The first methodology step consists in considering design extension condition by analysing both the combination of design basis events combined with the most credible common cause failure affecting the main prevention line, and multiple failures affecting safety systems used during normal plant operation. Then, this functional analysis is supplemented with probabilistic insights based on system reliability analysis in order to identify the components candidate for diversity. Finally, a detailed Failure Mode and Effect Analysis of the selected component allows the identification of critical redundant internal parts considering the probability of common cause failure affecting component safety functions. Diversity provisions to eliminate the occurrence or reduce the consequences of such common cause failures are then proposed. (author)
[en] In the framework of the European Horizon 2020 ESFR-SMART project, experimental data from Superphénix reactor are being recovered in order to support calibration and validation of the computational tools used for safety assessments of Generation-IV sodium fast reactors. Superphénix has been the largest sodium-cooled fast reactor ever built and operated in the world; it was connected to the grid in 1986 and shut down in 1997. During that period, a lot of measurements and experiments were performed and are available, being an important source of data for validation. In this sense, a new neutronics benchmark exercise has been proposed within ESFR-SMART Project based on the Superphénix start-up tests. Heterogeneous 3D core models for Superphénix reactor have been developed for both SCALE/KENO-VI and MCNP Monte Carlo neutron transport codes. The main goal of this work is to compare the results provided by each code with experimental data for a number of integral and local static parameters: criticality, worth of control rods, reactivity feedback coefficients and reaction rates in some specific core positions. A detailed comparison of calculated and experimental results has been carried out, encountering some discrepancies between MCNP and KENO-VI results even using the same nuclear data library (ENDF/B-VII.1). An exhaustive analysis about those discrepancies has been performed, which will yield further insight on the limitations of computational tools