HiRel: Hybrid Automated Reliability Predictor (HARP) integrated reliability tool system, (version 7.0). Volume 2: HARP tutorial

The Hybrid Automated Reliability Predictor (HARP) integrated Reliability (HiRel) tool system for reliability/availability prediction offers a toolbox of integrated reliability/availability programs that can be used to customize the user's application in a workstation or nonworkstation environment. The Hybrid Automated Reliability Predictor (HARP) tutorial provides insight into HARP modeling techniques and the interactive textual prompting input language via a step-by-step explanation and demonstration of HARP's fault occurrence/repair model and the fault/error handling models. Example applications are worked in their entirety and the HARP tabular output data are presented for each. Simple models are presented at first with each succeeding example demonstrating greater modeling power and complexity. This document is not intended to present the theoretical and mathematical basis for HARP

Results of the QUENCH-20 experiment with BWR test bundle [in press]

The experiment QUENCH-20 with BWR geometry simulation bundle was successfully conducted at KIT on 9th October 2019 in the framework of the international SAFEST project. The test bundle mock-up represented one quarter of a BWR fuel assembly with 24 electrically heated fuel rod simulators and two B4C control blades. The rod simulators were filled with Kr to an inner pressure of 5.5 bar. The pre-oxidation stage in a flowing gas mixture of steam and argon (each 3 g/s) and system pressure of 2 bar lasted 4 hours at the peak cladding temperature of 1250 K. The Zry-4 corner rod, withdrawn at the end of this stage, showed the maximal oxidation at elevations between 930 and 1020 mm with signs of breakaway. During the transient stage, the bundle was heated to a maximum temperature of 2000 K. The coolability of the bundle was decreased by its squeezing due to the shroud ductile deformation caused by an overpressure outside the shroud. The cladding radial strain and failures due to inner overpressure (about 4 bar) were observed at temperature about 1700 K and lasted about 200 s. During the period of rod failures also the first absorber melt relocation accompanied by shroud failure were registered. The interaction of B4C with the steel blade and the ZIRLO channel box were observed at elevations 650…950 mm with the formation of eutectic melt. The typical components of this melt are (Fe, Cr) borides and ZrB2 precipitated in steel or in Zr-steel eutectic melt. Massive absorber melt relocation was observed 50 s before the end of transition stage. Small fragments of the absorber melt moved down to the elevation of 50 mm. The melting point of Inconel spacer grids at 500 and 1050 mm was also reached at the end of the transition stage. The Inconel melt from the elevation 1050 mm relocated downwards through hot bundle regions to the Inconel grid spacer at 550 mm and later (during the escalation caused by quench) to 450 mm. This melt penetrated also under the damaged cladding oxide layer and formed molten eutectic mixtures between elevations 450 and 550 mm. The test was terminated by quench water injection with a flow rate of 50 g/s from the bundle bottom. Fast temperature escalation from 2000 to 2300 K during 20 s was observed due to the strongly exothermic oxidation reactions. As result, the metal part (prior β-Zr) of the claddings between 550 and 950 mm was melted, partially released into space between rods and partially relocated in the gap between pellet and outer oxide layer to 450 mm. In this case, the positive role of the oxide layer should be noted, which does not allow the melt to completely escape into the inter-rod space. It is thereby limiting the possibility of interactions of a large amount of melt with steam, which could significantly increase the exothermic oxidation processes and the escalation of temperatures. The distribution of the oxidation rate within each bundle cross section is very inhomogeneous: whereas the average outer ZrO2 layer thickness for the central rod (#1) at the elevation of 750 mm is 465 µm, the same parameter for the peripheral rod #24 is only 108 µm. The average oxidation rate of the inner cladding surface (due to interaction with steam and with ZrO2 pellets) is about 20% in comparison to the outer cladding oxidation. The bundle elevations 850 and 750 mm are mostly oxidized with average cladding ECR 33%. The oxidation of the melt relocated inside the rods was observed at elevations 550…950 mm. The mass spectrometer measured release of CO (12.6 g), CO2 (9.7 g) and CH4 (0.4 g) during the reflood as products of absorber oxidation; the corresponding B4C reacted mass was 41 g or 4.6% of the total B4C inventory. It is significantly lower than in the PWR bundle tests QUENCH-07 and QUENCH-09 containing central absorber rod with B4C pellets inserted into a thin stainless steel cladding and Zry-4 guide tubes (20% and 50% reacted B4C correspondingly). Hydrogen production during the reflood amounted to 32 g during the reflood (57.4 g during the whole test) including 10 g from B4C oxidation

The working group on the analysis and management of accidents (WGAMA): A historical review of major contributions

The Working Group on the Analysis and Management of Accidents (WGAMA) was created on December 31st, 1999 to assess and strengthen the technical basis needed for the prevention, mitigation and management of potential accidents in NPP and to facilitate international convergence on safety issues and accident management analyses and strategies. WGAMA addresses reactor coolant system thermal-hydraulics, in-vessel behaviour of degraded cores and in-vessel protection, containment behaviour and containment protection, and fission product (FP) release, transport, deposition and retention, for both current and advanced reactors. As a result, WGAMA contributions in thermal-hydraulics, computational fluid-dynamics (CFD) and severe accidents along the first two decades of the 21st century have been outstanding and are summarized in this paper. Beyond any doubt, the Fukushima-Daiichi accident heavily impacted WGAMA activities and the substantial outcomes produced in the accident aftermath are neatly identified in the paper. Beyond specific events, most importantly, around 50 technical reports have become reference material in the different fields covered by the group and they are gathered altogether in the reference section of the paper; a common outstanding feature in most of these reports is the recommendations included for further research, some of which have eventually given rise to some of the projects conducted or underway within the OECD framework. Far from declining, ongoing WGAMA activities are numerous and a number of them is already planned to be launched in the near future; a short mention to them is also included in this paper

HiRel: Hybrid Automated Reliability Predictor (HARP) integrated reliability tool system, (version 7.0). Volume 1: HARP introduction and user's guide

The Hybrid Automated Reliability Predictor (HARP) integrated Reliability (HiRel) tool system for reliability/availability prediction offers a toolbox of integrated reliability/availability programs that can be used to customize the user's application in a workstation or nonworkstation environment. HiRel consists of interactive graphical input/output programs and four reliability/availability modeling engines that provide analytical and simulative solutions to a wide host of reliable fault-tolerant system architectures and is also applicable to electronic systems in general. The tool system was designed to be compatible with most computing platforms and operating systems, and some programs have been beta tested, within the aerospace community for over 8 years. Volume 1 provides an introduction to the HARP program. Comprehensive information on HARP mathematical models can be found in the references

CoreSOAR Core Degradation State-of-the Art Report Update: Conclusions [in press]

In 1991 the CSNI published the first State-of-the-Art Report on In-Vessel Core Degradation, which was updated to 1995 under the EC 3rd Framework programme. These covered phenomena, experimental programmes, material data, main modelling codes, code assessments, identification of modelling needs, and conclusions including the needs for further research. This knowledge was fundamental to such safety issues as in-vessel melt retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. In the last 20 years, there has been much progress in understanding, with major experimental series finished, e.g. the integral in-reactor Phébus FP tests, while others have many tests completed, e.g. the electrically-heated QUENCH series on reflooding degraded rod bundles, and one test using a debris bed. The small-scale PRELUDE/PEARL experiments study debris bed quench, while LIVE examines melt pool behaviour in the lower head using simulant materials. The integral severe accident modelling codes, such as MELCOR and MAAP (USA) and ASTEC (Europe), encapsulate current knowledge in a quantitative way. After two EC-funded projects on the SARNET network of excellence, continued in NUGENIA, it is timely to take stock of the vast range of knowledge and technical improvements gained in the experimental and modelling areas. The CoreSOAR project, in NUGENIA/SARNET, drew together the experience of 11 European partners to update the state of the art in core degradation, finishing at the end of 2018. The review covered knowledge of phenomena, available integral experiments, separate-effects data, modelling codes and code validation, then drawing overall conclusions and identifying needs for further research. The final report serves as a reference for current and future research programmes concerning core degradation in NUGENIA, in other EC research projects such as in Horizon2020 and for projects under the auspices of OECD/NEA/CSNI

Study on thermal fragmentation characteristics of a superheated alumina droplet

International audienceIn the frame of the European Commission FP7 SAFEST project, IRSN proposed to experimentally investigate the steam explosion triggering mechanisms of a superheated alumina droplet falling into water, through a set of experiments in the Micro Interactions in Steam Explosion Energetics facility (MISTEE) at KTH. Since thermal fragmentation is considered to be a likely process for the triggering of Steam Explosions in the KROTOS tests (performed at CEA) with alumina, the ability of a single droplet of such material to undergo thermally induced fine fragmentation is studied on the MISTEE facility with a close-up visualization. A series of experiments were conducted, where droplets of molten alumina were discharged into a water pool and potentially exposed to a small pressure wave. The intense interactions were recorded with a high-speed camera along with the pressure in the droplet vicinity. The ability of alumina to undergo thermal fragmentation is expected to be firstly contingent on the stability of the vapour film enshrouding the melt droplet. The water and melt temperatures may then play a crucial role on the vapour film stability, and therefore on the observation of a steam explosion. Indeed, under high to moderate water sub-cooling conditions, experimental observations indicate that fine fragmentation of the melt can occur when the droplet is exposed to even a weak pressure wave, in the range of 0.15 MPa. In contrast, melt fine fragmentation is suppressed at low water sub-cooling conditions (less than 30 °C), where the formation of a thick vapour film (and large wake) is observed, and which is probably too stable to be destabilized by the weak pressure wave. The effect of the melt temperature on thermal fragmentation is also assessed. This parameter influences the solidification of the droplet and the strength of the explosion as it determines the available heat energy. In the present conditions, fine fragmentation of melt occurred even at quite low melt superheat (≈60 °C). For a high melt superheat (above 200 °C) a very energetic spontaneous steam explosion was observed. A physical analysis on the debris particles acquired indicates a mass median diameter of ≈100 µm, comparable to the one observed in the KROTOS alumina experiments. The MISTEE experimental results are finally used to assess the heat and mass transfer modelling of the coolant during the fragmentation process in the FCI code MC3D. © 2018 Elsevier Lt