Collaborative Systems Thinking: Towards an Understanding of Team-level Systems Thinking

As the engineering workforce ages, skills with long development periods are lost with retiring individuals faster than are younger engineers developing the skills. Systems thinking is one such skill. Recent research, (Davidz 2006), has shown the importance of experiential learning in systems thinking skill development. However, an engineering career begun today has fewer program experiences than in past decades because of extended program lifecycles and a reduction in the number of new large-scale engineering programs. This pattern is clearly visible in the aerospace industry, which (Stephens 2003) cites as already experiencing a systems thinking shortage. The ongoing research outlined in this paper explores systems thinking as an emergent property of teams. Collaborative systems thinking, a term coined by the authors to denote teamlevel systems thinking, may offer an opportunity to leverage and develop a skill in short supply by concentrating on the team in addition to the individual. This paper introduces the proposed definition for collaborative systems thinking, as developed by the authors, and the outlines the structure and progress of ongoing case research into the role of organizational culture and standard process usage in the development of collaborative systems thinking

A review of radionuclide behaviour in the primary system of a very-high-temperature reactor

During normal operation of (V)HTRs radiologically-significant contamination of the primary system will occur this being of prime importance in relation to depressurization accidents. This paper reviews information relevant to radiocontaminant transport in (V)HTR primary systems paying particular attention to chemical forms, interactions with dust and overall distribution. The primary-system environment comprises nuclear graphites, alloys, dust and high-purity helium into which low releases of the isotopes 134Cs, 137Cs, 90Sr, 110mAg, 131I, 135Xe and 85Kr can be anticipated. Additionally, proper treatment of radiological risk requires accounting for tritium. A likely gas-phase speciation of the chemically-active fission products is proposed:-for caesium and strontium, hydroxides would be dominant with iodides as minor species if a relatively low iodine concentration can be assumed;-for iodine, a split between CsI and HI are likely to dominate with atomic iodine as a minor species. Strong sorption of radionuclides onto carbonaceous dust can be expected. Such dust is likely to cover all surfaces in a pebble-bed (V)HTR so radionuclides will principally associate with this dust rather than underlying alloys. This is unlikely in prismatic (V)HTRs with lower and uneven dust deposits. Where caesium interacts with alloys strong implanting of a large fraction can occur via adsorption and reaction with low-concentration silicon. Silver shows no special affinity for carbonaceous dust but may interact preferentially with nickel-rich alloys, i.e., in the IHX and/or the gas turbine. Quantitative evaluations of radionuclide distribution are hampered by a lack of data regarding sorption onto the graphites, alloys and carbonaceous dust of modern (V)HTR systems; a long time will elapse before sufficient data are forthcoming. In the meantime, some form of best-estimate distribution and upper-bound concentration for contamination is needed if deterministic safety evaluations are to begin. This distribution will be different for pebble-bed and prismatic designs. © 2009 Elsevier B.V. All rights reserved

On the nature of aerosols produced during a severe accident of a water-cooled nuclear reactor

Particle behaviour depends strongly on classic characteristics, e.g., size, and less macroscopic ones involving structure and composition these being especially important in situations of strong differential forces on a particle, i.e., surface impact or intensely-shearing flows. The former situation may lead to particle deposition or break-up and re-entrainment (with potential accident-management implications). This paper reviews information on aerosols from prototypical experiments identifying common features and typical variations. It emerges that a particle comprising one-third metal, one-third metal oxide and one-third a mixture of fission-product species would not be out of place in any potential reactor-accident sequence. Particle shapes appear relatively compact without branching chain-like structures. On size and structure, aerosols in the upstream part of the primary circuit would comprise a near-lognormal population with AMMD no more than 2 μm and geometric standard deviation around 2, particles comprising agglomerates of highly-coordinated clusters as small as 0.1 μm. In the containment, aerosols can typically be represented by primary-circuit particles and their agglomerates though particular circumstances (core-concrete interaction, hot-leg accident sequence) can alter this simple picture. © 2008 Elsevier B.V. All rights reserved

Interpretation of fission-product transport behaviour in the Phébus FPT0 and FPT1 tests

Phébus FP studies the phenomenology of severe accidents in water-cooled nuclear reactors. Tests cover fuel rod degradation and behaviour of fission-products released via the coolant system into the containment. Analysis using computer codes aims to identify modelling weaknesses. Regarding fission-product behaviour in the coolant circuit, analyses of tests FPT0 and FPT1 are presented that used a standard version of a code, SOPHAEROS, with input data based solely on measured boundary conditions. Disagreements between calculated and experimental results are explored and plausible explanations presented. It is shown that in laminar conditions for a supersaturated vapour with competing homogeneous nucleation, heterogeneous nucleation and condensation on structures, codes can significantly underestimate the latter if entrance effects are ignored. Where thermophoresis dominates in hydrodynamically developed, weakly turbulent flow, codes can overestimate deposition; the likely explanation is underestimated mechanical resuspension. Concerning the transport of vapour species, it is shown that observations are compatible with passage of non-negligible amounts of the gas hydrogen iodide through the circuit. The final aspect of the exercise concerns deposit remobilization where this was different in the two tests and the understanding of which remains more speculative. Explanation invokes vibration of the apparatus producing mechanical resuspension in FPT0 and steam reacting with caesium deposits producing caesium hydroxide in FPT1. © 2005 Elsevier B.V. All rights reserved

Comparison of LWR and SFR in-containment source term: Similarities and differences

The fundamental differences between Light-Water Reactors (LWRs) and Sodium Fast Reactors (SFRs) (i.e., breeding ratio, neutron energy spectrum, power densities, coolant nature, reactor architecture and so on) entail major differences in safety aspects. In particular, the in-containment source terms in the event of a severe accident differ in major ways. Nevertheless, an in-depth analysis of similarities and differences based on the present available knowledge should allow assessment of the applicability of LWR safety-analysis tools to the SFR domain. This is the final goal of this paper. A thorough literature review indicates that the capabilities of present LWR safety-analysis codes to address in-containment SFR accident scenarios are uneven for different areas: while they appear acceptable for predicting aerosol evolution, they require significant expansion in other areas like aerosol generation and fission-product partitioning. Additionally, some areas in need of further experimental research are highlighted in this paper. © 2013 Elsevier B.V. All rights reserved

Progress in chemistry modelling for vapour and aerosol transport analyses

The ASTEC/Sophaeros computer code has been developed to model transport of reactive gases and aerosols in pipe flows with extensive coverage of chemical and physical phenomena. As part of the ASTEC severe-accident code co-developed by IRSN in France and GRS (Gesellschaft für Anlagen-und Reaktorsicherheit) in Germany, it is principally used to evaluate so-called radioactive source terms released to the environment in the event of potential accidents affecting water-cooled nuclear reactors. A major uncertainty in the past concerned the impact of chemistry on radioactivity transfer where the treatment of chemical phenomena was recognized as inadequate. Work described here, involving mainly systematic identification of relevant species and thermodynamic-data verification, has vastly improved this situation leading in particular to creation of a comprehensive database with fully-referenced sources of information. Further prospective improvements in this area are described. © Carl Hanser Verlag GmbH & Co. KG ISSN 1862-5282

In-containment source term in accident conditions in sodium-cooled fast reactors Data needs and model capabilities

International audienceSodium-cooled fast reactors (SFRs) are one out of the six technologies considered in the so-called Generation IV initiative. A full-scope safety analysis of this reactor type would need to have computation tools developed and properly validated. The present paper focuses on the review of currently available data and modeling capabilities for in-containment source term analysis in accident conditions. Generally speaking, it has been found that improvements to characterizing particles (i.e., density and shape) and particle-particle interaction processes are required. However, beyond any doubt, two of the strongest modeling needs for SFR accidents are aerosol generation which, in turn, means to properly model sodium vaporization, chemical reactions with the surrounding gas, nucleation of combustion products and primary particle agglomeration; and fission products partitioning. Current LWR integral codes, although not yet furnished with such models, look promising as a computational platform to be extended to the SFR domain. In any case, once they are adapted to anticipated SFR scenarios, extensive validation should be undertaken against a comprehensive and sound database. © 2011 Elsevier Ltd. All rights reserved

Modelling of fission-product transport in the reactor coolant system

The Phébus fission product (FP) programme studies the phenomenology of severe accidents in water-cooled nuclear reactors Five tests were performed in the frame of the programme covering fuel-rod degradation and FP behaviour released via the coolant system into the containment To model FP transport and behaviour in the coolant system, numerous physical and chemical phenomena have to be taken into account In the vapour phase, for example, FP speciation, vapour condensation and vapour/surface or vapour/aerosol reactions have to be considered The aerosol phase has to be modelled with nucleation, growth and deposition processes Finally, remobilisation phenomena like resuspension and revaporisation have to be taken into account for delayed release into the containment Four Phébus FP tests (FTP0, FPT1, FPT2, FPT3) have been modelled with the ASTEC/SOPHAEROS code Modelling shows an overall good estimation of retention for the main FPs (e.g.; I, Cs, Mo) Furthermore, a strong connection is revealed in the gaseous phase chemistry between I, Cs, Cd and Mo which has a great impact on gaseous iodine release into the containment The Phébus FP test modelling also exposes disagreement on FP retention when laminar gaseous flow is not well developed Finally, probably the most significant shortcoming in modelling that Phébus-FP tests highlighted concerns vapour-phase iodine-chemistry modelling at low temperature The study of this latter point is on going with the experimental programme ISTP/CHIP © 2013 Elsevier B.V All rights reserved

Progress in understanding fission-product behaviour in coated uranium-dioxide fuel particles

Supported by results of calculations performed with two analytical tools (MFPR, which takes account of physical and chemical mechanisms in calculating the chemical forms and physical locations of fission products in UO2, and MEPHISTA, a thermodynamic database), this paper presents an investigation of some important aspects of the fuel microstructure and chemical evolutions of irradiated TRISO particles. The following main conclusions can be identified with respect to irradiated TRISO fuel: first, the relatively low oxygen potential within the fuel particles with respect to PWR fuel leads to chemical speciation that is not typical of PWR fuels, e.g., the relatively volatile behaviour of barium; secondly, the safety-critical fission-product caesium is released from the urania kernel but the buffer and pyrolytic-carbon coatings could form an important chemical barrier to further migration (i.e., formation of carbides). Finally, significant releases of fission gases from the urania kernel are expected even in nominal conditions. © 2008 Elsevier B.V. All rights reserved

Investigation of thermo-catalytic decomposition of metal-iodide aerosols due to passage through hydrogen recombiners

Passive autocatalytic recombiners (PARs) are a means of preventing hydrogen accumulation in the containment building of a water-cooled nuclear reactor during an accident. A potential problem exists concerning suspended radioactive aerosols: particles passing through the catalytic elements are heated up with significant evaporation of more volatile chemical species. The aerosols, vapours and carrier-gas mixture may chemically react resulting potentially in conversion of easily-retained aerosol material into more troublesome vapours and gases. An experimental programme, RECI, demonstrated that potential exists for PARs to generate volatile forms of iodine, namely molecular iodine, by thermo-catalytic decomposition of metal-iodide aerosols. Here, analysis of RECI results aided by two computer codes, one a field code the other a lumped-parameter approach, provides significant insight into the iodide-iodine phenomenology where, in particular, the rapid cooling of the reacting mixture explains the persistence of volatile species downstream at ambient conditions. While understanding of the phenomenology has progressed, the current results cannot be extrapolated to the reactor case since further experiments are needed reproducing more closely expected accident conditions. © 2009 Elsevier B.V. All rights reserved