Application of coupled 3D neutron kinetic / therma hydraulic tools and methods to the safety analysis of PWRs
Angelo Lo Nigro - Università degli Studi di Pisa - [2005]
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  • Tesi completa: 165 pagine
  • Abstract
    The need for coupling between three dimensional thermal hydraulic system codes and three dimensional neutron kinetic originates from the consideration that many physical phenomena occurring within nuclear systems cannot be properly described with simplified neutron kinetic models. Typical examples of such phenomena are local power excursions, asymmetric cool downs, boron dilutions, axial offset shifts. Coupled 3D TH / 3D NK codes and methods, which have been developed in order to specifically address these transients, presently constitute an advanced tool, the use of which allows the achievement of detailed results associated with largely reduced uncertainties. The inclusion of 3D neutron kinetic core models in the safety analyses allows the execution of Best Estimate assessments of the interactions between core and loops, which in turn allow precisely quantifying safety margins and optimizing plant systems, providing safer and more cost-effective NPP designs.
    The purpose of the activity that was carried out in the frame of this thesis was to address the application of coupled 3D THSC /3D NKC to safety analyses of PWRs. In particular, three main NPP types have been selected to cover a wide range of characteristics and problematics of the present and future generations of PWRs: Westinghouse AP1000, TMI-1 and VVER1000. These plants were analyzed using the digital computer code RELAP5-3D, developed by INEEL.
    The NSSS models of these NPPs, set up and qualified in the first part of the activity, were used to analyze a list of reference transients.
    Three main topics were addressed in particular during the activity:
    • the development and qualification of an analysis procedure, which allows the performance of adequate coupled 3D NKC-THSC calculations capable of achieving trustworthy results,
    • the assessment and characterization of the reactivity feedbacks due boron concentration changes in the primary reactor coolant system (RCS), and
    • the coupling between the RCS and in containment models, allowing a proper description of the thermal hydraulic behavior of the complex system made by the primary and secondary reactor cooling systems and the containment.

    All these topics were addressed using the infrastructure developed, in terms of methods, qualified nodalization, neutron kinetic and system codes. As outlined in chapter 4, the objective of rationalizing and improving the analysis procedure adopted to carry out 3D NKC-THSC calculation was followed. The classical analysis procedure was improved by adding some intermediate steps, which can be used to validate the coupling and qualify the cross section set. In particular, in order to improve the assessment of the real added value connected with the adoption of coupled codes and techniques, three standard exercises were defined, which allow the alignment, between 0D NK and 3D NK models, of the reactivity feedbacks due to moderator density, doppler effect and boron concentration.
    As far as the activity on the boron is concerned, a procedure to calcolate and to qualify a set of boron feedback coefficients was been developed and successfully applied to the TMI-1 core. In addition, the issue of possibile recriticality following SBLOCA in PWRs was addressed in a systematic manner: in a first phase, the amount of deborated water having the potential to damage the core was assessed, in a second phase, a full transient simulation of a SBLOCA, associated to deboration and recriticality was represented. The issue of a RIA following a SBLOCA was studied considering its importance, which has been increased by recent reactivity insertion event tests that indicated that high burn-up fuel may be more susceptible to reactivity events than previously expected, and fuel failure may occur at fuel enthalpy values
    that were previously judged acceptable. The analyses carried out showed that RELAP5-3D is capable of representing all the main physical phenomena which occur in the core. Integration with CFD codes and methods is clearly advisable to reduce uncertainties on the mixing phenomena occurring in the downcomer and in the lower plenum, which RELAP5-3D, as well as the other thermal hydraulic system codes, is not capable of properly representing.
    In addition, as a bypass result of the activity, quite a detailed investigation on the opportunity and possibility of including a thermal hydraulic model of the containment system in the safety analysis of the NSSS was performed. The outcome of the investigation was the development of a methodology, which is based on off line calculations and separate effect models and which provides a very smooth and realistic containment model behavior. This allows a proper description of the thermal hydraulic behavior of the complex system made by the primary and secondary reactor cooling systems and the containment.
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