1,721,121 research outputs found
Progress in modeling in-containment source term with ASTEC-Na
No full textThe ASTEC-Na code is being developed to simulate any sort of postulated accidents in Na-cooled fast reactors, particularly severe accidents. A significant progress has been made in the development of source term models, their implementation in the code and the specific validation of the specific code module, hereafter named CPA∗, under the auspices of the JASMIN project. In this paper the fundamentals of models for Na pool fire thermal-hydraulics and particle generation and chemical ageing of airborne particulates are described. Based mostly on data gathered from the open literature, CPA∗ performance under conditions anticipated during Na pool fires has been assessed against AB1, AB2, F2 and EMIS10 experiments. Thermal-hydraulic estimates have shown acceptable generic trends with noticeable quantitative deviations, despite the highly parametrized models used. A similar statement concerning aerosol behavior following measurements tendency is also applicable. As for chemical ageing, the comparisons set indicate that further work is still necessary. Therefore, even though some significant progress has been achieved, it is unquestionable that further work needs to be done in the three areas addressed. Finally, it should be highlighted that one of the main outcomes of this work is the need of obtaining qualified data for models and codes validation, so that a thorough and sound model assessment and code validation can be conducted
In-containment source term predictability of ASTEC-Na: Major insights from data-predictions benchmarking
No full textModeling the containment response to a sodium pool fire is to be one of the key aspects of any comprehensive safety evaluation of the new generation of sodium cooled fast reactors. Through a peer review of earlier experimental investigations some useful data can be collected and then used for assessing the current analytical capabilities to model severe accidents or some of their specific aspects. This paper provides major insights into the in-containment aerosol behavior predictability of ASTEC-Na (CPA∗ module) during Na-pool fires. By comparing against tests from the ABCOVE (AB1 and AB2) and FAUNA (F2) programs, it has been shown that experimental trends can be roughly reproduced with a single-cell approach whenever natural convection is effective in making the vessel atmosphere uniform both thermally and in composition. Nonetheless, the present heavy parametrization of ASTEC-Na models should be avoided or strongly supported by further experimentation that allows setting sound default values, concerning both combustion energy distribution and aerosol formation and distribution. Anyway, the peer data review has highlighted that a meaningful comparison to predictions is not always feasible due to large data uncertainties, particularly at the beginning of Na burning. As for the particle ageing, the comparisons set seems to indicate that transformation from oxides to hydroxides is predicted to be too slow; nevertheless, a more extensive benchmarking should be conducted to confirm it. © 2017 Elsevier B.V
Summary of the activities and analyses carried out using USNRC CODES in 2022
No full textThe present document summarizes the ENEA research activities and the analyses conducted adopting USNRC codes in the year 2022
IAEA International Collaborative Standard Problem on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents
No full textSummary of the activities and analyses carried out using USNRC codes in 2023
No full textThe present document summarizes the ENEA research activities and the analyses conducted adopting USNRC codes in the year 2023
Preface for the special issue “NFT-05: Italy and Greece”: nuclear fission technology in Italy
No full textAfter the successful experience in the ‘60s and ‘70s, a bright future for nuclear energy suddenly became opaque after the Chernobyl accident; however, nuclear energy still can hold its promises. Noticeably, some research continued after that event, and valuable expertise has persisted in industry, research centers, and universities. The commitment to research on the safety aspects of current reactor generations and advanced reactors using passive safety systems should be underlined. Even though Large Reactor Units (LRUs) have demonstrated the capability to benefit a country like Italy, Small Modular Reactor (SMR) technology might have a role in fulfilling the zero carbon target. In fact, in the short term, light water SMR designs appear to be a target of the strategy for the Italian government to fulfil the zero-carbon target. Notably, there is also interest in the future deployment of Gen-4 reactors, particularly lead-cooled fast reactors, considering the expertise in the country. Greek scientists did not participate in the activities of the present VSI. Therefore, the discussion below deals with Italy only
Analyses with MELCOR code of an unmitigated SBO scenario with in vessel retention strategy applied to a generic PWR 900 MWe
No full textIn the framework of the In Vessel Melt Retention (IVMR) project, coordinated by IRSN, funded by Horizon 2020 Framework Programme of the European Commission and started in June 2015, ENEA is involved in the development of a generic PWR-900 analyses with MELCOR 2.2 code for benchmarking ASTEC code in relation to the phenomenology taking place when In Vessel Retention (IVR) strategy is applied. The IVR strategy for LWR is based on external reactor vessel cooling, by water flooding of the reactor cavity before vessel failure, aiming to avoid lower head failure and maintain the corium in the RPV. The severe accident scenario selected for this MELCOR 2.2 analyses with IVR application is an unmitigated Station Blackout (SBO). The target of this paper is to present the analyses of the main figure of merits related to this kind of transient (e.g. primary pressure, max intact cladding temperature, hydrogen production, zircaloy fraction oxidized, etc) and the corium degradation evolution. Particular attention will be focused on the characterization of the corium composition behaviour in the lower plenum and the consequent lower head thermal attack (e.g. axial profile along the lower head wall) that, through selected figure of merits (e.g. corium physical characteristics in the LP, maximum heat flux along the LH wall) will be compared also with the results obtained by ENEA_ASTEC calculation to show the differences between the codes due to the uncertainties on core degraded configurations implemented in severe accident codes and on the way the core relocates to the L
Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool
No full textThe pbenomenological analyses and thermal hydraulic characterization of a nuclear reactor arc the basis for its design and safety evaluation. In light of the impossibility and huge cost of performing meaningful experiments at full scale, sealed down expenmental tests - Integral Effect Test (IET) and Separate Effect Test (SET) • arc more feasible in developing "assessment database". The data arc useful in characterizing the prototype design and in the validation of computational tools for safety analysis. The analyses of system behaviors including component interactions in the Reactor Coolant System (RCS). the Containment System (PCV) and the RCS.'PCV coupled system have been extensively investigated using lETs in the past decades. Though several scaling methods, e.g. Linear, Power/Volume, Three level scaling, H2TS..., have been developed and applied in the IF.T and SET design, a direct extrapolation of the data to the prototype, i.e. the scalability, is in general not possible due to unavoidable scaling distortions. The scaling distortions arc related to many factors, mainly the complex geometry, multiple component interactions and two phase thermal hydraulic phenomena in steady state and transient condition of a nuclear reactor. The complex nature of scaling a nuclear reactor requires a large number of scaling parameters to be simultaneously fulfilled. In addition, physical construction and funding constraints demand that a scaling compromise is inevitable. Therefore a scaling approach, e.g. time preserved'not preserved, full height/reduced height, full pressure/reduced pressure, full power/reduced power..., has to be adopted in accordance with the objective of the IET or SET. Together with the scaling analysis. Best Estimate (BE) thermal hydraulic system code has been used for supporting experiment activity (design facilities, interpretation of results, etc) and for extrapolating results to full scale prototype conditions. Since the closure laws in the system code arc mainly based on scaled test data, the extrapolation of code results remains a challenging and open issue. Starting from a brief analysts of the main characteristics of IF.Ts and SETFs. the main objective of this paper is to analyze some IET scaling approaches used to the simulation of RCS responses which characterize the main scaling limits. The scaling approaches and their constraints in ROSA-III. FIST and PIPER-ONE facility will be used to analyze their impact to the experimental prediction in Small Break LOC'A counterpart tests. The liquid level behavior in the core and the core cladding temperature analysis are discussed used as judging criteria for the facilities sealing-up limits
The EC MUSA project on management and uncertainty of severe accidents: Main pillars and status
No full textIn the current state of maturity of severe accident codes, the time has come to foster the systematic application of Best Estimate Plus Uncertainties (BEPU) in this domain. The overall objective of the HORIZON-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” is to quantify the uncertainties of severe accident codes (e.g., ASTEC, MAAP, MELCOR, and AC2) when modeling reactor and spent fuel pools accident scenarios of Gen II and Gen III reactor designs for the prediction of the radiological source term. To do so, different Uncertainty Quantification (UQ) methodologies are to be used for the uncertainty and sensitivity analysis. Innovative AM measures will be considered in performing these UQ analyses, in addition to initial/boundary conditions and model parameters, to assess their impact on the source term prediction. This paper synthesizes the major pillars and the overall structure of the MUSA project, as well as the expectations and the progress made over the first year and a half of operation
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