Thermal–hydraulic phenomenology analysis and RELAP5/MOD3.3 code assessment for ATLAS 1% upper head SBLOCA test
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Inclou dades d'ús des de 2022
Cita com:
hdl:2117/334179
Tipus de documentProjecte Final de Màster Oficial
Data2020-07-20
Condicions d'accésAccés obert
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Abstract
In 2002, vessel head wall thinning that was identified at the Davis Besse reactor in the US shaded the lights on the importance of SB-LOCA’s in the upper head of RPV. The lower rates of coolant discharge through the break and primary circuit depressurization are the main characteristics in which SB-LOCA’s differ from LB-LOCA’s, and throughout the transient, three distinct and independent occurrences of core heat-up’s can potentially take place. SB-LOCA’s are troublesome scenarios for the development of EOP’s and modelling capabilities to adequately predict the overall T-H system response. This thesis contributes to the nuclear reactor safety from the thermal-hydraulic perspective, and it aims to develop and enhance a computational model of the ATLAS IET facility for the T-H system code RELAP5/MOD3.3. By applying the UPC-ANT Uncertainty Analysis methodology for code assessment, thermal-hydraulic phenomenology analysis is performed to analyze the model adequacy and accuracy to reproduce the most relevant T-H phenomena involved in Test B5.1 at the ATLAS facility, which is an SB-LOCA in the upper head. The work presented in the thesis is within the framework of UPC participation to the OECD/NEA ATLAS joint project 2. A variety of verification procedures are applied which focus on the statistical characteristics of sampled populations, the validity and robustness of the developed ATLAS-RELAP5 model by performing “null-transient” calculations. The implemented code assessment procedure is based on the adequacy of both the base-case and BEPU analysis calculations. It is concluded that the base-case calculation reproduces all the relevant T-H phenomena, either totally or partially. In essence, core heat-up and uncovery takes place ealier as compared to the experiment, about 900 seconds, due to quick clearing and refilling of loop-seals; thus, later events are accelerated in the base-case calculation. The BEPU analysis is generally adequate to perturb the relevant T-H phenomena and to envelope the experimental data sufficiently. T-H phenomena which are not “adequately” reproduced by the base-case model and are not “sufficiently” perturbed by the BEPU analysis reveal major and inherent model deficiencies and limitations, which include: (a) under-prediction of the downcomer collapsed water-level, maximum attained CET and PCT; (b) over-prediction of the strength of heat transfer to SG’s and natural circulation phenomena. The impact of uncertain input-parameters on a set of defined scalars is statistically analyzed. Those scalars represent some key T-H parameters characterizing the investigated scenario. Pearson’s and Spearman’s correlation analyses have shown that the break discharge coefficient is the most influenctial parameter. The adopted approach to treat code failures for some BEPU cases is validated based on statisical and phenomenological analyses. A list of potential model improvements for future development is provided, as inferred from the results of the implemented code assessment procedure.
MatèriesNuclear power plants -- Computer simulation, Nuclear power plants -- Safety measures, Centrals nuclears -- Simulació per ordinador, Centrals nuclears -- Mesures de seguretat
TitulacióMÀSTER UNIVERSITARI EN ENGINYERIA NUCLEAR (Pla 2012)
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alawad-tfm-submitted.pdf | 15,01Mb | Visualitza/Obre |