Borut Mavko - Academia.edu (original) (raw)

Papers by Borut Mavko

With the suggested guide-words approach connected to the critical components, a different viewpoi... more With the suggested guide-words approach connected to the critical components, a different viewpoint on nuclear safety attitudes is defined. This enables the identification, judgment, and improvement of the most vulnerable places in the plant. Any potential overlap in the duties and areas where a clear division of responsibilities is needed is thus revealed. Also, the need for communication between different

Proceedings of the …, 2003

... Title: Education and training in nuclear engineering and safety (NEPTUNO) and lessons learnt ... more ... Title: Education and training in nuclear engineering and safety (NEPTUNO) and lessons learnt from the "European Nuclear Education Network" (ENEN). Authors: Moons, F Safieh, J Por, G Lechak, T Mavko, B Raj Sehgal, B D'haeseleer, William. Issue Date: Nov-2003. ...

Transactions of the American Nuclear Society, 1992

ABSTRACT

Nuclear Engineering and Design, Oct 1, 2020

In view of many discussions on rethinking graduate education to better support the non-academic c... more In view of many discussions on rethinking graduate education to better support the non-academic careers, we have reviewed the doctoral theses, completed within graduate studies of nuclear engineering at the Jozef Stefan Institute and the University of Ljubljana (Slovenia) between 1993 and 2019. The theses are divided (in the present paper) in three groups: "basic", "applied" and "application-oriented basic". The merits of each kind are discussed, including from the point of view of nuclear engineers intending to continue their careers in the nuclear industry. As a specific case, the topics studied were reactor thermal hydraulics, severe accidents, ageing and integrity of components, and probabilistic safety assessment. The review shows a strong preponderance of basic and application-oriented basic theses over applied ones. It also appears that the predominantly academic orientation of the doctoral theses enabled not only academic, but also industrial careers.

Two SGTR transient scenarios with loss of secondary heat sink have been analyzed and compared usi... more Two SGTR transient scenarios with loss of secondary heat sink have been analyzed and compared using RELAP5/MOD3.1 thermal-hydraulic computer code. The findings of the two analyses show that the initiating event, causing the loss of heat sink influences operator actions and further course of the transient significantly. If main feedwater is lost first this leads to much different plant behavior as if reactor and turbine are tripped first. Nevertheless, both analyses have proven that if the operator ignored loss of secondary heat sink in the beginning of the transient and mitigated the consequences of SGTR first, leads to a successful outcome by applying the Feed and Bleed procedure after the SG dryout. The analyses have been performed for Krsko NPP, a two-loop Westinghouse PWR, 640 MWe, located in Slovenia.

Classifications of small-break loss-of-coolant accidents based on objective quantitative similari... more Classifications of small-break loss-of-coolant accidents based on objective quantitative similarity analysis are proposed. Accident scenarios were simulated in a two-loop pressurized water reactor plant with the RELAP5/MOD3.1 computer code for break sizes ranging from 1.27 cm (0.5 in.) to 15.2 cm (6 in.), with different availability of auxiliary feedwater system or reactor coolant pump trip delay. Similarities between different accident simulations were evaluated by comparing relevant time-dependent parameters with fast Fourier transform and correlation methods. Quantification of similarity between accident simulations could eventually lead to further development of the Code Scaling, Applicability and Uncertainty methodology.

The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and t... more The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and the ability of RELAP5/MOD3.2.2 Gamma to predict core overheating. The code prediction has been compared to the three experiments, one conducted on the separate effect test facility NEPTUN in Switzerland and the other two conducted on two integral test facilities, PMK-2 in Hungary and PACTEL facility in Finland. In the case of a series of boiloff experiments performed on the NEPTUN test facility the influence of the two correlations available in MOD3.2.2 Gamma for determining interphase drag has been studied. In the case of IAEA-SPE-4 experiment simulation on PMK-2 facility the main goal of the analysis was to study the adequate modeling of the hexagonal core channel with 19-rod bundle and the phenomena during the core uncovering. The third analyzed experiment, OECD-ISP-33, was performed on PACTEL facility to study different natural circulation modes during SB LOCA. The analysis also focused on the final stage of this SB LOCA experiment, when core dryout and heatup was observed due to gradual emptying of the primary system. Following the experience the appropriate modeling options have been used to achieve better representation of the important phenomena during the SB LOCA. 1 INTRODUCTION Certain deficiencies in the RELAP5/MOD3.2.x code [1] have revealed in the past few years, concerning the prediction of reactor core heatup in the late phases of LOCA. According to previous experience in modeling separate and integral effect tests ([2], [3], [4], [5], [6] and [7]) it was assumed that it should be sufficient to set up an adequate simulation model and make the appropriate selection of interphase drag and heat transfer correlation to successfully model the LOCA phenomena with RELAP5/MOD3.2.2 Gamma [1]. Special attention was paid to the core region model in order to capture the phenomena during core uncovering and heatup correctly and to prove the above hypothesis.

The test E11.2 "Hydrogen distribution in loop flow geometry", which was performed in the Heissdam... more The test E11.2 "Hydrogen distribution in loop flow geometry", which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.

Journal of heat transfer, Mar 16, 2001

This paper provides the scaling-up methodology that was applied from the BETHSY integral test fac... more This paper provides the scaling-up methodology that was applied from the BETHSY integral test facility to the real Framatome NPP (Nuclear Power Plant). The ISP-27 (International Standard Problem) transient scenario was used, based on test 9.1b. The objectives were to evaluate the ideal scaling-up of BETHSY facility for real NPP and to compare and analyse the physical phenomena known from experimental background with the phenomena predicted by RELAP5/MOD3.2 simulation of real NPP transient. Further, in order to test phenomenological scaling-up basis two models for RELAP5/MOD3.2 code were constructed differing in scaling criteria for the primary cooling system. Special attention was concentrated on heat structures scaling. Data were analysed through plotting plots and NPA’s (Nuclear Plant Analyzer) graphical presentation.

Proceedings of the …, 2003

Transactions of the American Nuclear Society, 1992

ABSTRACT

Nuclear Engineering and Design, Oct 1, 2020

Journal of heat transfer, Mar 16, 2001

Heat transfer coefficient in the fully developed turbulent channel flow bounded with walls heated... more Heat transfer coefficient in the fully developed turbulent channel flow bounded with walls heated with constant heat flux, is known to be a function of Reynolds and Prandtl number only (if the temperature is assumed to be a passive scalar). The present study, based on the results of the Direct Numerical Simulations reports a small difference of approximately 0.5% that exists in heat transfer rates near the wall of fully developed turbulent channel flow at the same Reynolds and Prandtl numbers, the same heat flux, but with different thermal activity ratios.

Update of MOD3.3 assessment calculations against two real transients at Krško NPP, caused by MSIV... more Update of MOD3.3 assessment calculations against two real transients at Krško NPP, caused by MSIV 1 and MSIV 2 inadvertent closure.

Assessment of RELAP5/MOD3.3 for fast transients (water hammer).

Numerical analysis of natural convection inside a heat generated fluid was performed for four dif... more Numerical analysis of natural convection inside a heat generated fluid was performed for four different spherical geometries that match the experimental vessels used by Asfia et al. [5-7]. The transient calculations were performed with the CFX 5.7 fluid dynamic software. The simulations show that the highest heat flux is just below the rim of the cavity and it can be 50 times higher than at the bottom. Based on the numerical results, the local values of heat transfer coefficient and the distributions of global Nusselt number were calculated. The present, three-dimensional simulation results were compared with the numerical results of Mayinger et al. [3] and Reineke et al. [4], and with the experimental data of Asfia et al. [5-7]. The agreement between the results that is well inside the experimental scatter verifies the selected modeling approach.