@inproceedings{MTMT:31618839,
	title = {Development of nuclear power plant basic principle simulators as tools for nuclear engineering education},
	url = {https://m2.mtmt.hu/api/publication/31618839},
	author = {Fehér, Sándor and Czifrus, Szabolcs and Lőrincz, J. and Varju, Tamás and Aszódi, Attila and Csom, Gyula},
	booktitle = {Varietas delectat... Complexity is the new normality},
	unique-id = {31618839},
	abstract = {Our Institute has been developing PC-based nuclear power plant (NPP) basic principle simulators for over 30 years. The main goal of this development is to provide the nuclear engineering education at the university with tools capable to illustrate the fundamental physical processes going on in the primary and secondary circuit of an NPP. The simulators were integrated into our training courses in the last three decades. The firstly developed primary circuit basic principle simulator, PC2 is still the most often and most widespread used one despite the fact that it was written for old DOS operating system and thus nowadays is of restricted applicability. Therefore, the development of a new PC2+ simulator started in 2016 based on the training experience accumulated and with support of a national research grant. Its scope of simulation is practically the same as that of the old PC2 but it has various model extensions and new features. The graphical user interface (visualization/interaction) program module and the simulation model computing module are two separate programs (executable codes) which may run on a single or on two different computers as well. This feature of PC2+ allows for using it via internet and opens new possibilities both in classroom exercises and distant learning. © 2020 SEFI 47th Annual Conference: Varietas Delectat... Complexity is the New Normality, Proceedings. All rights reserved.},
	keywords = {Nuclear power plants; Nuclear energy; Graphical user interfaces; Simulators; Distance education; Nuclear fuels; Engineering education; Simulation model; circuit simulation; Thermal hydraulics; Thermal hydraulics; Reactor physics; Reactor physics; Primary circuits; Nuclear power plant simulator; Primary circuit; Distant learning; Power plant simulator; Secondary circuit; Training experiences},
	year = {2020},
	pages = {380-389},
	orcid-numbers = {Varju, Tamás/0000-0002-2969-1867}
}

@article{MTMT:31449391,
	title = {Fast Neutron Irradiation Optimization of Thorium-Fueled SCWR Reactor Pressure Vessel},
	url = {https://m2.mtmt.hu/api/publication/31449391},
	author = {Ponya, Petra and Csom, Gyula and Fehér, Sándor},
	doi = {10.1115/1.4046354},
	journal-iso = {J NUCLEAR ENG RAD SCI},
	journal = {JOURNAL OF NUCLEAR ENGINEERING AND RADIATION SCIENCE},
	volume = {6},
	unique-id = {31449391},
	issn = {2332-8983},
	abstract = {Fast neutron irradiation causes embrittlement of the reactor pressure vessel (RPV) material; therefore, it may end operation life before design lifetime. Well-known method to recuperate crystal lattice dislocations is annealing. In the current version of thorium fueled supercritical water-cooled reactor (SCWR) design proposed by the Institute of Nuclear Technology at Budapest University of Technology and Economics (BME NTI), the supercritical fluid flows upward between the core barrel and the inner surface of the RPV thereby, the coolant would keep the RPV's temperature at similar to 500 degrees C. This reverse coolant flow direction would decrease the embrittlement of RPV by constant annealing. To minimize the fast neutron flux increase, a relatively thin shielding connected to the inner surface of the barrel could be used. This presents fast neutron irradiation analysis, performed for different settings of the shielding to reduce fast neutron flux reaching the inner surface of RPV.},
	year = {2020},
	eissn = {2332-8975}
}

@inproceedings{MTMT:32796964,
	title = {Fuel cycle studies of Generation IV reactors with the SITON v2.0 code and the FITXS burn-up scheme},
	url = {https://m2.mtmt.hu/api/publication/32796964},
	author = {Halász, Máté Gergely and Brolly, Áron Péter and Szieberth, Máté and Fehér, Sándor},
	booktitle = {Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17)},
	unique-id = {32796964},
	year = {2018},
	orcid-numbers = {Szieberth, Máté/0000-0001-9192-0352}
}

@article{MTMT:3134362,
	title = {Physical model of the nuclear fuel cycle simulation code SITON},
	url = {https://m2.mtmt.hu/api/publication/3134362},
	author = {Brolly, Áron Péter and Halász, Máté Gergely and Szieberth, Máté and Nagy, L and Fehér, Sándor},
	doi = {10.1016/j.anucene.2016.10.001},
	journal-iso = {ANN NUCL ENERGY},
	journal = {ANNALS OF NUCLEAR ENERGY},
	volume = {99},
	unique-id = {3134362},
	issn = {0306-4549},
	year = {2017},
	eissn = {1873-2100},
	pages = {471-483},
	orcid-numbers = {Szieberth, Máté/0000-0001-9192-0352}
}

@article{MTMT:3201111,
	title = {FITXS: A fast and flexible burn-up scheme based on the fitting of one-group cross-sections},
	url = {https://m2.mtmt.hu/api/publication/3201111},
	author = {Halász, Máté Gergely and Szieberth, Máté and Fehér, Sándor},
	doi = {10.1016/j.anucene.2017.02.010},
	journal-iso = {ANN NUCL ENERGY},
	journal = {ANNALS OF NUCLEAR ENERGY},
	volume = {104},
	unique-id = {3201111},
	issn = {0306-4549},
	keywords = {Breeding; Fuel cycle; transmutation; Gas-cooled Fast Reactor; Burn-up; Cross-section parametrization},
	year = {2017},
	eissn = {1873-2100},
	pages = {267-281},
	orcid-numbers = {Szieberth, Máté/0000-0001-9192-0352}
}

@article{MTMT:3120750,
	title = {Analysis of heat source distribution in internal circulating surface heat transfer molten salt reactor},
	url = {https://m2.mtmt.hu/api/publication/3120750},
	author = {German, Péter and Fehér, Sándor and Czifrus, Szabolcs and Yamaji, Bogdán},
	doi = {10.1016/j.pnucene.2016.07.018},
	journal-iso = {PROG NUCL ENERG},
	journal = {PROGRESS IN NUCLEAR ENERGY},
	volume = {92},
	unique-id = {3120750},
	issn = {0149-1970},
	abstract = {The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods. © 2016 Elsevier Ltd},
	keywords = {DEPOSITION; Monte Carlo; Molten materials; Monte Carlo methods; Graphite; Nuclear reactors; Fuels; Heat Transfer; Moderators; molten salt; Fused salts; Molten salt reactor; MCNP; Internal circulations; Control rods; Internal circulation; Heat deposition},
	year = {2016},
	eissn = {1878-4224},
	pages = {155-163}
}

@article{MTMT:3120735,
	title = {Reactor Dosimetry Aspects of the Service Life Extension of the Hungarian Paks NPP},
	url = {https://m2.mtmt.hu/api/publication/3120735},
	isbn = {9782759819294},
	author = {Zsolnay, Éva and Czifrus, Szabolcs and Fehér, Sándor and Hordósy, Gábor and Keresztúri, András and Kresz, N and Oszvald, F},
	doi = {10.1051/epjconf/201610602010},
	journal-iso = {EPJ WEB CONF},
	journal = {EPJ WEB OF CONFERENCES},
	volume = {106},
	unique-id = {3120735},
	issn = {2101-6275},
	year = {2016},
	eissn = {2100-014X}
}

@CONFERENCE{MTMT:2949545,
	title = {Development and application of SITON, a new fuel cycle simulation code},
	url = {https://m2.mtmt.hu/api/publication/2949545},
	author = {Áron, Brolly and Szieberth, Máté and Halász, Máté Gergely and Nagy, Lajos and Fehér, Sándor},
	booktitle = {Actinide and fission product partitioning and transmutation},
	unique-id = {2949545},
	year = {2015},
	pages = {404-415},
	orcid-numbers = {Szieberth, Máté/0000-0001-9192-0352}
}

@article{MTMT:2969965,
	title = {Methodology and conclusions of activation calculations of WWER-440 type nuclear power plants},
	url = {https://m2.mtmt.hu/api/publication/2969965},
	author = {Babcsány, Boglárka and Czifrus, Szabolcs and Fehér, Sándor},
	doi = {10.1016/j.nucengdes.2014.11.032},
	journal-iso = {NUCL ENG DES},
	journal = {NUCLEAR ENGINEERING AND DESIGN},
	volume = {284},
	unique-id = {2969965},
	issn = {0029-5493},
	year = {2015},
	eissn = {1872-759X},
	pages = {228-237}
}

@inproceedings{MTMT:2917806,
	title = {FUEL CYCLE STUDIES ON MINOR ACTINIDE BURNING IN GAS COOLED FAST REACTORS},
	url = {https://m2.mtmt.hu/api/publication/2917806},
	author = {Szieberth, Máté and Halász, Máté Gergely and Reiss, Tibor and Fehér, Sándor},
	booktitle = {Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13)},
	unique-id = {2917806},
	year = {2015},
	orcid-numbers = {Szieberth, Máté/0000-0001-9192-0352}
}