TY  - JOUR
AU  - Butenko, A V
AU  - Syresin, E M
AU  - Tyutyunnikov, S I
AU  - Batyaev, V F
AU  - Kulevoy, T V
AU  - Pavlov, K V
AU  - Rogov, V I
AU  - Titarenko, A. Yu.
AU  - Titarenko, Yu. E
AU  - Berlyand, A. V.
AU  - Berlyand, V A
AU  - Sobolevskiy, N M
AU  - Bobrovskiy, D V
AU  - Chumakov, A I
AU  - Saburov, V O
AU  - Soloviev, A N
AU  - Pešić, Milan P.
PY  - 2019
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/8787
AB  - Abstract: The problems of the metrological certification of beams of high-energy heavy charged particles (HCPs) and protons that will be used in the study—as well as testing for radiation resistance—of promising products of semiconductor micro- and nanoelectronics, solid-state microwave electronics, and micromechanical systems are considered. One of the main requirements for such beams is ensuring the desired range of linear energy transfer (LET). Two methods for changing the LET are considered, one of which is based on using the ions of various types (16O, 22Ne, 40Ar, 56Fe, 84Kr, 136Xe, 209Bi), and the other is based on using ion of the same type (197Au), but with different energies. The advantages of using both methods are considered and the problems arising when using the second method are analyzed. © 2019, Pleiades Publishing, Ltd.
T2  - Physics of Particles and Nuclei Letters
T1  - Analysis of Metrological Provision Problems of a Test Stand for Testing Radio-Electronic Products for Resistance to Irradiation with High-Energy Heavy Ions
VL  - 16
IS  - 6
SP  - 734
EP  - 743
DO  - 10.1134/S1547477119060098
ER  - 

@article{
author = "Butenko, A V and Syresin, E M and Tyutyunnikov, S I and Batyaev, V F and Kulevoy, T V and Pavlov, K V and Rogov, V I and Titarenko, A. Yu. and Titarenko, Yu. E and Berlyand, A. V. and Berlyand, V A and Sobolevskiy, N M and Bobrovskiy, D V and Chumakov, A I and Saburov, V O and Soloviev, A N and Pešić, Milan P.",
year = "2019",
abstract = "Abstract: The problems of the metrological certification of beams of high-energy heavy charged particles (HCPs) and protons that will be used in the study—as well as testing for radiation resistance—of promising products of semiconductor micro- and nanoelectronics, solid-state microwave electronics, and micromechanical systems are considered. One of the main requirements for such beams is ensuring the desired range of linear energy transfer (LET). Two methods for changing the LET are considered, one of which is based on using the ions of various types (16O, 22Ne, 40Ar, 56Fe, 84Kr, 136Xe, 209Bi), and the other is based on using ion of the same type (197Au), but with different energies. The advantages of using both methods are considered and the problems arising when using the second method are analyzed. © 2019, Pleiades Publishing, Ltd.",
journal = "Physics of Particles and Nuclei Letters",
title = "Analysis of Metrological Provision Problems of a Test Stand for Testing Radio-Electronic Products for Resistance to Irradiation with High-Energy Heavy Ions",
volume = "16",
number = "6",
pages = "734-743",
doi = "10.1134/S1547477119060098"
}

Butenko, A. V., Syresin, E. M., Tyutyunnikov, S. I., Batyaev, V. F., Kulevoy, T. V., Pavlov, K. V., Rogov, V. I., Titarenko, A. Yu., Titarenko, Yu. E., Berlyand, A. V., Berlyand, V. A., Sobolevskiy, N. M., Bobrovskiy, D. V., Chumakov, A. I., Saburov, V. O., Soloviev, A. N.,& Pešić, M. P.. (2019). Analysis of Metrological Provision Problems of a Test Stand for Testing Radio-Electronic Products for Resistance to Irradiation with High-Energy Heavy Ions. in Physics of Particles and Nuclei Letters, 16(6), 734-743.
https://doi.org/10.1134/S1547477119060098

Butenko AV, Syresin EM, Tyutyunnikov SI, Batyaev VF, Kulevoy TV, Pavlov KV, Rogov VI, Titarenko AY, Titarenko YE, Berlyand AV, Berlyand VA, Sobolevskiy NM, Bobrovskiy DV, Chumakov AI, Saburov VO, Soloviev AN, Pešić MP. Analysis of Metrological Provision Problems of a Test Stand for Testing Radio-Electronic Products for Resistance to Irradiation with High-Energy Heavy Ions. in Physics of Particles and Nuclei Letters. 2019;16(6):734-743.
doi:10.1134/S1547477119060098 .

Butenko, A V, Syresin, E M, Tyutyunnikov, S I, Batyaev, V F, Kulevoy, T V, Pavlov, K V, Rogov, V I, Titarenko, A. Yu., Titarenko, Yu. E, Berlyand, A. V., Berlyand, V A, Sobolevskiy, N M, Bobrovskiy, D V, Chumakov, A I, Saburov, V O, Soloviev, A N, Pešić, Milan P., "Analysis of Metrological Provision Problems of a Test Stand for Testing Radio-Electronic Products for Resistance to Irradiation with High-Energy Heavy Ions" in Physics of Particles and Nuclei Letters, 16, no. 6 (2019):734-743,
https://doi.org/10.1134/S1547477119060098 . .

TY  - JOUR
AU  - Pešić, Milan P.
PY  - 2018
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/7947
T2  - Nuclear Technology and Radiation Protection
T1  - Erratum: A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”
VL  - 33
IS  - 2
SP  - 230
UR  - https://hdl.handle.net/21.15107/rcub_vinar_7947
ER  - 

@article{
author = "Pešić, Milan P.",
year = "2018",
journal = "Nuclear Technology and Radiation Protection",
title = "Erratum: A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”",
volume = "33",
number = "2",
pages = "230",
url = "https://hdl.handle.net/21.15107/rcub_vinar_7947"
}

Pešić, M. P.. (2018). Erratum: A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”. in Nuclear Technology and Radiation Protection, 33(2), 230.
https://hdl.handle.net/21.15107/rcub_vinar_7947

Pešić MP. Erratum: A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”. in Nuclear Technology and Radiation Protection. 2018;33(2):230.
https://hdl.handle.net/21.15107/rcub_vinar_7947 .

Pešić, Milan P., "Erratum: A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”" in Nuclear Technology and Radiation Protection, 33, no. 2 (2018):230,
https://hdl.handle.net/21.15107/rcub_vinar_7947 .

TY  - JOUR
AU  - Pešić, Milan P.
PY  - 2018
UR  - http://www.doiserbia.nb.rs/Article.aspx?ID=1451-39941801001P
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/7787
AB  - German Nazi state conducted researches in nuclear technologies as an attempt to achieve various military goals. As the result of these researches, German scientists developed different, advanced nuclear technologies in years before and during World War II. In an attempt to develop the "Uranmaschinen", in which controlled release of high energy in fission process can be achieved, various approaches were examined, theoretically and experimentally. These studies were conducted under support of the German Nazi state and were known as the First and Second "Uranverain" (Uranium Society/Club). Versions of the "Uranmaschinen" were based, mainly, on natural uranium fuel and moderators of heavy water, regular water or paraffin. The latest known fission device was the subcritical nuclear fission reactor B-VIII, re-built in village Haigerloch, Bavaria, Southern Germany, in first months of 1945. It was a tank type device with natural uranium metal fuel and heavy water moderator, reflected by graphite. Radiation shielding of the device was achieved, primarily, by surrounding the reactor tank by regular water. The whole device construction was assembled inside a concrete hole in the floor of an underground cave, ex beer cellar. A recent neutronics study of this reactor was done, assuming fuel rods with lumped parameters approximation, by Italian Bologna University LIN (Laboratorio Ingegneria Nucleare) research group in 2009. This paper is a new approach to the neutronics study of the B-VIII reactor with an attempt to model real fuel-moderator geometry. This study points out many approximations and simplifications, made during the B-VIII material composition and geometry modeling, due to missing data. The paper investigates the influence to criticality of numerous uncertainties in the material compositions, mass densities and geometry of the facility. The Monte Carlo MCNP6.1 code with the latest ACE type neutron nuclear cross section data is used for that purpose. Additionally, an attempt of estimation of the uncertainty of the experimental result of the neutron multiplication was given. Differences in the calculated values of the neutron multiplication and the experimental one are investigated and tried to explain. These analyses show that the B-VIII was a subcritical device, as it was shown by the experimental results of the German scientists achieved in March-April 1945 in Haigerloch.
T2  - Nuclear Technology and Radiation Protection
T1  - A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”
T1  - Један нови приступ моделовању "B-VIII" система, крајњем достигнућу другог Уранијумског друштва
VL  - 33
IS  - 1
SP  - 1
EP  - 23
DO  - 10.2298/NTRP1801001P
ER  - 

@article{
author = "Pešić, Milan P.",
year = "2018",
abstract = "German Nazi state conducted researches in nuclear technologies as an attempt to achieve various military goals. As the result of these researches, German scientists developed different, advanced nuclear technologies in years before and during World War II. In an attempt to develop the "Uranmaschinen", in which controlled release of high energy in fission process can be achieved, various approaches were examined, theoretically and experimentally. These studies were conducted under support of the German Nazi state and were known as the First and Second "Uranverain" (Uranium Society/Club). Versions of the "Uranmaschinen" were based, mainly, on natural uranium fuel and moderators of heavy water, regular water or paraffin. The latest known fission device was the subcritical nuclear fission reactor B-VIII, re-built in village Haigerloch, Bavaria, Southern Germany, in first months of 1945. It was a tank type device with natural uranium metal fuel and heavy water moderator, reflected by graphite. Radiation shielding of the device was achieved, primarily, by surrounding the reactor tank by regular water. The whole device construction was assembled inside a concrete hole in the floor of an underground cave, ex beer cellar. A recent neutronics study of this reactor was done, assuming fuel rods with lumped parameters approximation, by Italian Bologna University LIN (Laboratorio Ingegneria Nucleare) research group in 2009. This paper is a new approach to the neutronics study of the B-VIII reactor with an attempt to model real fuel-moderator geometry. This study points out many approximations and simplifications, made during the B-VIII material composition and geometry modeling, due to missing data. The paper investigates the influence to criticality of numerous uncertainties in the material compositions, mass densities and geometry of the facility. The Monte Carlo MCNP6.1 code with the latest ACE type neutron nuclear cross section data is used for that purpose. Additionally, an attempt of estimation of the uncertainty of the experimental result of the neutron multiplication was given. Differences in the calculated values of the neutron multiplication and the experimental one are investigated and tried to explain. These analyses show that the B-VIII was a subcritical device, as it was shown by the experimental results of the German scientists achieved in March-April 1945 in Haigerloch.",
journal = "Nuclear Technology and Radiation Protection",
title = "A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”, Један нови приступ моделовању "B-VIII" система, крајњем достигнућу другог Уранијумског друштва",
volume = "33",
number = "1",
pages = "1-23",
doi = "10.2298/NTRP1801001P"
}

Pešić, M. P.. (2018). A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”. in Nuclear Technology and Radiation Protection, 33(1), 1-23.
https://doi.org/10.2298/NTRP1801001P

Pešić MP. A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”. in Nuclear Technology and Radiation Protection. 2018;33(1):1-23.
doi:10.2298/NTRP1801001P .

Pešić, Milan P., "A new approach on modeling of the B-VIII, the ultimate achievement of the second “Uranverain”" in Nuclear Technology and Radiation Protection, 33, no. 1 (2018):1-23,
https://doi.org/10.2298/NTRP1801001P . .

TY  - JOUR
AU  - Pešić, Milan P.
PY  - 2016
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/1063
AB  - Demand on the availability of well-defined reactor experiments for validation of computer codes for use in nuclear industry and nuclear technology is everlasting. Users must be confident of the results obtained by the proven computer codes and nuclear data libraries chosen in the models. The well-defined (mostly historical) and evaluated reactor experiments (about 5000 in 2015) were collected continuously as the benchmarks within the frame of the OECD/NEA international projects ICSBEP (since 1995) and IRPhEP (since 2003). The Handbooks of the Projects are published in electronic forms (at the NEA web site of the OECD and at a DVD media) every year. This study is aimed to (a) examine and evaluate reactor basic experiments, carried out in the lattice of the natural uranium metal fuel in the heavy water of the RB critical assembly first core in 1958, and (b) demonstrate their possibility for validation of modern nuclear data libraries. These RB reactor basic experiments include: (1) approach to criticality, (2) determination of the reactivity gradient at the D2O critical level, (3) measurement of the dependence of the D2O critical level on the D2O temperature, i.e. dependence of the reactivity with change in the D2O temperature; (4) the critical reactor geometrical parameter (buckling) measurements, (5) the migration length measurements, (6) determination of the neutron multiplication factor in the infinite lattice, and (7) the safety rods reactivity measurements. Results of the experiments are compared to the results obtained using modern nuclear data libraries of the ACE type by applying the MCNP6.1, a well-known and proven computer code based on the Monte Carlo method. A short overview of these experiments (done at the RB assembly) is shown. A brief description of the neutron ACE type nuclear data libraries (created in the LANL, based on the ENDF/B-VII.0 and ENDF/B-VII.1 files, or created in the OECD/NEA, based on the JEFF-3.2 evaluated nuclear data files), used in this validation study, is given. The benchmark models used for this validation study are described and the obtained results were analyzed. It is concluded that most of these reactor basic experiments, carried out in the lattice of the natural uranium metal fuel rods and the heavy water of the RB critical assembly, can be used as the benchmarks for validation of new nuclear data libraries. It may be done after further evaluations of influence of missing data, information and uncertainties in the material composition and geometry dimensions have been prepared according to the IRPhEP criteria and standards.
T2  - Nuclear technology and radiation protection
T1  - A Study on the Use of the Reactor Basic Experiments in the U-D2o Lattices of the Rb Critical Assembly for Validation of Modern Nuclear Data Libraries
VL  - 31
IS  - 1
SP  - 1
EP  - 15
DO  - 10.2298/NTRP1601001P
ER  - 

@article{
author = "Pešić, Milan P.",
year = "2016",
abstract = "Demand on the availability of well-defined reactor experiments for validation of computer codes for use in nuclear industry and nuclear technology is everlasting. Users must be confident of the results obtained by the proven computer codes and nuclear data libraries chosen in the models. The well-defined (mostly historical) and evaluated reactor experiments (about 5000 in 2015) were collected continuously as the benchmarks within the frame of the OECD/NEA international projects ICSBEP (since 1995) and IRPhEP (since 2003). The Handbooks of the Projects are published in electronic forms (at the NEA web site of the OECD and at a DVD media) every year. This study is aimed to (a) examine and evaluate reactor basic experiments, carried out in the lattice of the natural uranium metal fuel in the heavy water of the RB critical assembly first core in 1958, and (b) demonstrate their possibility for validation of modern nuclear data libraries. These RB reactor basic experiments include: (1) approach to criticality, (2) determination of the reactivity gradient at the D2O critical level, (3) measurement of the dependence of the D2O critical level on the D2O temperature, i.e. dependence of the reactivity with change in the D2O temperature; (4) the critical reactor geometrical parameter (buckling) measurements, (5) the migration length measurements, (6) determination of the neutron multiplication factor in the infinite lattice, and (7) the safety rods reactivity measurements. Results of the experiments are compared to the results obtained using modern nuclear data libraries of the ACE type by applying the MCNP6.1, a well-known and proven computer code based on the Monte Carlo method. A short overview of these experiments (done at the RB assembly) is shown. A brief description of the neutron ACE type nuclear data libraries (created in the LANL, based on the ENDF/B-VII.0 and ENDF/B-VII.1 files, or created in the OECD/NEA, based on the JEFF-3.2 evaluated nuclear data files), used in this validation study, is given. The benchmark models used for this validation study are described and the obtained results were analyzed. It is concluded that most of these reactor basic experiments, carried out in the lattice of the natural uranium metal fuel rods and the heavy water of the RB critical assembly, can be used as the benchmarks for validation of new nuclear data libraries. It may be done after further evaluations of influence of missing data, information and uncertainties in the material composition and geometry dimensions have been prepared according to the IRPhEP criteria and standards.",
journal = "Nuclear technology and radiation protection",
title = "A Study on the Use of the Reactor Basic Experiments in the U-D2o Lattices of the Rb Critical Assembly for Validation of Modern Nuclear Data Libraries",
volume = "31",
number = "1",
pages = "1-15",
doi = "10.2298/NTRP1601001P"
}

Pešić, M. P.. (2016). A Study on the Use of the Reactor Basic Experiments in the U-D2o Lattices of the Rb Critical Assembly for Validation of Modern Nuclear Data Libraries. in Nuclear technology and radiation protection, 31(1), 1-15.
https://doi.org/10.2298/NTRP1601001P

Pešić MP. A Study on the Use of the Reactor Basic Experiments in the U-D2o Lattices of the Rb Critical Assembly for Validation of Modern Nuclear Data Libraries. in Nuclear technology and radiation protection. 2016;31(1):1-15.
doi:10.2298/NTRP1601001P .

Pešić, Milan P., "A Study on the Use of the Reactor Basic Experiments in the U-D2o Lattices of the Rb Critical Assembly for Validation of Modern Nuclear Data Libraries" in Nuclear technology and radiation protection, 31, no. 1 (2016):1-15,
https://doi.org/10.2298/NTRP1601001P . .

TY  - JOUR
AU  - Pešić, Milan P.
AU  - Titarenko, Yury E.
AU  - Batyaev, Viacheslav F.
AU  - Pavlov, Kirill V.
AU  - Tffarenko, Alexey Yu.
AU  - Zhivun, Valeriy M.
AU  - Igumnov, Mikhail M.
AU  - Konev, Viacheslav N.
AU  - Legostaev, Vladimir O.
PY  - 2015
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/529
AB  - Verification of neutron fission cross-sections of minor actinides from some recently available evaluated nuclear data libraries was carried out by comparison of the reaction rates calculated by the MCNP6.1 computer code to the experimental values. The experimental samples, containing thin layers of U-235, Np-237, (238,239,240,241)pu, (242)mAin, Cm-243, (245)cm, and Cm-247, deposited on metal support and foils of U-235 (pseudo-alloy Al-27 + U-235), U-238, In-nat, Zn-64, Al-27, and multi-component sample alloy Al-27 + (55)mn Cu-nat + Lu-nat Au-197, were irradiated in the channels of the tank containing fluorine salts 0.52NaF + 0.48ZrF(4), labelled as the Micromodel Salt Blanket, inserted in the lattice centre of the MAKET heavy water critical assembly at the Institute for Theoretical and Experimental Physics, Moscow. This paper is a continuation of earlier initiated scientific-research activities carried out for validation of the evaluated fission cross-sections of actinides that were supposed to be used for the quality examination of the fuel design of the accelerator driven systems or fast reactors, and consequently, determination of transmutation rates of actinides, and therefore, determination of operation parameters of these reactor facilities. These scientific-research activities were carried out within a frame of scientific projects supported by the International Science and Technology Center and the International Atomic Energy Agency co-ordinated research activities, from 1999 to 2010. Obtained results confirm that further research is needed in evaluations in order to establish better neutron cross-section data for the minor actinides and selected nuclides which could be used in the accelerator driven systems or fast reactors.
T2  - Nuclear technology and radiation protection
T1  - Validation of Minor Actinides Fission Neutron Cross-Sections
VL  - 30
IS  - 1
SP  - 1
EP  - 10
DO  - 10.2298/NTRP1501001P
ER  - 

@article{
author = "Pešić, Milan P. and Titarenko, Yury E. and Batyaev, Viacheslav F. and Pavlov, Kirill V. and Tffarenko, Alexey Yu. and Zhivun, Valeriy M. and Igumnov, Mikhail M. and Konev, Viacheslav N. and Legostaev, Vladimir O.",
year = "2015",
abstract = "Verification of neutron fission cross-sections of minor actinides from some recently available evaluated nuclear data libraries was carried out by comparison of the reaction rates calculated by the MCNP6.1 computer code to the experimental values. The experimental samples, containing thin layers of U-235, Np-237, (238,239,240,241)pu, (242)mAin, Cm-243, (245)cm, and Cm-247, deposited on metal support and foils of U-235 (pseudo-alloy Al-27 + U-235), U-238, In-nat, Zn-64, Al-27, and multi-component sample alloy Al-27 + (55)mn Cu-nat + Lu-nat Au-197, were irradiated in the channels of the tank containing fluorine salts 0.52NaF + 0.48ZrF(4), labelled as the Micromodel Salt Blanket, inserted in the lattice centre of the MAKET heavy water critical assembly at the Institute for Theoretical and Experimental Physics, Moscow. This paper is a continuation of earlier initiated scientific-research activities carried out for validation of the evaluated fission cross-sections of actinides that were supposed to be used for the quality examination of the fuel design of the accelerator driven systems or fast reactors, and consequently, determination of transmutation rates of actinides, and therefore, determination of operation parameters of these reactor facilities. These scientific-research activities were carried out within a frame of scientific projects supported by the International Science and Technology Center and the International Atomic Energy Agency co-ordinated research activities, from 1999 to 2010. Obtained results confirm that further research is needed in evaluations in order to establish better neutron cross-section data for the minor actinides and selected nuclides which could be used in the accelerator driven systems or fast reactors.",
journal = "Nuclear technology and radiation protection",
title = "Validation of Minor Actinides Fission Neutron Cross-Sections",
volume = "30",
number = "1",
pages = "1-10",
doi = "10.2298/NTRP1501001P"
}

Pešić, M. P., Titarenko, Y. E., Batyaev, V. F., Pavlov, K. V., Tffarenko, A. Yu., Zhivun, V. M., Igumnov, M. M., Konev, V. N.,& Legostaev, V. O.. (2015). Validation of Minor Actinides Fission Neutron Cross-Sections. in Nuclear technology and radiation protection, 30(1), 1-10.
https://doi.org/10.2298/NTRP1501001P

Pešić MP, Titarenko YE, Batyaev VF, Pavlov KV, Tffarenko AY, Zhivun VM, Igumnov MM, Konev VN, Legostaev VO. Validation of Minor Actinides Fission Neutron Cross-Sections. in Nuclear technology and radiation protection. 2015;30(1):1-10.
doi:10.2298/NTRP1501001P .

Pešić, Milan P., Titarenko, Yury E., Batyaev, Viacheslav F., Pavlov, Kirill V., Tffarenko, Alexey Yu., Zhivun, Valeriy M., Igumnov, Mikhail M., Konev, Viacheslav N., Legostaev, Vladimir O., "Validation of Minor Actinides Fission Neutron Cross-Sections" in Nuclear technology and radiation protection, 30, no. 1 (2015):1-10,
https://doi.org/10.2298/NTRP1501001P . .

TY  - JOUR
AU  - Pešić, Milan P.
PY  - 2012
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/5067
AB  - A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option) and mode n p e, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally) in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally) in the phantoms tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and, for the first time, a Cause and Effect diagram has been created in Cause Mapping methodology and shown in Appendix A.2.
T2  - Nuclear technology and radiation protection
T1  - Estimation of Doses Received By Operators in the 1958 Rb Reactor Accident Using the Mcnp5 Computer Code Simulation
VL  - 27
IS  - 3
SP  - 199
EP  - 221
DO  - 10.2298/NTRP1203199P
ER  - 

@article{
author = "Pešić, Milan P.",
year = "2012",
abstract = "A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option) and mode n p e, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally) in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally) in the phantoms tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and, for the first time, a Cause and Effect diagram has been created in Cause Mapping methodology and shown in Appendix A.2.",
journal = "Nuclear technology and radiation protection",
title = "Estimation of Doses Received By Operators in the 1958 Rb Reactor Accident Using the Mcnp5 Computer Code Simulation",
volume = "27",
number = "3",
pages = "199-221",
doi = "10.2298/NTRP1203199P"
}

Pešić, M. P.. (2012). Estimation of Doses Received By Operators in the 1958 Rb Reactor Accident Using the Mcnp5 Computer Code Simulation. in Nuclear technology and radiation protection, 27(3), 199-221.
https://doi.org/10.2298/NTRP1203199P

Pešić MP. Estimation of Doses Received By Operators in the 1958 Rb Reactor Accident Using the Mcnp5 Computer Code Simulation. in Nuclear technology and radiation protection. 2012;27(3):199-221.
doi:10.2298/NTRP1203199P .

Pešić, Milan P., "Estimation of Doses Received By Operators in the 1958 Rb Reactor Accident Using the Mcnp5 Computer Code Simulation" in Nuclear technology and radiation protection, 27, no. 3 (2012):199-221,
https://doi.org/10.2298/NTRP1203199P . .

TY  - JOUR
AU  - Pešić, Milan P.
PY  - 2012
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/4353
AB  - In 1958, the experimental RB reactor was designed as a heavy water critical assembly with natural uranium metal rods. It was the first nuclear fission critical facility at the Boris Kidric (now Vinca) Institute of Nuclear Sciences in the former Yugoslavia. The first non-reflected, unshielded core was assembled in an aluminium tank, at a distance of around 4 m from all adjacent surfaces, so as to achieve as low as possible neutron back reflection to the core. The 2% enriched uranium metal and 80% enriched uranium dioxide (dispersed in aluminum) fuel elements (known as slugs) were obtained from the USSR in 1960 and 1976, respectively. The so-called clean cores of the RB reactor were assembled from a single type of fuel elements. The mixed cores of the RB reactor, assembled from two or three types of different fuel elements, were also positioned in heavy water. Both types of cores can be composed as square lattices with different pitches, covering a range of 7 cm to 24 cm. A radial heavy water reflector of various thicknesses usually surrounds the cores. Up to 2006, four sets of clean cores (44 core configurations) have been accepted as criticality benchmarks and included into the OECD ICSBEP Handbook. The RB mixed core 39/1978 was made of 31 natural uranium metal rods positioned in heavy water, in a lattice with a pitch of 8 root 2 cm and 78*9 low enriched uranium metal slugs placed in heavy water in lattice pitches of 8 cm and 8 root 2 cm. This RB multipart core has now been proposed to the ICSBEP as a possible new U-D2O criticality benchmark, due to its complex irregular lattice which is almost impossible to simulate by computer codes that require axial symmetry or a regular lattice. Criticality results presented in this paper were obtained in calculations carried out by using recent versions of the MCNP5 and KENO V.a computer codes and various libraries of the neutron cross-sections data. Our results confirm that the proposed RB reactor complex core, RB 39/1978, may well prove to be a new U-D2O benchmark criticality system for validating reactor design computer codes and data libraries recommended by the ICSPEP.
T2  - Nuclear technology and radiation protection
T1  - A Proposal for a New U-D2o Criticality Benchmark: Rb Reactor Core 39/1978
VL  - 27
IS  - 1
SP  - 75
EP  - 83
DO  - 10.2298/NTRP1201075P
ER  - 

@article{
author = "Pešić, Milan P.",
year = "2012",
abstract = "In 1958, the experimental RB reactor was designed as a heavy water critical assembly with natural uranium metal rods. It was the first nuclear fission critical facility at the Boris Kidric (now Vinca) Institute of Nuclear Sciences in the former Yugoslavia. The first non-reflected, unshielded core was assembled in an aluminium tank, at a distance of around 4 m from all adjacent surfaces, so as to achieve as low as possible neutron back reflection to the core. The 2% enriched uranium metal and 80% enriched uranium dioxide (dispersed in aluminum) fuel elements (known as slugs) were obtained from the USSR in 1960 and 1976, respectively. The so-called clean cores of the RB reactor were assembled from a single type of fuel elements. The mixed cores of the RB reactor, assembled from two or three types of different fuel elements, were also positioned in heavy water. Both types of cores can be composed as square lattices with different pitches, covering a range of 7 cm to 24 cm. A radial heavy water reflector of various thicknesses usually surrounds the cores. Up to 2006, four sets of clean cores (44 core configurations) have been accepted as criticality benchmarks and included into the OECD ICSBEP Handbook. The RB mixed core 39/1978 was made of 31 natural uranium metal rods positioned in heavy water, in a lattice with a pitch of 8 root 2 cm and 78*9 low enriched uranium metal slugs placed in heavy water in lattice pitches of 8 cm and 8 root 2 cm. This RB multipart core has now been proposed to the ICSBEP as a possible new U-D2O criticality benchmark, due to its complex irregular lattice which is almost impossible to simulate by computer codes that require axial symmetry or a regular lattice. Criticality results presented in this paper were obtained in calculations carried out by using recent versions of the MCNP5 and KENO V.a computer codes and various libraries of the neutron cross-sections data. Our results confirm that the proposed RB reactor complex core, RB 39/1978, may well prove to be a new U-D2O benchmark criticality system for validating reactor design computer codes and data libraries recommended by the ICSPEP.",
journal = "Nuclear technology and radiation protection",
title = "A Proposal for a New U-D2o Criticality Benchmark: Rb Reactor Core 39/1978",
volume = "27",
number = "1",
pages = "75-83",
doi = "10.2298/NTRP1201075P"
}

Pešić, M. P.. (2012). A Proposal for a New U-D2o Criticality Benchmark: Rb Reactor Core 39/1978. in Nuclear technology and radiation protection, 27(1), 75-83.
https://doi.org/10.2298/NTRP1201075P

Pešić MP. A Proposal for a New U-D2o Criticality Benchmark: Rb Reactor Core 39/1978. in Nuclear technology and radiation protection. 2012;27(1):75-83.
doi:10.2298/NTRP1201075P .

Pešić, Milan P., "A Proposal for a New U-D2o Criticality Benchmark: Rb Reactor Core 39/1978" in Nuclear technology and radiation protection, 27, no. 1 (2012):75-83,
https://doi.org/10.2298/NTRP1201075P . .

TY  - JOUR
AU  - Pešić, Milan P.
PY  - 2011
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/4505
AB  - New validation of the well-known Monte Carlo code MCNP5 against measured criticality and kinetics data for the coupled fast-thermal HERBE System at the Reactor B critical assembly is shown in this paper. Results of earlier calculations of these criticality and kinetics parameters, done by combination of transport and diffusion codes using two-dimension geometry model are compared to results of new calculations carried out by the MCNP5 code in three-dimension geometry. Satisfactory agreements in comparison of new results with experimental data, in spite complex heterogeneous composition of the HERBE core, are achieved confirming that MCNP5 code could apply successfully to study on HERBE kinetics parameters after uncertainties in impurities in material compositions and positions of fuel elements in fast zone were removed.
T2  - Nuclear technology and radiation protection
T1  - Mcnp5 Study on Kinetics Parameters of Coupled Fast-Thermal System Herbe
VL  - 26
IS  - 2
SP  - 134
EP  - 139
DO  - 10.2298/NTRP1102134P
ER  - 

@article{
author = "Pešić, Milan P.",
year = "2011",
abstract = "New validation of the well-known Monte Carlo code MCNP5 against measured criticality and kinetics data for the coupled fast-thermal HERBE System at the Reactor B critical assembly is shown in this paper. Results of earlier calculations of these criticality and kinetics parameters, done by combination of transport and diffusion codes using two-dimension geometry model are compared to results of new calculations carried out by the MCNP5 code in three-dimension geometry. Satisfactory agreements in comparison of new results with experimental data, in spite complex heterogeneous composition of the HERBE core, are achieved confirming that MCNP5 code could apply successfully to study on HERBE kinetics parameters after uncertainties in impurities in material compositions and positions of fuel elements in fast zone were removed.",
journal = "Nuclear technology and radiation protection",
title = "Mcnp5 Study on Kinetics Parameters of Coupled Fast-Thermal System Herbe",
volume = "26",
number = "2",
pages = "134-139",
doi = "10.2298/NTRP1102134P"
}

Pešić, M. P.. (2011). Mcnp5 Study on Kinetics Parameters of Coupled Fast-Thermal System Herbe. in Nuclear technology and radiation protection, 26(2), 134-139.
https://doi.org/10.2298/NTRP1102134P

Pešić MP. Mcnp5 Study on Kinetics Parameters of Coupled Fast-Thermal System Herbe. in Nuclear technology and radiation protection. 2011;26(2):134-139.
doi:10.2298/NTRP1102134P .

Pešić, Milan P., "Mcnp5 Study on Kinetics Parameters of Coupled Fast-Thermal System Herbe" in Nuclear technology and radiation protection, 26, no. 2 (2011):134-139,
https://doi.org/10.2298/NTRP1102134P . .

TY  - CONF
AU  - Beličev, Petar
AU  - Spasić-Jokić, Vesna
AU  - Mayer, S.
AU  - Milosevic, M.
AU  - Ilic, R.
AU  - Pešić, Milan P.
PY  - 2010
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/6899
AB  - The Metal -Oxide Semiconductor Field-Effect-Transistor (MOSFET, RadFET) is frequently used as a sensor of ionizing radiation in nuclear-medicine, diagnostic-radiology, radiotherapy quality-assurance and in the nuclear and space industries. We focused our investigations on calculating the energy response of a p-type RadFET to low-energy photons in range from 12 keV to 2 MeV and on understanding the influence of uncertainties in the composition and geometry of the device in calculating the energy response function. All results were normalized to unit air kerma incident on the RadFET for incident photon energy of 1.1 MeV. The calculations of the energy response characteristics of a RadFET radiation detector were performed via Monte Carlo simulations using the MCNPX code and for a limited number of incident photon energies the FOTELP code was also used for the sake of comparison. The geometry of the RadFET was modeled as a simple stack of appropriate materials. Our goal was to obtain results with statistical uncertainties better than 1% (fulfilled in MCNPX calculations for all incident energies which resulted in simulations with 1 - 2x10(9) histories.
C3  - Journal of Physics: Conference Series
T1  - Monte Carlo Calculation of the Energy Response Characteristics of a RadFET Radiation Detector
VL  - 238
DO  - 10.1088/1742-6596/238/1/012030
ER  - 

@conference{
author = "Beličev, Petar and Spasić-Jokić, Vesna and Mayer, S. and Milosevic, M. and Ilic, R. and Pešić, Milan P.",
year = "2010",
abstract = "The Metal -Oxide Semiconductor Field-Effect-Transistor (MOSFET, RadFET) is frequently used as a sensor of ionizing radiation in nuclear-medicine, diagnostic-radiology, radiotherapy quality-assurance and in the nuclear and space industries. We focused our investigations on calculating the energy response of a p-type RadFET to low-energy photons in range from 12 keV to 2 MeV and on understanding the influence of uncertainties in the composition and geometry of the device in calculating the energy response function. All results were normalized to unit air kerma incident on the RadFET for incident photon energy of 1.1 MeV. The calculations of the energy response characteristics of a RadFET radiation detector were performed via Monte Carlo simulations using the MCNPX code and for a limited number of incident photon energies the FOTELP code was also used for the sake of comparison. The geometry of the RadFET was modeled as a simple stack of appropriate materials. Our goal was to obtain results with statistical uncertainties better than 1% (fulfilled in MCNPX calculations for all incident energies which resulted in simulations with 1 - 2x10(9) histories.",
journal = "Journal of Physics: Conference Series",
title = "Monte Carlo Calculation of the Energy Response Characteristics of a RadFET Radiation Detector",
volume = "238",
doi = "10.1088/1742-6596/238/1/012030"
}

Beličev, P., Spasić-Jokić, V., Mayer, S., Milosevic, M., Ilic, R.,& Pešić, M. P.. (2010). Monte Carlo Calculation of the Energy Response Characteristics of a RadFET Radiation Detector. in Journal of Physics: Conference Series, 238.
https://doi.org/10.1088/1742-6596/238/1/012030

Beličev P, Spasić-Jokić V, Mayer S, Milosevic M, Ilic R, Pešić MP. Monte Carlo Calculation of the Energy Response Characteristics of a RadFET Radiation Detector. in Journal of Physics: Conference Series. 2010;238.
doi:10.1088/1742-6596/238/1/012030 .

Beličev, Petar, Spasić-Jokić, Vesna, Mayer, S., Milosevic, M., Ilic, R., Pešić, Milan P., "Monte Carlo Calculation of the Energy Response Characteristics of a RadFET Radiation Detector" in Journal of Physics: Conference Series, 238 (2010),
https://doi.org/10.1088/1742-6596/238/1/012030 . .

TY  - JOUR
AU  - Marinković, Predrag
AU  - Pešić, Milan P.
PY  - 2009
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/3736
AB  - Monte Carlo techniques were usually used to study the effect of scattered photons on a radiographic X-ray image, Such approach is accurate, but computer time consuming. On the other hand, the exposure buildup factors can be used as approximate and efficient assessment to account for the scattering of X-rays. This method uses the known radiography parameters to find the resulting detector exposure due to both scattered and un-collided photons. A model for radiography simulation, based on X-ray dose buildup factor, is proposed. This model includes non-uniform attenuation in voxelized object of imaging (patient body tissue). Composition of patient body is considered as a multi-layer structure. Various empirical formulas exist for multi-layer structure calculations and they all calculate multi-layer buildup factors by combining single-layer buildup factors. The proposed model is convenient in cases when more exact techniques (like Monte Carlo) are not economical. (C) 2009 The Japan Society of Applied Physics
T2  - Japanese Journal of Applied Physics
T1  - Radiography Simulation Based on Exposure Buildup Factors for Multilayer Structures
VL  - 48
IS  - 6
DO  - 10.1143/JJAP.48.066002
ER  - 

@article{
author = "Marinković, Predrag and Pešić, Milan P.",
year = "2009",
abstract = "Monte Carlo techniques were usually used to study the effect of scattered photons on a radiographic X-ray image, Such approach is accurate, but computer time consuming. On the other hand, the exposure buildup factors can be used as approximate and efficient assessment to account for the scattering of X-rays. This method uses the known radiography parameters to find the resulting detector exposure due to both scattered and un-collided photons. A model for radiography simulation, based on X-ray dose buildup factor, is proposed. This model includes non-uniform attenuation in voxelized object of imaging (patient body tissue). Composition of patient body is considered as a multi-layer structure. Various empirical formulas exist for multi-layer structure calculations and they all calculate multi-layer buildup factors by combining single-layer buildup factors. The proposed model is convenient in cases when more exact techniques (like Monte Carlo) are not economical. (C) 2009 The Japan Society of Applied Physics",
journal = "Japanese Journal of Applied Physics",
title = "Radiography Simulation Based on Exposure Buildup Factors for Multilayer Structures",
volume = "48",
number = "6",
doi = "10.1143/JJAP.48.066002"
}

Marinković, P.,& Pešić, M. P.. (2009). Radiography Simulation Based on Exposure Buildup Factors for Multilayer Structures. in Japanese Journal of Applied Physics, 48(6).
https://doi.org/10.1143/JJAP.48.066002

Marinković P, Pešić MP. Radiography Simulation Based on Exposure Buildup Factors for Multilayer Structures. in Japanese Journal of Applied Physics. 2009;48(6).
doi:10.1143/JJAP.48.066002 .

Marinković, Predrag, Pešić, Milan P., "Radiography Simulation Based on Exposure Buildup Factors for Multilayer Structures" in Japanese Journal of Applied Physics, 48, no. 6 (2009),
https://doi.org/10.1143/JJAP.48.066002 . .

TY  - JOUR
AU  - Pešić, Milan P.
AU  - Nešković, Nebojša B.
PY  - 2006
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/6589
AB  - Study of a small accelerator-driven subcritical research reactor in the Vinca Institute of Nuclear Sciences was initiated in 1999. The idea was to extract a beam of medium-energy protons or deuterons from the TESLA accelerator installation, and to transport and inject it into the reactor. The reactor core was to be composed of the highly enriched uranium fuel elements. The reactor was designated as ADSRR-H. Since the use of this type of fuel elements was not recommended any more, the study of a small accelerator-driven subcritical research reactor employing the low-enriched uranium fuel elements began in 2004. The reactor was designated as ADSRR-L. We compare here the results of the initial computer simulations of ADSRR-H and ADSRR-L. The results have confirmed that our concept could be the basis for designing and construction of a low neutron flux model of the proposed accelerator-driven subcritical power reactor to be moderated and cooled by lead. Our objective is to study the physics and technologies necessary to design and construct ADSRR-L. The reactor would be used for development of nuclear techniques and technologies, and for basic and applied research in neutron physics, metrology, radiation protection and radiobiology. (c) 2006 Elsevier B.V. All rights reserved.
T2  - Nuclear Instruments and Methods in Physics Research. Section A: Accelerators, Spectrometers, Detectors, and Associated Equipment
T1  - Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation
VL  - 562
IS  - 2
SP  - 642
EP  - 645
DO  - 10.1016/j.nima.2006.02.173
ER  - 

@article{
author = "Pešić, Milan P. and Nešković, Nebojša B.",
year = "2006",
abstract = "Study of a small accelerator-driven subcritical research reactor in the Vinca Institute of Nuclear Sciences was initiated in 1999. The idea was to extract a beam of medium-energy protons or deuterons from the TESLA accelerator installation, and to transport and inject it into the reactor. The reactor core was to be composed of the highly enriched uranium fuel elements. The reactor was designated as ADSRR-H. Since the use of this type of fuel elements was not recommended any more, the study of a small accelerator-driven subcritical research reactor employing the low-enriched uranium fuel elements began in 2004. The reactor was designated as ADSRR-L. We compare here the results of the initial computer simulations of ADSRR-H and ADSRR-L. The results have confirmed that our concept could be the basis for designing and construction of a low neutron flux model of the proposed accelerator-driven subcritical power reactor to be moderated and cooled by lead. Our objective is to study the physics and technologies necessary to design and construct ADSRR-L. The reactor would be used for development of nuclear techniques and technologies, and for basic and applied research in neutron physics, metrology, radiation protection and radiobiology. (c) 2006 Elsevier B.V. All rights reserved.",
journal = "Nuclear Instruments and Methods in Physics Research. Section A: Accelerators, Spectrometers, Detectors, and Associated Equipment",
title = "Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation",
volume = "562",
number = "2",
pages = "642-645",
doi = "10.1016/j.nima.2006.02.173"
}

Pešić, M. P.,& Nešković, N. B.. (2006). Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation. in Nuclear Instruments and Methods in Physics Research. Section A: Accelerators, Spectrometers, Detectors, and Associated Equipment, 562(2), 642-645.
https://doi.org/10.1016/j.nima.2006.02.173

Pešić MP, Nešković NB. Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation. in Nuclear Instruments and Methods in Physics Research. Section A: Accelerators, Spectrometers, Detectors, and Associated Equipment. 2006;562(2):642-645.
doi:10.1016/j.nima.2006.02.173 .

Pešić, Milan P., Nešković, Nebojša B., "Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation" in Nuclear Instruments and Methods in Physics Research. Section A: Accelerators, Spectrometers, Detectors, and Associated Equipment, 562, no. 2 (2006):642-645,
https://doi.org/10.1016/j.nima.2006.02.173 . .

TY  - JOUR
AU  - Pešić, Milan P.
PY  - 2003
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/2676
AB  - The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments, contributing to the preservation of a rather small number of heavy water benchmark critical experiments.
T2  - Nuclear Science and Engineering
T1  - Criticality safety of low-enriched uranium and high-enriched uranium fuel elements in heavy water lattices
VL  - 145
IS  - 2
SP  - 225
EP  - 233
DO  - 10.13182/NSE03-A2378
ER  - 

@article{
author = "Pešić, Milan P.",
year = "2003",
abstract = "The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments, contributing to the preservation of a rather small number of heavy water benchmark critical experiments.",
journal = "Nuclear Science and Engineering",
title = "Criticality safety of low-enriched uranium and high-enriched uranium fuel elements in heavy water lattices",
volume = "145",
number = "2",
pages = "225-233",
doi = "10.13182/NSE03-A2378"
}

Pešić, M. P.. (2003). Criticality safety of low-enriched uranium and high-enriched uranium fuel elements in heavy water lattices. in Nuclear Science and Engineering, 145(2), 225-233.
https://doi.org/10.13182/NSE03-A2378

Pešić MP. Criticality safety of low-enriched uranium and high-enriched uranium fuel elements in heavy water lattices. in Nuclear Science and Engineering. 2003;145(2):225-233.
doi:10.13182/NSE03-A2378 .

Pešić, Milan P., "Criticality safety of low-enriched uranium and high-enriched uranium fuel elements in heavy water lattices" in Nuclear Science and Engineering, 145, no. 2 (2003):225-233,
https://doi.org/10.13182/NSE03-A2378 . .

TY  - CONF
AU  - Stefanovic, D
AU  - Pešić, Milan P.
PY  - 2002
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/6471
AB  - Reactor physics deals with theoretical and applied investigations with the aim to use the obtained results in nuclear engineering for designing, building and safe operating of nuclear reactors. Among scientific and engineering disciplines, commonly named nuclear technologies, reactor physics is a discipline that directly stems from nuclear sciences and it has not had an analogue base in the standard technologies. The basic studies cover theoretical and experimental determination of static and dynamic parameters of fission systems, burn up of nuclear fuel, investigation of parameters of nuclear engineering materials, prediction of behaviour of fission systems during probable accidents and estimation of their consequences. The basic and roadmap of development in reactor physics and engineering in Serbia during the second half of the 20th century are given in this paper.
T1  - Physics of nuclear reactors
VL  - 104
SP  - 51
EP  - 67
UR  - https://hdl.handle.net/21.15107/rcub_vinar_6471
ER  - 

@conference{
author = "Stefanovic, D and Pešić, Milan P.",
year = "2002",
abstract = "Reactor physics deals with theoretical and applied investigations with the aim to use the obtained results in nuclear engineering for designing, building and safe operating of nuclear reactors. Among scientific and engineering disciplines, commonly named nuclear technologies, reactor physics is a discipline that directly stems from nuclear sciences and it has not had an analogue base in the standard technologies. The basic studies cover theoretical and experimental determination of static and dynamic parameters of fission systems, burn up of nuclear fuel, investigation of parameters of nuclear engineering materials, prediction of behaviour of fission systems during probable accidents and estimation of their consequences. The basic and roadmap of development in reactor physics and engineering in Serbia during the second half of the 20th century are given in this paper.",
title = "Physics of nuclear reactors",
volume = "104",
pages = "51-67",
url = "https://hdl.handle.net/21.15107/rcub_vinar_6471"
}

Stefanovic, D.,& Pešić, M. P.. (2002). Physics of nuclear reactors. , 104, 51-67.
https://hdl.handle.net/21.15107/rcub_vinar_6471

Stefanovic D, Pešić MP. Physics of nuclear reactors. 2002;104:51-67.
https://hdl.handle.net/21.15107/rcub_vinar_6471 .

Stefanovic, D, Pešić, Milan P., "Physics of nuclear reactors", 104 (2002):51-67,
https://hdl.handle.net/21.15107/rcub_vinar_6471 .

TY  - JOUR
AU  - Pešić, Milan P.
AU  - Ninković, Marko M.
PY  - 1999
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/2266
AB  - The RE experimental reactor has operated at Vinca Institute of Nuclear Sciences since the end of April 1958. Tn this paper, neutron and gamma-ray spectra and corresponding dose quantities near the reactor, calculated by using the MCNP(TM) code, are compared to the measured values during the Third International Intercomparison Experiment on Nuclear Accident Dosimetry carried out at the RE reactor in 1973. Discrepancies in the correlation declared power of the reactor-dose rates are found. Good agreements are obtained between measured and calculated neutron and gamma-ray spectra, and corresponding absorbed doses in air, but only after the reactor declared power is multiplied by a correction factor, determined in this study.
T2  - Health Physics
T1  - Comparison of the MCNP (TM) calculated and measured radiation field quantities near the RB reactor
VL  - 77
IS  - 3
SP  - 276
EP  - 281
DO  - 10.1097/00004032-199909000-00005
ER  - 

@article{
author = "Pešić, Milan P. and Ninković, Marko M.",
year = "1999",
abstract = "The RE experimental reactor has operated at Vinca Institute of Nuclear Sciences since the end of April 1958. Tn this paper, neutron and gamma-ray spectra and corresponding dose quantities near the reactor, calculated by using the MCNP(TM) code, are compared to the measured values during the Third International Intercomparison Experiment on Nuclear Accident Dosimetry carried out at the RE reactor in 1973. Discrepancies in the correlation declared power of the reactor-dose rates are found. Good agreements are obtained between measured and calculated neutron and gamma-ray spectra, and corresponding absorbed doses in air, but only after the reactor declared power is multiplied by a correction factor, determined in this study.",
journal = "Health Physics",
title = "Comparison of the MCNP (TM) calculated and measured radiation field quantities near the RB reactor",
volume = "77",
number = "3",
pages = "276-281",
doi = "10.1097/00004032-199909000-00005"
}

Pešić, M. P.,& Ninković, M. M.. (1999). Comparison of the MCNP (TM) calculated and measured radiation field quantities near the RB reactor. in Health Physics, 77(3), 276-281.
https://doi.org/10.1097/00004032-199909000-00005

Pešić MP, Ninković MM. Comparison of the MCNP (TM) calculated and measured radiation field quantities near the RB reactor. in Health Physics. 1999;77(3):276-281.
doi:10.1097/00004032-199909000-00005 .

Pešić, Milan P., Ninković, Marko M., "Comparison of the MCNP (TM) calculated and measured radiation field quantities near the RB reactor" in Health Physics, 77, no. 3 (1999):276-281,
https://doi.org/10.1097/00004032-199909000-00005 . .

TY  - JOUR
AU  - Milošević, Miodrag J.
AU  - Pešić, Milan P.
PY  - 1998
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/2124
AB  - The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described.
T2  - Nuclear Engineering and Design
T1  - An analysis of burnup reactivity credit for reactor RA spent fuel storage
VL  - 180
IS  - 1
SP  - 29
EP  - 35
DO  - 10.1016/S0029-5493(97)00302-6
ER  - 

@article{
author = "Milošević, Miodrag J. and Pešić, Milan P.",
year = "1998",
abstract = "The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described.",
journal = "Nuclear Engineering and Design",
title = "An analysis of burnup reactivity credit for reactor RA spent fuel storage",
volume = "180",
number = "1",
pages = "29-35",
doi = "10.1016/S0029-5493(97)00302-6"
}

Milošević, M. J.,& Pešić, M. P.. (1998). An analysis of burnup reactivity credit for reactor RA spent fuel storage. in Nuclear Engineering and Design, 180(1), 29-35.
https://doi.org/10.1016/S0029-5493(97)00302-6

Milošević MJ, Pešić MP. An analysis of burnup reactivity credit for reactor RA spent fuel storage. in Nuclear Engineering and Design. 1998;180(1):29-35.
doi:10.1016/S0029-5493(97)00302-6 .

Milošević, Miodrag J., Pešić, Milan P., "An analysis of burnup reactivity credit for reactor RA spent fuel storage" in Nuclear Engineering and Design, 180, no. 1 (1998):29-35,
https://doi.org/10.1016/S0029-5493(97)00302-6 . .

TY  - JOUR
AU  - Pešić, Milan P.
AU  - Cupać, Svjetlana
PY  - 1996
UR  - https://vinar.vin.bg.ac.rs/handle/123456789/1955
T2  - Nuclear Engineering International
T1  - Working in a difficult regime
VL  - 41
IS  - 498
SP  - 40
EP  - 41
UR  - https://hdl.handle.net/21.15107/rcub_vinar_1955
ER  - 

@article{
author = "Pešić, Milan P. and Cupać, Svjetlana",
year = "1996",
journal = "Nuclear Engineering International",
title = "Working in a difficult regime",
volume = "41",
number = "498",
pages = "40-41",
url = "https://hdl.handle.net/21.15107/rcub_vinar_1955"
}

Pešić, M. P.,& Cupać, S.. (1996). Working in a difficult regime. in Nuclear Engineering International, 41(498), 40-41.
https://hdl.handle.net/21.15107/rcub_vinar_1955

Pešić MP, Cupać S. Working in a difficult regime. in Nuclear Engineering International. 1996;41(498):40-41.
https://hdl.handle.net/21.15107/rcub_vinar_1955 .

Pešić, Milan P., Cupać, Svjetlana, "Working in a difficult regime" in Nuclear Engineering International, 41, no. 498 (1996):40-41,
https://hdl.handle.net/21.15107/rcub_vinar_1955 .