Esmat Amin - Academia.edu (original) (raw)
Papers by Esmat Amin
Applied Radiation and Isotopes, 2016
This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative ... more This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code.
Axial burnup is an important factor in criticality safety and thermo-hydraulic calculation. In th... more Axial burnup is an important factor in criticality safety and thermo-hydraulic calculation. In the present study, the effect of axial distribution of burnup on power distribution and isotope inventory were evaluated using combination of the WIMSD-5B and MCNP5 codes for PWR fuel rods assembly. The fuel rods of assembly were divided into 5, 8 and 10 zones. The system was normalized to the total thermal power 16.89172 MWth, which was derived using the fuel assembly load of 458.599 kg. The MCNP5 code has been used to perform three dimensional neutron physics analysis while WIMSD-5B was used for generation of number densities at various stages of fuel burnup. The results were compared with data from the Takahama-3 Benchmark. The calculated-to-experimental and previous calculated results for ratios of U-235, U-236, U-238 and many important nuclides show good agreement. Both results for the isotopic inventory and the power distribution emphasize the importance of considering the axial vari...
The Accelerator Driven Systems (ADS), although receive a great deal of attention of many research... more The Accelerator Driven Systems (ADS), although receive a great deal of attention of many researchers worldwide, is still in the development stages. Transmutation of plutonium and minor actinides in the ADS is being envisaged for the purpose of reducing the long-term radiotoxic inventory of spent fuel in a possibly, cleaner and safer way than at present and producing energy and neutron sources. High energy accelerators appear to be a promising way to incinerate heavy actinides. The ADS consist of a sub-critical assembly driven by accelerator delivering a proton beam on a target to produce neutrons by spallation. The target constitutes the physical and functional interface between the accelerator and the sub-critical reactor. For this reason it is probably the most innovative component of the ADS. The target design is a key issue to investigate in studying the ADS performances, namely the number of neutrons emitted per incident particle, the mean energy deposited in the target per neu...
Scientific Reports, 2019
Thorium-plutonium mixed oxide, (Th,Pu)OX, is currently used as an alternative fuel in the light w... more Thorium-plutonium mixed oxide, (Th,Pu)OX, is currently used as an alternative fuel in the light water reactors in the world. The main objective of this paper is not only to show the benefits of using the thorium, but mainly to study how the way thorium is introduced in the fuel affects the neutron parameters. Among these benefits is the possibility of extending the operating cycle length and the reduction of the increasing stockpiles of plutonium. The first investigated method is introducing thorium as (Th,Pu)OX. The second one is a homogeneous model of thorium plutonium oxide. It is carried out by adding an amount of plutonium separated from the uranium oxide cycle at 50 GWd/ton of heavy metal to the same amount of thorium. Thus, we studied three assemblies; the reference assembly is uranium oxide of 4.2% enrichment containing borated water as a moderator of concentration 500 ppm (part per million) of B-10. The second is a (Th,Pu)OX and the third one is an assembly with homogenized...
Journal of Nuclear Energy. Parts A/B. Reactor Science and Technology, 1965
... equation (1). As a result, using the convolution rule for the integral terms, we obtain the e... more ... equation (1). As a result, using the convolution rule for the integral terms, we obtain the equation for the Laplace transform of the function'F(u) denoted by q(p): T(p) = hj99(p)Fj(p) 1 h;T(p)PZ(p) + 1, (A.1) where Fj(p) fj(u)evu du 1 +p fi [1 exp [r.(1 p)1 (A.2) 700 MYA YOUSEF, EA ...
Routine neutron calculations for a nuclear power reactor are usually performed using deterministi... more Routine neutron calculations for a nuclear power reactor are usually performed using deterministic transport and diffusion codes. Such codes need few group cross sections. The group cross-sections are generated using transport codes. For the global reactor calculations one needs cell parameters averaged over the fuel assembly. The reactor assembly geometry and materials are reactor type dependent. In the present work few group cross sections averaged over the fuel assembly are generated using MCNP5 for five types of generation IV reactors namely; HTGR, HTGR+BP, SCWR, UOX/MOX, and SFR together with PWR and VVER. The results are compared with previous calculations. The present results have relative statistical error in the range of 2% to 4%. The present results are compared with published calculations and a good agreement was obtained.
In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were perform... more In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed usingMCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. (e effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. (e reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. (e calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff and the Doppler coefficient (DC) are found to be in agreement with the design values. It is found that the core loaded with uranium and thorium has lower delayed neutron fraction than the uranium oxide core.(emoderator temperature coefficients of the uranium-thorium core are found to be higher than those of the uranium core. Results indicated that thorium has lower production of minor actinides (MAs)...
Radiation Physics and Chemistry
Progress in Nuclear Energy
Nuclear Technology and Radiation Protection
In this paper, the feasibility of high-level radioactive waste transmutation in accelerator drive... more In this paper, the feasibility of high-level radioactive waste transmutation in accelerator driven system sub-critical reactor assembly, has been studied for two zone's model and with three different core configurations. The inner zone has a fast neutron spectrum and the outer one has a thermal neutron spectrum. The subcritical core is coupled with external neutron source of energy 14 MeV (D-T source). The effects of high level waste isotopes sample (238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm, and 245Cm) distribution on the neutron spectrum and burnup performance in the inner zone have been investigated and discussed, by proposed three core configurations non-uniform, uniform, and spiral. The burnup calculations have been performed for one-year operation cycle for all the all proposed models. This work shows that one can effectively transmute most of the actual minor actinides isotopes in the inner fast spectrum zone of the proposed system, with optimal distribution ...
Arab Journal of Nuclear Sciences and Applications
The feasibility of utilizing Thorium-Plutonium Mixed Oxide in the Westinghouse AP1000 Advanced Pa... more The feasibility of utilizing Thorium-Plutonium Mixed Oxide in the Westinghouse AP1000 Advanced Passive pressurized water reactor is examined under steady-state, beginning of life (BOL) conditions. Initial core loading of the reactor consists of three types of UO 2 fuel assemblies with different enrichment in U235, as follows: 2.35w/o, 3.40w/o and 4.45w/o. In this paper, one-third of the UO 2 fuel assemblies are replaced by (Th-Pu)O 2 fuel assemblies in two arrangements: the first one assumes a blanket of (Th-Pu)O 2 fuel which replaces the 4.45% enriched UO 2 fuel assemblies surrounding the low enriched UO 2 fuel assemblies, and in the second arrangement some of the UO 2 fuel assemblies are replaced in a way creating a ring of (Th-Pu)O 2 fuel in the core. The reactor is modeled using QUARK computer code. The required cross-section data for QUARK calculations have been generated using WIMSD5 lattice cell code. The results of the steady state analysis show that introducing (Th-Pu)O 2 fuel into AP1000 would not negatively impact the reactor's safety as the criteria mentioned in design control document are met. For (Th-Pu)O 2 fuel blanket and ring loading arrangements, the calculated power peaking factor is less or equal to the design limit. Over the length of the hot channel, the Minimum Departure from Nucleate Boiling Ratio (MDNBR) varies with a minimum above the design limit for the considered (Th-Pu)O 2 fuel assemblies loading arrangements. This work provides the basis for studying Th-based fuel behavior and thermal hydraulic analysis of AP1000 using Th-based fuel in order to evaluate the safety aspect of various core loading patterns under anticipated and accidental conditions.
Annals of Nuclear Energy, 2016
International Journal of Nuclear Energy Science and Technology, 2015
Kerntechnik, 2015
The present study aims to showthe effect of different cross section libraries on MINERVE reactors... more The present study aims to showthe effect of different cross section libraries on MINERVE reactors afety parameters.T he MCNP5c alculation model of the MINERVE reactorf acility is used to determine core and safety parameters such as axial and radial fission rate distributions,c ontrol rodw orth and spectral indices.D ifferent neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor.MINERVE provides alarge experimental basis for the improvement of the cross section databases.T he current study calculatesthese parametersbyMCNP5code using three different cross section libraries(the continuous energy cross sections of the ENDFB-VI,T 16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO 2 ,R 1MOX, R2UO 2 ,a nd R2MOXc ore configurations,r epresentative of a LWRl oadedw ith UO 2 ,m ixed oxide matrix, over moderated UO 2 and over moderated MOX, respectively.T he study aims to determine the most suitable cross sectionl ibrary to be used with MCNP5r eactor calculation code for light water reactor fuel lattice.T he MCNP5 results were compared with the experimental results. Analyse von Sicherheitsparametern für den MINERVE-Reaktor. Ziel dieses Beitrags ist es,d en Einfluss verschiedener Wirkungsquerschnitts-Bibliotheken auf die Sicherheitsparameter des MINERVE Reaktors zu untersuchen. Das MCNP5-Rechenmodell der MINERVE-Reaktoranlage wird verwendet zur Bestimmung von Sicherheitsparametern wiez.B.axiale und radiale Spaltratenverteilung, Steuerstabgüte und spektralenI ndizes.V erschiedene Neutronenspektren wurden durch Änderung des experimentellen Gitters innerhalb des MINERVE-Reaktors erzeugt. MINERVE bietet eine große experimentelle Basis für die Verbesserung von Wirkungsquerschnitts-Datenbanken. In der derzeitigen Studie werden diese Parameterm it Hilfe des MCNP5-Codes unter Verwendung von drei verschiedenen Wirkungsquerschnitts-Bibliotheken (ENDFB-VI, T16-2003 und ENDF/B-VII.1) berechnet. Neutronenphysikalische Simulationsrechnungen wurden für R1UO 2 ,R 1MOX, R2UO 2 ,u nd R2MOX-Kernkonfigurationen durchgeführt. Ziel war es,d ie am besten geeignete Wirkungsquerschnitts-Bibliothek für die MCNP5-Berechnungen von Brennstoffgittern in Leichtwasserreaktorenz ub estimmen. Die MCNP5-Rechenergebnisse wurden mit experimentellen Ergebnissen verglichen.
Transmutation of plutonium and minor actinides in accelerator-driven systems (ADS) is being envis... more Transmutation of plutonium and minor actinides in accelerator-driven systems (ADS) is being envisaged for the purpose of reducing the long-term radiotoxic inventory of spent fuel nuclear reactor. The target is the physical and functional interface between the accelerator and the sub-critical reactor in the ADS, so it is probably the most innovative component of the ADS. The performances of ADS are characterized by the number of neutrons emitted per incident proton, the mean energy deposited in the target for neutron produced, the neutron spectrum, and the spallation product distribution. Neutron multiplicity is the number of neutrons produced per one beam particle. The present paper focuses on the production of neutrons in the spallation reactions. The neutrons produced in spallation reactions can be characterized by their energy and spatial distributions and multiplicity. This paper focus on the validation of Monte Carlo code (MCNPX) [1, 2] for study of neutrons production in the s...
The present study aims to assess radiation dose received by the target tumor cells and non-target... more The present study aims to assess radiation dose received by the target tumor cells and non-target organs during breast radiotherapy. The human body with its details, was modeled using three dimensional Monte Carlo Nuclear Particles Code (MCNP-4B). The dose distribution from 60 Co γ-rays, with average energy 1.25 MeV, in two fields in human theoretical model was calculated at selected points using the same code. Monte Carlo computer calculations of the photon spectra and the ratios of doses at the surfaces and in some of the internal organs of the model were also performed. In order to validate the calculated data, dose distribution at the same selected points were measured during a real radiotherapy practice using ARP phantom and TLD. Moreover, depth dose distribution was established by modeling water phantom with capacity 80 cm 3 and measured by ion chamber 0.125. Both measured and calculated data were compared. The comparison showed that the calculated and the measured data have t...
Applied Radiation and Isotopes, 2016
This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative ... more This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code.
Axial burnup is an important factor in criticality safety and thermo-hydraulic calculation. In th... more Axial burnup is an important factor in criticality safety and thermo-hydraulic calculation. In the present study, the effect of axial distribution of burnup on power distribution and isotope inventory were evaluated using combination of the WIMSD-5B and MCNP5 codes for PWR fuel rods assembly. The fuel rods of assembly were divided into 5, 8 and 10 zones. The system was normalized to the total thermal power 16.89172 MWth, which was derived using the fuel assembly load of 458.599 kg. The MCNP5 code has been used to perform three dimensional neutron physics analysis while WIMSD-5B was used for generation of number densities at various stages of fuel burnup. The results were compared with data from the Takahama-3 Benchmark. The calculated-to-experimental and previous calculated results for ratios of U-235, U-236, U-238 and many important nuclides show good agreement. Both results for the isotopic inventory and the power distribution emphasize the importance of considering the axial vari...
The Accelerator Driven Systems (ADS), although receive a great deal of attention of many research... more The Accelerator Driven Systems (ADS), although receive a great deal of attention of many researchers worldwide, is still in the development stages. Transmutation of plutonium and minor actinides in the ADS is being envisaged for the purpose of reducing the long-term radiotoxic inventory of spent fuel in a possibly, cleaner and safer way than at present and producing energy and neutron sources. High energy accelerators appear to be a promising way to incinerate heavy actinides. The ADS consist of a sub-critical assembly driven by accelerator delivering a proton beam on a target to produce neutrons by spallation. The target constitutes the physical and functional interface between the accelerator and the sub-critical reactor. For this reason it is probably the most innovative component of the ADS. The target design is a key issue to investigate in studying the ADS performances, namely the number of neutrons emitted per incident particle, the mean energy deposited in the target per neu...
Scientific Reports, 2019
Thorium-plutonium mixed oxide, (Th,Pu)OX, is currently used as an alternative fuel in the light w... more Thorium-plutonium mixed oxide, (Th,Pu)OX, is currently used as an alternative fuel in the light water reactors in the world. The main objective of this paper is not only to show the benefits of using the thorium, but mainly to study how the way thorium is introduced in the fuel affects the neutron parameters. Among these benefits is the possibility of extending the operating cycle length and the reduction of the increasing stockpiles of plutonium. The first investigated method is introducing thorium as (Th,Pu)OX. The second one is a homogeneous model of thorium plutonium oxide. It is carried out by adding an amount of plutonium separated from the uranium oxide cycle at 50 GWd/ton of heavy metal to the same amount of thorium. Thus, we studied three assemblies; the reference assembly is uranium oxide of 4.2% enrichment containing borated water as a moderator of concentration 500 ppm (part per million) of B-10. The second is a (Th,Pu)OX and the third one is an assembly with homogenized...
Journal of Nuclear Energy. Parts A/B. Reactor Science and Technology, 1965
... equation (1). As a result, using the convolution rule for the integral terms, we obtain the e... more ... equation (1). As a result, using the convolution rule for the integral terms, we obtain the equation for the Laplace transform of the function'F(u) denoted by q(p): T(p) = hj99(p)Fj(p) 1 h;T(p)PZ(p) + 1, (A.1) where Fj(p) fj(u)evu du 1 +p fi [1 exp [r.(1 p)1 (A.2) 700 MYA YOUSEF, EA ...
Routine neutron calculations for a nuclear power reactor are usually performed using deterministi... more Routine neutron calculations for a nuclear power reactor are usually performed using deterministic transport and diffusion codes. Such codes need few group cross sections. The group cross-sections are generated using transport codes. For the global reactor calculations one needs cell parameters averaged over the fuel assembly. The reactor assembly geometry and materials are reactor type dependent. In the present work few group cross sections averaged over the fuel assembly are generated using MCNP5 for five types of generation IV reactors namely; HTGR, HTGR+BP, SCWR, UOX/MOX, and SFR together with PWR and VVER. The results are compared with previous calculations. The present results have relative statistical error in the range of 2% to 4%. The present results are compared with published calculations and a good agreement was obtained.
In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were perform... more In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed usingMCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. (e effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. (e reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. (e calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff and the Doppler coefficient (DC) are found to be in agreement with the design values. It is found that the core loaded with uranium and thorium has lower delayed neutron fraction than the uranium oxide core.(emoderator temperature coefficients of the uranium-thorium core are found to be higher than those of the uranium core. Results indicated that thorium has lower production of minor actinides (MAs)...
Radiation Physics and Chemistry
Progress in Nuclear Energy
Nuclear Technology and Radiation Protection
In this paper, the feasibility of high-level radioactive waste transmutation in accelerator drive... more In this paper, the feasibility of high-level radioactive waste transmutation in accelerator driven system sub-critical reactor assembly, has been studied for two zone's model and with three different core configurations. The inner zone has a fast neutron spectrum and the outer one has a thermal neutron spectrum. The subcritical core is coupled with external neutron source of energy 14 MeV (D-T source). The effects of high level waste isotopes sample (238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm, and 245Cm) distribution on the neutron spectrum and burnup performance in the inner zone have been investigated and discussed, by proposed three core configurations non-uniform, uniform, and spiral. The burnup calculations have been performed for one-year operation cycle for all the all proposed models. This work shows that one can effectively transmute most of the actual minor actinides isotopes in the inner fast spectrum zone of the proposed system, with optimal distribution ...
Arab Journal of Nuclear Sciences and Applications
The feasibility of utilizing Thorium-Plutonium Mixed Oxide in the Westinghouse AP1000 Advanced Pa... more The feasibility of utilizing Thorium-Plutonium Mixed Oxide in the Westinghouse AP1000 Advanced Passive pressurized water reactor is examined under steady-state, beginning of life (BOL) conditions. Initial core loading of the reactor consists of three types of UO 2 fuel assemblies with different enrichment in U235, as follows: 2.35w/o, 3.40w/o and 4.45w/o. In this paper, one-third of the UO 2 fuel assemblies are replaced by (Th-Pu)O 2 fuel assemblies in two arrangements: the first one assumes a blanket of (Th-Pu)O 2 fuel which replaces the 4.45% enriched UO 2 fuel assemblies surrounding the low enriched UO 2 fuel assemblies, and in the second arrangement some of the UO 2 fuel assemblies are replaced in a way creating a ring of (Th-Pu)O 2 fuel in the core. The reactor is modeled using QUARK computer code. The required cross-section data for QUARK calculations have been generated using WIMSD5 lattice cell code. The results of the steady state analysis show that introducing (Th-Pu)O 2 fuel into AP1000 would not negatively impact the reactor's safety as the criteria mentioned in design control document are met. For (Th-Pu)O 2 fuel blanket and ring loading arrangements, the calculated power peaking factor is less or equal to the design limit. Over the length of the hot channel, the Minimum Departure from Nucleate Boiling Ratio (MDNBR) varies with a minimum above the design limit for the considered (Th-Pu)O 2 fuel assemblies loading arrangements. This work provides the basis for studying Th-based fuel behavior and thermal hydraulic analysis of AP1000 using Th-based fuel in order to evaluate the safety aspect of various core loading patterns under anticipated and accidental conditions.
Annals of Nuclear Energy, 2016
International Journal of Nuclear Energy Science and Technology, 2015
Kerntechnik, 2015
The present study aims to showthe effect of different cross section libraries on MINERVE reactors... more The present study aims to showthe effect of different cross section libraries on MINERVE reactors afety parameters.T he MCNP5c alculation model of the MINERVE reactorf acility is used to determine core and safety parameters such as axial and radial fission rate distributions,c ontrol rodw orth and spectral indices.D ifferent neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor.MINERVE provides alarge experimental basis for the improvement of the cross section databases.T he current study calculatesthese parametersbyMCNP5code using three different cross section libraries(the continuous energy cross sections of the ENDFB-VI,T 16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO 2 ,R 1MOX, R2UO 2 ,a nd R2MOXc ore configurations,r epresentative of a LWRl oadedw ith UO 2 ,m ixed oxide matrix, over moderated UO 2 and over moderated MOX, respectively.T he study aims to determine the most suitable cross sectionl ibrary to be used with MCNP5r eactor calculation code for light water reactor fuel lattice.T he MCNP5 results were compared with the experimental results. Analyse von Sicherheitsparametern für den MINERVE-Reaktor. Ziel dieses Beitrags ist es,d en Einfluss verschiedener Wirkungsquerschnitts-Bibliotheken auf die Sicherheitsparameter des MINERVE Reaktors zu untersuchen. Das MCNP5-Rechenmodell der MINERVE-Reaktoranlage wird verwendet zur Bestimmung von Sicherheitsparametern wiez.B.axiale und radiale Spaltratenverteilung, Steuerstabgüte und spektralenI ndizes.V erschiedene Neutronenspektren wurden durch Änderung des experimentellen Gitters innerhalb des MINERVE-Reaktors erzeugt. MINERVE bietet eine große experimentelle Basis für die Verbesserung von Wirkungsquerschnitts-Datenbanken. In der derzeitigen Studie werden diese Parameterm it Hilfe des MCNP5-Codes unter Verwendung von drei verschiedenen Wirkungsquerschnitts-Bibliotheken (ENDFB-VI, T16-2003 und ENDF/B-VII.1) berechnet. Neutronenphysikalische Simulationsrechnungen wurden für R1UO 2 ,R 1MOX, R2UO 2 ,u nd R2MOX-Kernkonfigurationen durchgeführt. Ziel war es,d ie am besten geeignete Wirkungsquerschnitts-Bibliothek für die MCNP5-Berechnungen von Brennstoffgittern in Leichtwasserreaktorenz ub estimmen. Die MCNP5-Rechenergebnisse wurden mit experimentellen Ergebnissen verglichen.
Transmutation of plutonium and minor actinides in accelerator-driven systems (ADS) is being envis... more Transmutation of plutonium and minor actinides in accelerator-driven systems (ADS) is being envisaged for the purpose of reducing the long-term radiotoxic inventory of spent fuel nuclear reactor. The target is the physical and functional interface between the accelerator and the sub-critical reactor in the ADS, so it is probably the most innovative component of the ADS. The performances of ADS are characterized by the number of neutrons emitted per incident proton, the mean energy deposited in the target for neutron produced, the neutron spectrum, and the spallation product distribution. Neutron multiplicity is the number of neutrons produced per one beam particle. The present paper focuses on the production of neutrons in the spallation reactions. The neutrons produced in spallation reactions can be characterized by their energy and spatial distributions and multiplicity. This paper focus on the validation of Monte Carlo code (MCNPX) [1, 2] for study of neutrons production in the s...
The present study aims to assess radiation dose received by the target tumor cells and non-target... more The present study aims to assess radiation dose received by the target tumor cells and non-target organs during breast radiotherapy. The human body with its details, was modeled using three dimensional Monte Carlo Nuclear Particles Code (MCNP-4B). The dose distribution from 60 Co γ-rays, with average energy 1.25 MeV, in two fields in human theoretical model was calculated at selected points using the same code. Monte Carlo computer calculations of the photon spectra and the ratios of doses at the surfaces and in some of the internal organs of the model were also performed. In order to validate the calculated data, dose distribution at the same selected points were measured during a real radiotherapy practice using ARP phantom and TLD. Moreover, depth dose distribution was established by modeling water phantom with capacity 80 cm 3 and measured by ion chamber 0.125. Both measured and calculated data were compared. The comparison showed that the calculated and the measured data have t...