Pub Date : 2018-03-30DOI: 10.4236/WJNST.2018.82007
M. M. Haque, Alamin, Md Faruq Hossain, Md. Sanowar Hossain, Md. Selim Reza
In total 184 studies, which included thyroid uptake and scintigraphy, were performed in 68 hyperthyroid patients: 67% female and 33% male to investigate their thyroid conditions. The aim of the present study was to illustrate the role of uptake and scintigraphy tests in determining the thyroid status of hyperthyroid patients. The uptake study was performed by oral administration of 100 - 200 μCi of 131I as sodium-iodide and counting the radioactivity at 2 and 24 hrs, whereas thyroid scintigraphy was performed 20 minutes after an intravenous injection of 2 - 4 mCi of 99mTc-pertechnetate. The present results of thyroid uptake and scintigraphy successfully identified the thyroid condition in different states. The present results were also compared with some reported data and found to be fair in good agreement.
{"title":"Contribution of Radioiodine Thyroid Uptake and Scintigraphy to the Diagnosis of Hyperthyroidism","authors":"M. M. Haque, Alamin, Md Faruq Hossain, Md. Sanowar Hossain, Md. Selim Reza","doi":"10.4236/WJNST.2018.82007","DOIUrl":"https://doi.org/10.4236/WJNST.2018.82007","url":null,"abstract":"In total 184 studies, which included thyroid uptake and scintigraphy, were performed in 68 hyperthyroid patients: 67% female and 33% male to investigate their thyroid conditions. The aim of the present study was to illustrate the role of uptake and scintigraphy tests in determining the thyroid status of hyperthyroid patients. The uptake study was performed by oral administration of 100 - 200 μCi of 131I as sodium-iodide and counting the radioactivity at 2 and 24 hrs, whereas thyroid scintigraphy was performed 20 minutes after an intravenous injection of 2 - 4 mCi of 99mTc-pertechnetate. The present results of thyroid uptake and scintigraphy successfully identified the thyroid condition in different states. The present results were also compared with some reported data and found to be fair in good agreement.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"08 1","pages":"70-77"},"PeriodicalIF":0.0,"publicationDate":"2018-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44989398","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-03-30DOI: 10.4236/WJNST.2018.82006
P. E. Umbehaun, W. M. Torres, J. Souza, M. Yamaguchi, A. T. E. Silva, R. N. Mesquita, N. Scuro, D. A. Andrade
This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.
{"title":"Thermal Hydraulic Analysis Improvement for the IEA-R1 Research Reactor and Fuel Assembly Design Modification","authors":"P. E. Umbehaun, W. M. Torres, J. Souza, M. Yamaguchi, A. T. E. Silva, R. N. Mesquita, N. Scuro, D. A. Andrade","doi":"10.4236/WJNST.2018.82006","DOIUrl":"https://doi.org/10.4236/WJNST.2018.82006","url":null,"abstract":"This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"08 1","pages":"54-69"},"PeriodicalIF":0.0,"publicationDate":"2018-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48987843","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-03-30DOI: 10.4236/wjnst.2018.82003
K. Ibikunle, U. Sadiq, Y. V. Ibrahim, S. Jonah
The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Test Reactors (RERTR) program. Since 2006, the reduction in the fuel enrichment of MSNR facilities from greater than 90% HEU cores to less than 20% LEU cores has been embarked upon. Consequently in this work, the physics parameters of three dispersion LEU fuels, which include U3Si, U3Si2, and U9Mo enriched to 19.75% were determined by the MCNP code to investigate their suitability for the conversion of NIRR-1 to LEU. The following reactor core physics parameters were computed for the LEU fuel options: clean cold core excess reactivity (ρex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, βeff and prompt neutron lifetime, lf). Results are compared with experimental and calculated data of the current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of commercial MNSR in general and NIRR-1 in particular from HEU to LEU.
{"title":"MCNP Simulation of Physics Parameters of Dispersion Fuels for Conversion of NIRR-1 to LEU","authors":"K. Ibikunle, U. Sadiq, Y. V. Ibrahim, S. Jonah","doi":"10.4236/wjnst.2018.82003","DOIUrl":"https://doi.org/10.4236/wjnst.2018.82003","url":null,"abstract":"The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Test Reactors (RERTR) program. Since 2006, the reduction in the fuel enrichment of MSNR facilities from greater than 90% HEU cores to less than 20% LEU cores has been embarked upon. Consequently in this work, the physics parameters of three dispersion LEU fuels, which include U3Si, U3Si2, and U9Mo enriched to 19.75% were determined by the MCNP code to investigate their suitability for the conversion of NIRR-1 to LEU. The following reactor core physics parameters were computed for the LEU fuel options: clean cold core excess reactivity (ρex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, βeff and prompt neutron lifetime, lf). Results are compared with experimental and calculated data of the current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of commercial MNSR in general and NIRR-1 in particular from HEU to LEU.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"08 1","pages":"23-29"},"PeriodicalIF":0.0,"publicationDate":"2018-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43352065","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-01-26DOI: 10.4236/WJNST.2018.81002
F. Menegus
Ideas, solely related on the nuclear shell model, fail to give an interpretation of the experimental central role of 54Xe in the asymmetric fission of actinides. The same is true for the β-delayed fission of 180Tl to 80Kr and 100Ru. The representation of the natural isotopes, in the Z-Neutron Excess plane, suggests the importance of the of the Neutron Excess evolution mode in the fragments of the asymmetric actinide fission and in the fragments of the β-delayed fission of 180Tl. The evolution mode of the Neutron Excess, hinged at Kr and Xe, is directed by the 50 and 82 neutron magic numbers. The present isotope representation offers a frame for the interpretation of the post fission evaporation of neutrons, higher for the AL compared to the AH fragments, a tenet in nuclear fission. Further enlightened is the functional meaning of the 50 proton magic number, marking the start of the yield rise of the AH fragments in actinide fission.
{"title":"A Suggestion Complementing the Magic Numbers Interpretation of the Nuclear Fission Phenomena","authors":"F. Menegus","doi":"10.4236/WJNST.2018.81002","DOIUrl":"https://doi.org/10.4236/WJNST.2018.81002","url":null,"abstract":"Ideas, solely related on the nuclear shell model, fail to give an interpretation of the experimental central role of 54Xe in the asymmetric fission of actinides. The same is true for the β-delayed fission of 180Tl to 80Kr and 100Ru. The representation of the natural isotopes, in the Z-Neutron Excess plane, suggests the importance of the of the Neutron Excess evolution mode in the fragments of the asymmetric actinide fission and in the fragments of the β-delayed fission of 180Tl. The evolution mode of the Neutron Excess, hinged at Kr and Xe, is directed by the 50 and 82 neutron magic numbers. The present isotope representation offers a frame for the interpretation of the post fission evaporation of neutrons, higher for the AL compared to the AH fragments, a tenet in nuclear fission. Further enlightened is the functional meaning of the 50 proton magic number, marking the start of the yield rise of the AH fragments in actinide fission.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"08 1","pages":"11-22"},"PeriodicalIF":0.0,"publicationDate":"2018-01-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41450605","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2017-08-09DOI: 10.4236/WJNST.2017.74018
B. M. Mweetwa, E. Ampomah-Amoako, E. Akaho
The Program for the Analysis of Reactor Transients/Argonne National Laboratory (PARET/ANL) code was used to predict the thermal hydraulic behaviour of the Ghana Research Reactor-1 after adding 9.0 mm of beryllium to the top shim tray of the core. The core was analysed for reactivity insertions 2.1 mk, 3.0 mk, 4.0 mk, 5.0 mk and 6.7 mk, respectively. The reactor is still safe to operate in the range 2.1 mk to 4.0 mk. However, 2.1 mk would be ideal since the reactor automatic shutdown (SCRAM) is set not to exceed 120% of reactor nominal power.
{"title":"Transient Studies of Ghana Research Reactor-1 after Nineteen (19) Years of Operation Using PARET/ANL Code","authors":"B. M. Mweetwa, E. Ampomah-Amoako, E. Akaho","doi":"10.4236/WJNST.2017.74018","DOIUrl":"https://doi.org/10.4236/WJNST.2017.74018","url":null,"abstract":"The Program for the Analysis of Reactor Transients/Argonne National Laboratory (PARET/ANL) code was used to predict the thermal hydraulic behaviour of the Ghana Research Reactor-1 after adding 9.0 mm of beryllium to the top shim tray of the core. The core was analysed for reactivity insertions 2.1 mk, 3.0 mk, 4.0 mk, 5.0 mk and 6.7 mk, respectively. The reactor is still safe to operate in the range 2.1 mk to 4.0 mk. However, 2.1 mk would be ideal since the reactor automatic shutdown (SCRAM) is set not to exceed 120% of reactor nominal power.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"07 1","pages":"223-231"},"PeriodicalIF":0.0,"publicationDate":"2017-08-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47289380","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2017-08-09DOI: 10.4236/WJNST.2017.74022
Rubina Nasir
Effect of kinetic model parameters on fission product (I-129) activity from fuel to coolant in PWRs has been studied in this work. First a computational model was developed for fission product release into primary coolant using ORIGEN-2 as subroutine. The model is based on set of differential equations of kinetic model which includes fuel-to-gap release model, gap-to-coolant leakage model, and Booths diffusion model. A Matlab based computer program FPAPC (Fission Product Activity in Primary Coolant) was developed. Variations of I-129 activity in Primary Heat Transport System were computed and computed values of i-129 were found in good agreement and deviations were within 2% - 3% of already published data values. Finally, the effects of coolant purification rate, diffusion constant and gas escape rate on I-129 activity were studied and results indicated that the coolant purification rate is the most sensitive parameter for fission product activity in primary circuit. For changes of 5% in steps from −10% to +10% in the coolant purification rate constant (Β), the activity variation after 200 days of reactor operation was 23.1% for the change.
{"title":"Effect of Kinetic Parameters on I-129 Activity from Fuel to Coolant in PWRs","authors":"Rubina Nasir","doi":"10.4236/WJNST.2017.74022","DOIUrl":"https://doi.org/10.4236/WJNST.2017.74022","url":null,"abstract":"Effect of kinetic model parameters on fission product (I-129) activity from fuel to coolant in PWRs has been studied in this work. First a computational model was developed for fission product release into primary coolant using ORIGEN-2 as subroutine. The model is based on set of differential equations of kinetic model which includes fuel-to-gap release model, gap-to-coolant leakage model, and Booths diffusion model. A Matlab based computer program FPAPC (Fission Product Activity in Primary Coolant) was developed. Variations of I-129 activity in Primary Heat Transport System were computed and computed values of i-129 were found in good agreement and deviations were within 2% - 3% of already published data values. Finally, the effects of coolant purification rate, diffusion constant and gas escape rate on I-129 activity were studied and results indicated that the coolant purification rate is the most sensitive parameter for fission product activity in primary circuit. For changes of 5% in steps from −10% to +10% in the coolant purification rate constant (Β), the activity variation after 200 days of reactor operation was 23.1% for the change.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"07 1","pages":"284-291"},"PeriodicalIF":0.0,"publicationDate":"2017-08-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45046007","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2017-08-09DOI: 10.4236/WJNST.2017.74024
G. Singh, P. M. Khot, Pradeep Kumar, Chetan Baghra, R. Bhatt, P. G. Behere
This paper presents a study on the process engineering aspects of relevance to the industrial implementation of ThO2 and (Th, U)O2 mixed oxide (MOX) pellet type fuel manufacturing. The paper in particular focuses on the recycling of thoria based fuel production scrap which is an economically important component in the fuel manufacturing process. The thoria based fuels are envisaged for Advanced Heavy Water Reactor (AHWR) and other reactors important to the Indian Nuclear Power Programme. A process was developed for recycling the chemically clean, off-specification and defective sintered ThO2 and (Th, U)O2 MOX nuclear fuel pellets. ThO2 doesn’t undergo oxidation or reduction and thus, more traditional methods of recycling are impractical. The integrated process was developed by combining three basic approaches of recycling namely mechanical micronisation, air oxidation (for MOX) and microwave dissolution-denitration. A thorough investigation of the influence of several variables as heating method, UO2 content, fluoride and polyvinyl alcohol (PVA) addition during microwave dissolution-denitration was recorded on the product characteristics. The suitability evaluation of the recycled powder for re-fabrication of the fuel was carried out by analyzing the particle size, BET specific surface area, phase using XRD, bulk density and impurities. The physical and chemical properties of recycled powder obtained from the sintered (Th1-y, Uy)O2 (y; 0 - 30 wt%) pellets advocate 100% utilisation for fuel re-fabrication. Recycled ThO2 by integrated process showed distinctly high sinterability compared to standard powder evaluated in terms of surface area and particle size.
{"title":"An Integrated Process for Recycling of ThO 2 Based Mixed Oxide Rejected Nuclear Fuel Pellets","authors":"G. Singh, P. M. Khot, Pradeep Kumar, Chetan Baghra, R. Bhatt, P. G. Behere","doi":"10.4236/WJNST.2017.74024","DOIUrl":"https://doi.org/10.4236/WJNST.2017.74024","url":null,"abstract":"This paper presents a study on the process engineering aspects of relevance to the industrial implementation of ThO2 and (Th, U)O2 mixed oxide (MOX) pellet type fuel manufacturing. The paper in particular focuses on the recycling of thoria based fuel production scrap which is an economically important component in the fuel manufacturing process. The thoria based fuels are envisaged for Advanced Heavy Water Reactor (AHWR) and other reactors important to the Indian Nuclear Power Programme. A process was developed for recycling the chemically clean, off-specification and defective sintered ThO2 and (Th, U)O2 MOX nuclear fuel pellets. ThO2 doesn’t undergo oxidation or reduction and thus, more traditional methods of recycling are impractical. The integrated process was developed by combining three basic approaches of recycling namely mechanical micronisation, air oxidation (for MOX) and microwave dissolution-denitration. A thorough investigation of the influence of several variables as heating method, UO2 content, fluoride and polyvinyl alcohol (PVA) addition during microwave dissolution-denitration was recorded on the product characteristics. The suitability evaluation of the recycled powder for re-fabrication of the fuel was carried out by analyzing the particle size, BET specific surface area, phase using XRD, bulk density and impurities. The physical and chemical properties of recycled powder obtained from the sintered (Th1-y, Uy)O2 (y; 0 - 30 wt%) pellets advocate 100% utilisation for fuel re-fabrication. Recycled ThO2 by integrated process showed distinctly high sinterability compared to standard powder evaluated in terms of surface area and particle size.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"7 1","pages":"309-330"},"PeriodicalIF":0.0,"publicationDate":"2017-08-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48687309","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2017-08-09DOI: 10.4236/WJNST.2017.74019
A. Ivanchin
Theoretical physics makes a wide use of differential equations for which only a potential solution is applied. The possibility that these equations may have a non-potential solution is ruled out and not considered. In this paper an exact non-potential solution of the continuity equation is described. The electric field of an elementary charged particle consists of two components: the known Potential Component (PC) produced by the charge and the earlier unknown Non-potential Component (NC) with a zero charge. Charged particles have both components, while a neutron has only the NC. The proton and neutron NC ensures similarity of their properties. The PC is spherically symmetric and NC is axisymmetric. Therefore, to describe an elementary particle, one should take into account both its spatial coordinates and the NC orientation. The particle interaction is determined by their NC mutual orientation. Neglecting the latter leads to indefiniteness of the interaction result. In a homogeneous electric field, the force acting on the NC is zero. Therefore, a charged particle possessing the NC will behave like a potential one. In an inhomogeneous field, the situation is principally different. Due to the NC there occurs an interaction between a neutron and a proton. The non-potential field results in the existence of two types of neutrons: a neutron and an antineutron. A neutron repels from a proton ensuring scattering of neutrons on protons. An antineutron is attracted to a proton leading to its annihilation. The NC produces the magnetic dipole moment of an elementary particle.
{"title":"Electrostatic Theory of Elementary Particles","authors":"A. Ivanchin","doi":"10.4236/WJNST.2017.74019","DOIUrl":"https://doi.org/10.4236/WJNST.2017.74019","url":null,"abstract":"Theoretical physics makes a wide use of differential equations for which only a potential solution is applied. The possibility that these equations may have a non-potential solution is ruled out and not considered. In this paper an exact non-potential solution of the continuity equation is described. The electric field of an elementary charged particle consists of two components: the known Potential Component (PC) produced by the charge and the earlier unknown Non-potential Component (NC) with a zero charge. Charged particles have both components, while a neutron has only the NC. The proton and neutron NC ensures similarity of their properties. The PC is spherically symmetric and NC is axisymmetric. Therefore, to describe an elementary particle, one should take into account both its spatial coordinates and the NC orientation. The particle interaction is determined by their NC mutual orientation. Neglecting the latter leads to indefiniteness of the interaction result. In a homogeneous electric field, the force acting on the NC is zero. Therefore, a charged particle possessing the NC will behave like a potential one. In an inhomogeneous field, the situation is principally different. Due to the NC there occurs an interaction between a neutron and a proton. The non-potential field results in the existence of two types of neutrons: a neutron and an antineutron. A neutron repels from a proton ensuring scattering of neutrons on protons. An antineutron is attracted to a proton leading to its annihilation. The NC produces the magnetic dipole moment of an elementary particle.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"07 1","pages":"232-251"},"PeriodicalIF":0.0,"publicationDate":"2017-08-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42314969","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2017-08-09DOI: 10.4236/WJNST.2017.74021
H. Borbón-Nuñez, C. Furetta
The aim of this paper is to give some simplified expressions related to the peak shape method. The modified equations have been used to calculate the activation energy (E) of commercial thermoluminescent dosimeters (TLD), as well as of ZnO thermoluminescent material produced in laboratory; the values so determined have been compared to the values obtained using the classical expressions of the peak shape method. The modified equations proposed are as a function of peak shape parameters or the peak temperature at the maximum. This expression could be useful to obtain approximated E values in the case of complex glow curves as well, when the peaks are not well resolved but the peak temperature at the maximum may be easily determined.
{"title":"Activation Energy of Modified Peak Shape Equations","authors":"H. Borbón-Nuñez, C. Furetta","doi":"10.4236/WJNST.2017.74021","DOIUrl":"https://doi.org/10.4236/WJNST.2017.74021","url":null,"abstract":"The aim of this paper is to give some simplified expressions related to the peak shape method. The modified equations have been used to calculate the activation energy (E) of commercial thermoluminescent dosimeters (TLD), as well as of ZnO thermoluminescent material produced in laboratory; the values so determined have been compared to the values obtained using the classical expressions of the peak shape method. The modified equations proposed are as a function of peak shape parameters or the peak temperature at the maximum. This expression could be useful to obtain approximated E values in the case of complex glow curves as well, when the peaks are not well resolved but the peak temperature at the maximum may be easily determined.","PeriodicalId":61566,"journal":{"name":"核科学与技术国际期刊(英文)","volume":"7 1","pages":"274-283"},"PeriodicalIF":0.0,"publicationDate":"2017-08-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41573450","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2017-08-09DOI: 10.4236/WJNST.2017.74020
M. Silverman
Residence time in a flow measurement of radioactivity is the time spent by a pre-determined quantity of radioactive sample in the flow cell. In a recent report of the measurement of indoor radon by passive diffusion in an open volume (i.e. no flow cell or control volume), the concept of residence time was generalized to apply to measurement conditions with random, rather than directed, flow. The generalization, leading to a quantity Δtr, involved use of a) a phenomenological alpha-particle range function to calculate the effective detection volume, and b) a phenomenological description of diffusion by Fick’s law to determine the effective flow velocity. This paper examines the residence time in passive diffusion from the micro-statistical perspective of single-particle continuous Brownian motion. The statistical quantity “mean residence time” Tr is derived from the Green’s function for unbiased single-particle diffusion and is shown to be consistent with Δtr. The finite statistical lifetime of the randomly moving radioactive atom plays an essential part. For stable particles, Tr is of infinite duration, whereas for an unstable particle (such as 222Rn), with diffusivity D and decay rate λ, Tr is approximately the effective size of the detection region divided by the characteristic diffusion velocity . Comparison of the mean residence time with the time of first passage (or exit time) in the theory of stochastic processes shows the conditions under which the two measures of time are equivalent and helps elucidate the connection between the phenomenological and statistical descriptions of radon diffusion.
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