Pub Date : 2019-07-08DOI: 10.1504/IJNEST.2019.10022535
B. Khuwaileh, A. Ishag
Nuclear energy is a promising source of power, proven viable in the cogeneration of electricity and water alike. However, a major challenge for (but not limited to) nuclear power generation is the maximisation of the power plant efficiency. Operating power plants with high to maximum efficiency has a profound effect on energy prices and environmental conditions for obvious reasons. One proposed strategy is to utilise energy storage systems for later discharge of power. However, this option entails transmission losses and a considerable capital cost. Therefore, this work explores the potential of water desalination as a proxy for energy storage systems in nuclear power plants. The current work explores various water desalination technologies and compares their performance in terms of the economics, water quality and production capacity. Three case studies have been adapted including APR1400, SMART and NuScale technologies. On the desalination side, Reverse Osmosis (RO), Multi-Stage Flash (MSF), Multi-Effect Distillation (MED) and hybrid combinations were studied. Results indicate that various desalination techniques can replace energy storage systems with justifiable capital cost and yet provide fresh water with acceptable quality. Specifically, RO can use the excess power produced via nuclear reactors during low demand periods with relatively low costs, without introducing new radiation release pathways.
{"title":"On the potential of water desalination as a proxy for energy storage systems in nuclear power plants","authors":"B. Khuwaileh, A. Ishag","doi":"10.1504/IJNEST.2019.10022535","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10022535","url":null,"abstract":"Nuclear energy is a promising source of power, proven viable in the cogeneration of electricity and water alike. However, a major challenge for (but not limited to) nuclear power generation is the maximisation of the power plant efficiency. Operating power plants with high to maximum efficiency has a profound effect on energy prices and environmental conditions for obvious reasons. One proposed strategy is to utilise energy storage systems for later discharge of power. However, this option entails transmission losses and a considerable capital cost. Therefore, this work explores the potential of water desalination as a proxy for energy storage systems in nuclear power plants. The current work explores various water desalination technologies and compares their performance in terms of the economics, water quality and production capacity. Three case studies have been adapted including APR1400, SMART and NuScale technologies. On the desalination side, Reverse Osmosis (RO), Multi-Stage Flash (MSF), Multi-Effect Distillation (MED) and hybrid combinations were studied. Results indicate that various desalination techniques can replace energy storage systems with justifiable capital cost and yet provide fresh water with acceptable quality. Specifically, RO can use the excess power produced via nuclear reactors during low demand periods with relatively low costs, without introducing new radiation release pathways.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41649957","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 : 2019-07-08DOI: 10.1504/IJNEST.2019.10022534
Abdessamad DIDI, A. Dadouch, M. Bencheikh, Otman Jai, O. Hajjaji
The spallation target plays an important role in the construction of an accelerator-driven system. Its purpose is to generate a neutron flux produced by cascaded spallation reactions using heavy nuclei, the latter being bombarded by the high-intensity proton beam. In this study, we have examined several heavy materials, such as uranium, thorium, tungsten, tantalum, lead, bismuth and mercury. The aim of this is to optimise a high-intensity neutron flux to be useful in several fields of applications, such as medicine and transmutation of nuclear waste. In this paper, we have shown when the spallation target changes the neutron flux varies. For example, uranium and thorium two materials producing a very intense amount of neutrons followed by lead, tungsten, mercury and bismuth and lately tantalum. We found these results by the variation of the proton beams energy from 0.1 GeV to 3 GeV, then with the variation of the geometry. Finally, we validated this study with experimental and theoretical results.
{"title":"New study of various target neutron yields from spallation reactions using a high-energy proton beam","authors":"Abdessamad DIDI, A. Dadouch, M. Bencheikh, Otman Jai, O. Hajjaji","doi":"10.1504/IJNEST.2019.10022534","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10022534","url":null,"abstract":"The spallation target plays an important role in the construction of an accelerator-driven system. Its purpose is to generate a neutron flux produced by cascaded spallation reactions using heavy nuclei, the latter being bombarded by the high-intensity proton beam. In this study, we have examined several heavy materials, such as uranium, thorium, tungsten, tantalum, lead, bismuth and mercury. The aim of this is to optimise a high-intensity neutron flux to be useful in several fields of applications, such as medicine and transmutation of nuclear waste. In this paper, we have shown when the spallation target changes the neutron flux varies. For example, uranium and thorium two materials producing a very intense amount of neutrons followed by lead, tungsten, mercury and bismuth and lately tantalum. We found these results by the variation of the proton beams energy from 0.1 GeV to 3 GeV, then with the variation of the geometry. Finally, we validated this study with experimental and theoretical results.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49488986","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 : 2019-07-08DOI: 10.1504/IJNEST.2019.10022533
Q. J. Tarbool, K. S. Jassim, A. Abojassim
Itemised studies of the surface property of the inter-nucleus potential in heavy-ion reactions have been achieved using large-angle quasi-elastic scattering at sub-barrier energies close to the Coulomb barrier height for 63Li + 6430Zn and 73Li + 6430Zn systems. In this paper, the nuclear potential has been expressed using Wood Saxon (WS) formula. The effect of rotational deformation was included for the nucleus 6430Zn with ground state rotational band up to the 4+ states. The Single-Channel (SC) and Coupled-Channels (CC) calculations have been carried out to elicit the diffuseness parameter of the nuclear potential as well as the potential depth, these calculations have been performed by using (CQEL) program which is considered the latest version of computer code (CCFULL). The chi square method, χ², played an important role in determining the best-fitted value of the diffuseness parameter. Through CC calculations with inert projectile and rotational target for 63Li + 6430Zn and 73Li + 6430Zn systems, full compatibility of diffuseness parameter has been achieved with the standard value 0.63 fm with χ² = 0.130 and χ² = 0.163 respectively, while the SC calculations give 0.64 fm and 0.65 fm, respectively.
{"title":"Surface diffuseness parameter with quasi-elastic scattering for some heavy-ion systems","authors":"Q. J. Tarbool, K. S. Jassim, A. Abojassim","doi":"10.1504/IJNEST.2019.10022533","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10022533","url":null,"abstract":"Itemised studies of the surface property of the inter-nucleus potential in heavy-ion reactions have been achieved using large-angle quasi-elastic scattering at sub-barrier energies close to the Coulomb barrier height for 63Li + 6430Zn and 73Li + 6430Zn systems. In this paper, the nuclear potential has been expressed using Wood Saxon (WS) formula. The effect of rotational deformation was included for the nucleus 6430Zn with ground state rotational band up to the 4+ states. The Single-Channel (SC) and Coupled-Channels (CC) calculations have been carried out to elicit the diffuseness parameter of the nuclear potential as well as the potential depth, these calculations have been performed by using (CQEL) program which is considered the latest version of computer code (CCFULL). The chi square method, χ², played an important role in determining the best-fitted value of the diffuseness parameter. Through CC calculations with inert projectile and rotational target for 63Li + 6430Zn and 73Li + 6430Zn systems, full compatibility of diffuseness parameter has been achieved with the standard value 0.63 fm with χ² = 0.130 and χ² = 0.163 respectively, while the SC calculations give 0.64 fm and 0.65 fm, respectively.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41774812","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 : 2019-04-26DOI: 10.1504/ijnest.2019.10020879
Carlos Filio López, A. Torres, Jaime Hernandez Galeana, Juan Galicia Aragon
A model of the Triga Mark III reactor of the National Institute for Nuclear Research (ININ) of Mexico was developed with the Monte Carlo codes Serpent and MCNP. The models were verified and validated by means of the experiments carried out as part of the starting tests to change the mixed fuel to Low Enrichment Fuels (LEU) of the TRIGA reactor core. The reactor data used in the V&V process consisted of fuel loading measurements, simulating the different stages of loading of fuel elements to the core to reach the reactor core criticality and the additional loading to achieve the reactivity excess to operate the reactor, as well as the evaluation of the shutdown margin reactivity and the control rods worth. The validated models constitute a trustworthy computational tool to analyse the most important neutronic core parameters as well as to have the numerical capabilities for fuel utilisation studies and analysis for extension of experimental facilities.
墨西哥国家核研究所（ININ）的Triga Mark III反应堆的模型是用蒙特卡罗代码Serpent和MCNP开发的。这些模型通过作为启动试验的一部分进行的实验进行了验证和验证，以将TRIGA反应堆堆芯的混合燃料改为低浓缩燃料（LEU）。V&V过程中使用的反应堆数据包括燃料装载测量，模拟燃料元件装载到堆芯以达到堆芯临界的不同阶段，以及实现反应堆运行反应性过剩的额外装载，以及停堆裕度反应性和控制棒价值的评估。经验证的模型构成了一个值得信赖的计算工具，可以分析最重要的中子核心参数，并具有燃料利用率研究和实验设施扩展分析的数值能力。
{"title":"Fuel loading, criticality and control rod worth calculations of the Triga Mark III reactor using Serpent and MCNP","authors":"Carlos Filio López, A. Torres, Jaime Hernandez Galeana, Juan Galicia Aragon","doi":"10.1504/ijnest.2019.10020879","DOIUrl":"https://doi.org/10.1504/ijnest.2019.10020879","url":null,"abstract":"A model of the Triga Mark III reactor of the National Institute for Nuclear Research (ININ) of Mexico was developed with the Monte Carlo codes Serpent and MCNP. The models were verified and validated by means of the experiments carried out as part of the starting tests to change the mixed fuel to Low Enrichment Fuels (LEU) of the TRIGA reactor core. The reactor data used in the V&V process consisted of fuel loading measurements, simulating the different stages of loading of fuel elements to the core to reach the reactor core criticality and the additional loading to achieve the reactivity excess to operate the reactor, as well as the evaluation of the shutdown margin reactivity and the control rods worth. The validated models constitute a trustworthy computational tool to analyse the most important neutronic core parameters as well as to have the numerical capabilities for fuel utilisation studies and analysis for extension of experimental facilities.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43597760","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 : 2019-04-26DOI: 10.1504/IJNEST.2019.10020874
J. François, E. García-Cervantes, C. Martin-Del-Campo
The objective of this paper is to study the reactivity behaviour and the isotopic fuel performance of an ASTRID-like reactor under a breed and burn strategy for two core designs: with oxide and with metallic fuel. The proposed reshuffling scheme was simulated with MCNP6 and the JEFF-3.2 cross-sections library to extend the fuel life by two more cycles. Our findings showed that the implementation of the reshuffling scheme enhanced the fuel utilisation, obtaining cycle extensions of 805 days and 1305 days, over the first 365 days, for the oxide and metallic fuelled designs, respectively. For the metallic-fuelled design the breeding of Pu-239 achieved a production of 577.7 kg, which represents a production rate of 126.3 kg/EFPY. The conversion rate for the metallic-fuelled design was 1.06 and 0.96 for the oxide-fuelled design. Regarding the coolant void reactivity worth, for the oxide-fuelled design it becomes positive after the second fuel reshuffling.
{"title":"A preliminary comparative study between oxide and metallic fuelled ASTRID-like reactor under a B&B strategy","authors":"J. François, E. García-Cervantes, C. Martin-Del-Campo","doi":"10.1504/IJNEST.2019.10020874","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10020874","url":null,"abstract":"The objective of this paper is to study the reactivity behaviour and the isotopic fuel performance of an ASTRID-like reactor under a breed and burn strategy for two core designs: with oxide and with metallic fuel. The proposed reshuffling scheme was simulated with MCNP6 and the JEFF-3.2 cross-sections library to extend the fuel life by two more cycles. Our findings showed that the implementation of the reshuffling scheme enhanced the fuel utilisation, obtaining cycle extensions of 805 days and 1305 days, over the first 365 days, for the oxide and metallic fuelled designs, respectively. For the metallic-fuelled design the breeding of Pu-239 achieved a production of 577.7 kg, which represents a production rate of 126.3 kg/EFPY. The conversion rate for the metallic-fuelled design was 1.06 and 0.96 for the oxide-fuelled design. Regarding the coolant void reactivity worth, for the oxide-fuelled design it becomes positive after the second fuel reshuffling.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45791980","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 : 2019-04-26DOI: 10.1504/IJNEST.2019.10020878
K. Khattab, G. Saba
The calculated weekly specific activities of 99Mo and 99mTc produced from the irradiation of the MoO3 targets in the Miniature Neutron Source Reactor (MNSR) are presented in this paper. The productions of the 99Mo and 99mTc were modelled and calculated through a set of differential equations using the Mathcad program. The resonance self-shielding factor (Gres) was calculated using the MATSSF and MCNP4C codes. The effects of the physical parameters such as the neutron flux and irradiation time on the weekly specific activities of 99Mo and 99mTc have been analysed. It was found that the optimum irradiation scheme was achieved when the MNSR was operated for an extended period of 5 hours a day for 5 days a week at the neutron flux of 7.5×1011 n.cm‒2.s‒1. The weekly specific activities for the 99Mo and 99mTc which can be produced in the MNSR, filling each of the five inner irradiation sites with 20 g of the MoO3 targets and using the regular natural convection method to cool the reactor, were 7.20 and 5.05 mCi.g‒1, respectively.
本文介绍了微型中子源反应堆(MNSR)中MoO3靶体辐照产生的99Mo和99mTc的周比活度计算结果。使用Mathcad程序通过一组微分方程对99Mo和99mTc的生产进行了建模和计算。利用MATSSF和MCNP4C代码计算了谐振自屏蔽系数(Gres)。分析了中子通量和辐照时间等物理参数对99Mo和99mTc周比活度的影响。结果表明，在中子通量为7.5×1011 n.cm-2.s-1的情况下，每周5天，每天延长运行5小时，可获得最佳辐照方案。在用20 g MoO3靶材填充5个内辐照点并采用常规自然对流方式冷却反应堆时，微核反应堆可产生的99Mo和99mTc的周比活度分别为7.20和5.05 mCi。分别g1。
{"title":"Feasibility study for production of 99Mo and 99mTc by the neutron activation of 98Mo in the MNSR reactor","authors":"K. Khattab, G. Saba","doi":"10.1504/IJNEST.2019.10020878","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10020878","url":null,"abstract":"The calculated weekly specific activities of 99Mo and 99mTc produced from the irradiation of the MoO3 targets in the Miniature Neutron Source Reactor (MNSR) are presented in this paper. The productions of the 99Mo and 99mTc were modelled and calculated through a set of differential equations using the Mathcad program. The resonance self-shielding factor (Gres) was calculated using the MATSSF and MCNP4C codes. The effects of the physical parameters such as the neutron flux and irradiation time on the weekly specific activities of 99Mo and 99mTc have been analysed. It was found that the optimum irradiation scheme was achieved when the MNSR was operated for an extended period of 5 hours a day for 5 days a week at the neutron flux of 7.5×1011 n.cm‒2.s‒1. The weekly specific activities for the 99Mo and 99mTc which can be produced in the MNSR, filling each of the five inner irradiation sites with 20 g of the MoO3 targets and using the regular natural convection method to cool the reactor, were 7.20 and 5.05 mCi.g‒1, respectively.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43003830","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 : 2019-04-26DOI: 10.1504/IJNEST.2019.10020880
H. Qadr
The purpose of this work was to investigate and quantify the linear attenuation coefficient and the buildup factor for different materials. The linear attenuation coefficient of absorber materials such as graphite was (0.097 cm‒1), whereas it was observed (0.136 cm‒1) for aluminium, and lead was (0.596 cm‒1). By using the gamma radiation energies emitted from 60Co source with 1332 keV, experimental and theoretical values are in a good agreement. Attenuation coefficient was measured by using counts of good geometry and bad geometry. The result shows that the linear attenuation is higher for lead and better radiation shielding compared with graphite and aluminium. Furthermore, buildup factor decreases with increasing thickness of the absorber material.
{"title":"Calculation for gamma ray buildup factor for aluminium, graphite and lead","authors":"H. Qadr","doi":"10.1504/IJNEST.2019.10020880","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10020880","url":null,"abstract":"The purpose of this work was to investigate and quantify the linear attenuation coefficient and the buildup factor for different materials. The linear attenuation coefficient of absorber materials such as graphite was (0.097 cm‒1), whereas it was observed (0.136 cm‒1) for aluminium, and lead was (0.596 cm‒1). By using the gamma radiation energies emitted from 60Co source with 1332 keV, experimental and theoretical values are in a good agreement. Attenuation coefficient was measured by using counts of good geometry and bad geometry. The result shows that the linear attenuation is higher for lead and better radiation shielding compared with graphite and aluminium. Furthermore, buildup factor decreases with increasing thickness of the absorber material.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45449779","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 : 2019-04-26DOI: 10.1504/IJNEST.2019.10020881
I. B. Rivas-Ortiz, D. S. Dominguez, C. Hernández, S. M. Iglesias
At present, neutron density calculation in non-multiplying media is relevant in many areas of engineering and science. In this paper, we propose the Extended Linear Discontinuous (ELD) method in multi-group discrete ordinates formulation, originally formulated for one-energy group fixed-source problems with isotropic scattering source in slab geometry. The proposed auxiliary equations are uncoupled on angular directions and combine the linear discontinuous approximation of the finite element method and the quasi-analytical general solution of the spectral nodal method. Thus, we can implement an efficient and simple algorithm using the conventional source iteration scheme for the sweeping equations. Numerical results for benchmark problems are presented to illustrate the accuracy and computational performance of the ELD method. The work shows that the main advantages of the proposed method are that the numerical scheme is stable for coarse-meshes, and its numerical results are more accurate than those generated by the Diamond Difference (DD) and Linear Discontinuous (LD) methods.
{"title":"A multi-group extended linear discontinuous method for fixed-source discrete ordinates problems in slab geometry","authors":"I. B. Rivas-Ortiz, D. S. Dominguez, C. Hernández, S. M. Iglesias","doi":"10.1504/IJNEST.2019.10020881","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10020881","url":null,"abstract":"At present, neutron density calculation in non-multiplying media is relevant in many areas of engineering and science. In this paper, we propose the Extended Linear Discontinuous (ELD) method in multi-group discrete ordinates formulation, originally formulated for one-energy group fixed-source problems with isotropic scattering source in slab geometry. The proposed auxiliary equations are uncoupled on angular directions and combine the linear discontinuous approximation of the finite element method and the quasi-analytical general solution of the spectral nodal method. Thus, we can implement an efficient and simple algorithm using the conventional source iteration scheme for the sweeping equations. Numerical results for benchmark problems are presented to illustrate the accuracy and computational performance of the ELD method. The work shows that the main advantages of the proposed method are that the numerical scheme is stable for coarse-meshes, and its numerical results are more accurate than those generated by the Diamond Difference (DD) and Linear Discontinuous (LD) methods.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48176565","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 : 2019-04-26DOI: 10.1504/IJNEST.2019.10020877
H. Manjunatha, L. Seenappa, K. Sridhar
We have formulated a simple semi-empirical formulae for photon energy absorption buildup factors in the energy region 0.015-15 MeV, atomic number range 1≤Z≤92 and for mean free path up to 40 mfp. The results produced by the present formulae agree well with the data available in the literature. This semi-empirical formula may be extended to any compounds/mixtures/biological samples. This semi-empirical formula finds importance in the calculations of buildup factors of any materials which are required for radiation shielding, nuclear engineering, radiotherapy and nuclear medicine.
{"title":"Semi-empirical formula for photon energy absorption buildup factors of elements and compounds","authors":"H. Manjunatha, L. Seenappa, K. Sridhar","doi":"10.1504/IJNEST.2019.10020877","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10020877","url":null,"abstract":"We have formulated a simple semi-empirical formulae for photon energy absorption buildup factors in the energy region 0.015-15 MeV, atomic number range 1≤Z≤92 and for mean free path up to 40 mfp. The results produced by the present formulae agree well with the data available in the literature. This semi-empirical formula may be extended to any compounds/mixtures/biological samples. This semi-empirical formula finds importance in the calculations of buildup factors of any materials which are required for radiation shielding, nuclear engineering, radiotherapy and nuclear medicine.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44304062","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 : 2019-04-26DOI: 10.1504/IJNEST.2019.10020882
U. Shukla, Rishabh Bajpai
With the increasing demand in electricity and rising temperature of Earth due to global warming, nuclear power plants can address the current needs. Development in the realm of nuclear energy has become a necessity in order to fulfil the present need. The present paper will summarise the basic knowledge regarding the nuclear power plant and current status of nuclear energy in India. Moreover, the paper presents some limitations to nuclear energy. This review paper will be helpful for the beginners in the field of nuclear power plant.
{"title":"Nuclear power plants in India: achieving clean and green energy","authors":"U. Shukla, Rishabh Bajpai","doi":"10.1504/IJNEST.2019.10020882","DOIUrl":"https://doi.org/10.1504/IJNEST.2019.10020882","url":null,"abstract":"With the increasing demand in electricity and rising temperature of Earth due to global warming, nuclear power plants can address the current needs. Development in the realm of nuclear energy has become a necessity in order to fulfil the present need. The present paper will summarise the basic knowledge regarding the nuclear power plant and current status of nuclear energy in India. Moreover, the paper presents some limitations to nuclear energy. This review paper will be helpful for the beginners in the field of nuclear power plant.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42114627","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}
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