Pub Date : 2017-11-21DOI: 10.1504/IJNEST.2017.10009081
M. I. Sayyed
With increasing use of ionising radiations (in particular the gamma rays and X-rays) in medicine, many dangerous diseases may occur. Hence, it is necessary to restrict and control exposure of human beings to these radiations. In this study we have investigated the radioprotective effectiveness of some medications sold at community pharmacies in Tabuk, Saudi Arabia. The data were collected and recorded for 20 drugs commonly used for different medical purposes. In order to investigate the effectiveness of these radioprotectives in terms of absorption of low and high energy photons, the effective atomic number (Zeff) of ten drugs for total photon interaction in the energy range of 1 keV to 15 MeV using WinXCom were calculated. In addition, by Geometric-Progression (G-P) method, the energy absorption (EABF) and exposure build-up factors (EBF) for incident photon energy 0.015 MeV to 15 MeV up to penetration depths of 40 mean free paths (mfp) were calculated for the ten drugs. Among the selected compounds, Captopril and Cefixime have the maximum value of Zeff, while the minimum EBF and EABF were found for Mesna, Cramastine, Thiotepa and Busflan; therefore, they are appealing as radioprotective compounds.
{"title":"Radioprotective potential of some medicines used in Tabuk, Saudi Arabia, to minimise the effects of the ionising radiations","authors":"M. I. Sayyed","doi":"10.1504/IJNEST.2017.10009081","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10009081","url":null,"abstract":"With increasing use of ionising radiations (in particular the gamma rays and X-rays) in medicine, many dangerous diseases may occur. Hence, it is necessary to restrict and control exposure of human beings to these radiations. In this study we have investigated the radioprotective effectiveness of some medications sold at community pharmacies in Tabuk, Saudi Arabia. The data were collected and recorded for 20 drugs commonly used for different medical purposes. In order to investigate the effectiveness of these radioprotectives in terms of absorption of low and high energy photons, the effective atomic number (Zeff) of ten drugs for total photon interaction in the energy range of 1 keV to 15 MeV using WinXCom were calculated. In addition, by Geometric-Progression (G-P) method, the energy absorption (EABF) and exposure build-up factors (EBF) for incident photon energy 0.015 MeV to 15 MeV up to penetration depths of 40 mean free paths (mfp) were calculated for the ten drugs. Among the selected compounds, Captopril and Cefixime have the maximum value of Zeff, while the minimum EBF and EABF were found for Mesna, Cramastine, Thiotepa and Busflan; therefore, they are appealing as radioprotective compounds.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"205"},"PeriodicalIF":0.0,"publicationDate":"2017-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47515139","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 : 2017-11-21DOI: 10.1504/IJNEST.2017.10009085
F. A. Majeed
The effect of the breakup channel on fusion reactions of weakly bound systems by means of semi-classical and full quantum mechanical approaches has been discussed. The total fusion reaction cross-section σfus and the fusion barrier distribution Dfus for the systems 4He+64Zn, 6Li+208Pb and 7Li+24Mg have been calculated. The inclusion of the breakup channel enhances the calculations of the fusion cross-section markedly below the Coulomb barrier and hindrance above the Coulomb barrier in comparison to the experimental data. The semi-classical calculations agree reasonably with the full quantum mechanical treatment and they were able to reproduce the experimental data in details for the total fusion reaction cross-section σfus and the fusion barrier distribution Dfus.
{"title":"The role of the breakup channel on the fusion reaction of light and weakly bound nuclei","authors":"F. A. Majeed","doi":"10.1504/IJNEST.2017.10009085","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10009085","url":null,"abstract":"The effect of the breakup channel on fusion reactions of weakly bound systems by means of semi-classical and full quantum mechanical approaches has been discussed. The total fusion reaction cross-section σfus and the fusion barrier distribution Dfus for the systems 4He+64Zn, 6Li+208Pb and 7Li+24Mg have been calculated. The inclusion of the breakup channel enhances the calculations of the fusion cross-section markedly below the Coulomb barrier and hindrance above the Coulomb barrier in comparison to the experimental data. The semi-classical calculations agree reasonably with the full quantum mechanical treatment and they were able to reproduce the experimental data in details for the total fusion reaction cross-section σfus and the fusion barrier distribution Dfus.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"218"},"PeriodicalIF":0.0,"publicationDate":"2017-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44156462","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 : 2017-11-21DOI: 10.1504/IJNEST.2017.10009089
Xuejing Li
An electromagnetic flow meter (EMFM) with non-insulation pipe wall that may be used in the Fast Reactor Test Facility (FARET) has been designed and partially tested. The internal pipe wall of EMFMs must be non-conductive to prevent generated electromotive force from short circuiting. Usually the inside of metallic pipes is lined with insulating material. The lining limits the applicable temperature range of measured fluid and also its reliability. A new structure is proposed, in which the insulating liner is eliminated and metallic pipe instead of non-insulation material. Also a servo system is applied. Therefore, the output signal is exactly the same as that of conventional EMFMs. In this paper, an analytical method based on conducting wall boundary conditions and experimental results is described.
{"title":"Electromagnetic flow meter with non-insulation pipe wall for liquid sodium in nuclear reactors","authors":"Xuejing Li","doi":"10.1504/IJNEST.2017.10009089","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10009089","url":null,"abstract":"An electromagnetic flow meter (EMFM) with non-insulation pipe wall that may be used in the Fast Reactor Test Facility (FARET) has been designed and partially tested. The internal pipe wall of EMFMs must be non-conductive to prevent generated electromotive force from short circuiting. Usually the inside of metallic pipes is lined with insulating material. The lining limits the applicable temperature range of measured fluid and also its reliability. A new structure is proposed, in which the insulating liner is eliminated and metallic pipe instead of non-insulation material. Also a servo system is applied. Therefore, the output signal is exactly the same as that of conventional EMFMs. In this paper, an analytical method based on conducting wall boundary conditions and experimental results is described.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"251"},"PeriodicalIF":0.0,"publicationDate":"2017-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43720766","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 : 2017-08-16DOI: 10.1504/IJNEST.2017.10006735
R. Chambon, A. Hébert, J. Taforeau
In order to better optimise the fuel energy efficiency and perform safety analyses in PWRs, the fuel power distribution has to be known as accurately as possible, ideally in each pin. However, this level of detail is lost when core calculations are performed with cross-sections homogenised over the fuel assembly. The pin power reconstruction (PPR) method can be used to get back this level of detail as accurately as possible in a small additional computing time frame compared to pin-by-pin full-core calculations. The DRAGON5 lattice code and the PARCS core code were recently interfaced. For this study, all the missing parts to be able to perform PPR were introduced in the newly developed system DRAGON5/PARCS. A major component was to set the methodology to compute the corner and assembly discontinuity factors in DRAGON5. Verification tests were performed on 12 configurations of 3x3 clusters where simulations in transport theory and in diffusion theory followed by pin-power reconstruction were compared.
{"title":"Pin power reconstruction: a new capability in the DRAGON5-PARCS neutronic system","authors":"R. Chambon, A. Hébert, J. Taforeau","doi":"10.1504/IJNEST.2017.10006735","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10006735","url":null,"abstract":"In order to better optimise the fuel energy efficiency and perform safety analyses in PWRs, the fuel power distribution has to be known as accurately as possible, ideally in each pin. However, this level of detail is lost when core calculations are performed with cross-sections homogenised over the fuel assembly. The pin power reconstruction (PPR) method can be used to get back this level of detail as accurately as possible in a small additional computing time frame compared to pin-by-pin full-core calculations. The DRAGON5 lattice code and the PARCS core code were recently interfaced. For this study, all the missing parts to be able to perform PPR were introduced in the newly developed system DRAGON5/PARCS. A major component was to set the methodology to compute the corner and assembly discontinuity factors in DRAGON5. Verification tests were performed on 12 configurations of 3x3 clusters where simulations in transport theory and in diffusion theory followed by pin-power reconstruction were compared.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"99"},"PeriodicalIF":0.0,"publicationDate":"2017-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47880769","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 : 2017-08-16DOI: 10.1504/IJNEST.2017.10006741
P. Najafi, S. Talebi
Safe operation and maximum burning of fuel material are two issues that are recently taken into consideration in nuclear fuel rod fabrication industry. Any failure in nuclear fuel and cladding, such as local melting or cracking, may cause release of radioactive fission fragments to the reactor coolant, which is an undesirable issue from the reactor safety point of view. Hence, one of the most important issues in nuclear industry is the preservation of fuel rod integrity during its lifetime. The performance of the irradiated fuel inside the reactor core is affected by various complex phenomena. The main objective of the present paper is to develop valid physical models and an accurate numerical method to study of Water-Water Energetic Reactor (VVER) fuel rod performance. The obtained model can be applied to simulate fuel performance during its lifetime. The correlations used in physical models are chosen in such a way that main parameters, such as gap pressure and fuel centre temperature, can be estimated accurately. The obtained results are in acceptable agreement with available data.
{"title":"Development of a method for analysis of thermal performance of VVER fuel","authors":"P. Najafi, S. Talebi","doi":"10.1504/IJNEST.2017.10006741","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10006741","url":null,"abstract":"Safe operation and maximum burning of fuel material are two issues that are recently taken into consideration in nuclear fuel rod fabrication industry. Any failure in nuclear fuel and cladding, such as local melting or cracking, may cause release of radioactive fission fragments to the reactor coolant, which is an undesirable issue from the reactor safety point of view. Hence, one of the most important issues in nuclear industry is the preservation of fuel rod integrity during its lifetime. The performance of the irradiated fuel inside the reactor core is affected by various complex phenomena. The main objective of the present paper is to develop valid physical models and an accurate numerical method to study of Water-Water Energetic Reactor (VVER) fuel rod performance. The obtained model can be applied to simulate fuel performance during its lifetime. The correlations used in physical models are chosen in such a way that main parameters, such as gap pressure and fuel centre temperature, can be estimated accurately. The obtained results are in acceptable agreement with available data.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"120"},"PeriodicalIF":0.0,"publicationDate":"2017-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42060024","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 : 2017-08-16DOI: 10.1504/IJNEST.2017.10006743
S. Kenawy, E. Elmaghraby
In the present work, we describe the use of a multi-gamma lines source to determine the linear attenuation coefficient of ceramic material. High precision germanium detector is used to assess gamma-ray transmission. Geometry is adopted to enhance detection of 210Pb gamma-ray line at 46.2 keV. Cascade summing was followed up. The effective density of micro-porous Al2O3-SiC-ZrO2 ceramics was determined by comparing the measured linear attenuation coefficients at different energies with corresponding values computed by XCOM database. Five different composing ratios are investigated, for the three mixed materials. The results illustrate the applicability of high-resolution gamma-ray attenuation in the determination of effective density of ceramics and suitability of Al2O3-SiC composite as a candidate alternative for fuel cladding.
{"title":"Experimental determination of effective density of Al2O3-SiC-ZrO2 ceramics porous phase using gamma-ray attenuation","authors":"S. Kenawy, E. Elmaghraby","doi":"10.1504/IJNEST.2017.10006743","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10006743","url":null,"abstract":"In the present work, we describe the use of a multi-gamma lines source to determine the linear attenuation coefficient of ceramic material. High precision germanium detector is used to assess gamma-ray transmission. Geometry is adopted to enhance detection of 210Pb gamma-ray line at 46.2 keV. Cascade summing was followed up. The effective density of micro-porous Al2O3-SiC-ZrO2 ceramics was determined by comparing the measured linear attenuation coefficients at different energies with corresponding values computed by XCOM database. Five different composing ratios are investigated, for the three mixed materials. The results illustrate the applicability of high-resolution gamma-ray attenuation in the determination of effective density of ceramics and suitability of Al2O3-SiC composite as a candidate alternative for fuel cladding.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"163"},"PeriodicalIF":0.0,"publicationDate":"2017-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49665866","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 : 2017-08-16DOI: 10.1504/IJNEST.2017.10006744
S. M. Shauddin, M. S. Mahmood, M.J.H. Khan
IAEA benchmark problem for VVER-1000 reactor fresh fuel assemblies has been evaluated through Monte Carlo (MC) simulation. Infinite multiplication factor (k∞) values are calculated for different hexagonal fuel assemblies with 2 wt%, 3 wt%, 3.22 wt% and 3.3 wt% enriched U-235. MC simulations have been performed with the computer code MCNP at different temperatures 120°C, 278°C, 280°C, 302°C and 305°C for three different conditions: (i) 0 ppm soluble boron concentration, (ii) 1000 ppm soluble boron concentration, and (iii) full insertion of control rod clusters (absorber material, B4C) without boron in the moderator. The results are in good agreement with the deterministic calculations as reported in IAEA-TECDOC-847.
{"title":"Criticality benchmark of VVER-1000 fresh fuel assembly with MCNP","authors":"S. M. Shauddin, M. S. Mahmood, M.J.H. Khan","doi":"10.1504/IJNEST.2017.10006744","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10006744","url":null,"abstract":"IAEA benchmark problem for VVER-1000 reactor fresh fuel assemblies has been evaluated through Monte Carlo (MC) simulation. Infinite multiplication factor (k∞) values are calculated for different hexagonal fuel assemblies with 2 wt%, 3 wt%, 3.22 wt% and 3.3 wt% enriched U-235. MC simulations have been performed with the computer code MCNP at different temperatures 120°C, 278°C, 280°C, 302°C and 305°C for three different conditions: (i) 0 ppm soluble boron concentration, (ii) 1000 ppm soluble boron concentration, and (iii) full insertion of control rod clusters (absorber material, B4C) without boron in the moderator. The results are in good agreement with the deterministic calculations as reported in IAEA-TECDOC-847.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"175"},"PeriodicalIF":0.0,"publicationDate":"2017-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44979041","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 : 2017-08-16DOI: 10.1504/IJNEST.2017.10006742
S. Meireles, A. Z. Mesquita, M. Q. Antolin, D. Campolina, D. A. Palma, M. A. Menezes
The IPR-R1 TRIGA Mark I research reactor is located at the Nuclear Technology Development Centre (CDTN), in Belo Horizonte, Brazil. It is operating for more than 50 years and was successfully simulated before. However, new techniques and methods used in nuclear reactors analysis make a further simulation inevitable. In this manuscript, the computational model of an initial core of the IPR-R1 TRIGA reactor was developed employing two different Monte Carlo codes, MCNPX and Serpent 2, to simulate the neutronics behaviour. A new model is suggested, more complete, to improve the simulations results making the model more close the experimental data. This work explores how changes could be inserted in order to make the model closer to reality and if such participation would be noticeable in both codes used. The neutronic parameters obtained from these simulations performed in Serpent 2 are compared to MCNPX simulation results at the same conditions, and the results are compared with previous experimental data.
IPR-R1 TRIGA Mark I研究反应堆位于巴西贝洛奥里藏特的核技术开发中心（CDTN）。它已经运行了50多年，以前曾成功模拟过。然而，在核反应堆分析中使用的新技术和方法使得进一步的模拟不可避免。在本文中，IPR-R1 TRIGA反应堆初始堆芯的计算模型是使用两种不同的蒙特卡罗代码MCNPX和Serpent 2开发的，以模拟中子学行为。提出了一个更完整的新模型，以改进模拟结果，使模型更接近实验数据。这项工作探讨了如何插入更改，以使模型更接近现实，以及这种参与是否在使用的两种代码中都很明显。将从Serpent 2中进行的这些模拟中获得的中子参数与相同条件下的MCNPX模拟结果进行比较，并将结果与先前的实验数据进行比较。
{"title":"Influence of geometry in TRIGA reactor criticality calculation and reactivity determination using Serpent 2 and MCNPX codes","authors":"S. Meireles, A. Z. Mesquita, M. Q. Antolin, D. Campolina, D. A. Palma, M. A. Menezes","doi":"10.1504/IJNEST.2017.10006742","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10006742","url":null,"abstract":"The IPR-R1 TRIGA Mark I research reactor is located at the Nuclear Technology Development Centre (CDTN), in Belo Horizonte, Brazil. It is operating for more than 50 years and was successfully simulated before. However, new techniques and methods used in nuclear reactors analysis make a further simulation inevitable. In this manuscript, the computational model of an initial core of the IPR-R1 TRIGA reactor was developed employing two different Monte Carlo codes, MCNPX and Serpent 2, to simulate the neutronics behaviour. A new model is suggested, more complete, to improve the simulations results making the model more close the experimental data. This work explores how changes could be inserted in order to make the model closer to reality and if such participation would be noticeable in both codes used. The neutronic parameters obtained from these simulations performed in Serpent 2 are compared to MCNPX simulation results at the same conditions, and the results are compared with previous experimental data.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"146"},"PeriodicalIF":0.0,"publicationDate":"2017-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49662551","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 : 2017-08-16DOI: 10.1504/IJNEST.2017.10006745
G. Espinosa-Paredes, J. François, H. Sánchez-Mora, A. Pérez-Valseca, C. Martin-Del-Campo
The aim of this paper is to make a comparative study of two concepts of Lead-Cooled Fast Reactor (LFR) fuel assemblies, from a point of view of the thermofluids performance. The sub-channel analysis approach was applied to determine the temperature distribution in the fuel, in the cladding and in the lead-coolant. The mathematical model is fully transient and takes into account the heat transfer in an annular fuel pellet design. The thermofluid is modelled with a mass, energy and momentum balance with thermal expansion effects. The neutronic processes are modelled with point kinetic equations for power generation with feedback fuel temperature and expansion effects. The numerical experiments consider steady-state and transient behaviours. The numerical comparison shows that a hexagonal assembly is an option to compact the size of the LFR core design. This option leads to higher temperature in the fuel and the cladding than in the case of a rectangular assembly design. Results show the LFR with square array is more sensitive to power changes than the hexagonal array at the same nominal power and with the same transient conditions.
{"title":"Study on the temperature distributions in fuel assemblies of lead-cooled fast reactors","authors":"G. Espinosa-Paredes, J. François, H. Sánchez-Mora, A. Pérez-Valseca, C. Martin-Del-Campo","doi":"10.1504/IJNEST.2017.10006745","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10006745","url":null,"abstract":"The aim of this paper is to make a comparative study of two concepts of Lead-Cooled Fast Reactor (LFR) fuel assemblies, from a point of view of the thermofluids performance. The sub-channel analysis approach was applied to determine the temperature distribution in the fuel, in the cladding and in the lead-coolant. The mathematical model is fully transient and takes into account the heat transfer in an annular fuel pellet design. The thermofluid is modelled with a mass, energy and momentum balance with thermal expansion effects. The neutronic processes are modelled with point kinetic equations for power generation with feedback fuel temperature and expansion effects. The numerical experiments consider steady-state and transient behaviours. The numerical comparison shows that a hexagonal assembly is an option to compact the size of the LFR core design. This option leads to higher temperature in the fuel and the cladding than in the case of a rectangular assembly design. Results show the LFR with square array is more sensitive to power changes than the hexagonal array at the same nominal power and with the same transient conditions.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"183"},"PeriodicalIF":0.0,"publicationDate":"2017-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41371280","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 : 2017-07-12DOI: 10.1504/IJNEST.2017.10005997
G. Ajeesh, S. Bhowmik, V. Sivakumar, L. Varshney, Virendra Kumar, M. Abraham, J. Epaarachchi
This research highlights the effect of radiation, chemical and thermal environments on mechanical and thermal properties of polyetherimide (PEI) composites. The tests are conducted on specimens made from PEI and PEI reinforced with modified Carbon Nano Fibre (CNF). The specimens are subjected to gamma radiation doses of 5 MGy, which is equivalent to the cumulative dose of radiation from spent nuclear fuel until the end of complete radioactivity. The exposed samples are further subjected to highly corrosive and thermal environments. Studies under transmission electron microscopy reveal that there is a uniform dispersion of modified CNF in PEI. Differential Scanning Calorimetry (DSC) and Thermo Gravimetric Analysis (TGA) indicate that there are no significant changes in thermal properties of PEI and PEI composite when exposed to aggressive environments. It is observed that there is a marginal loss in the tensile strength of polymeric samples when exposed to gamma radiation and thermal environments. PEI samples when subjected to alkaline corrosive environments show significant loss in the tensile strength. There is a significant decrease in the molecular weight of PEI under alkaline corrosive environments as seen from Gel Permeable Chromatography (GPC).
{"title":"Effects of high energy radiation and thermo-chemical environments on polyetherimide composites: futuristic approach to nuclear waste storage","authors":"G. Ajeesh, S. Bhowmik, V. Sivakumar, L. Varshney, Virendra Kumar, M. Abraham, J. Epaarachchi","doi":"10.1504/IJNEST.2017.10005997","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10005997","url":null,"abstract":"This research highlights the effect of radiation, chemical and thermal environments on mechanical and thermal properties of polyetherimide (PEI) composites. The tests are conducted on specimens made from PEI and PEI reinforced with modified Carbon Nano Fibre (CNF). The specimens are subjected to gamma radiation doses of 5 MGy, which is equivalent to the cumulative dose of radiation from spent nuclear fuel until the end of complete radioactivity. The exposed samples are further subjected to highly corrosive and thermal environments. Studies under transmission electron microscopy reveal that there is a uniform dispersion of modified CNF in PEI. Differential Scanning Calorimetry (DSC) and Thermo Gravimetric Analysis (TGA) indicate that there are no significant changes in thermal properties of PEI and PEI composite when exposed to aggressive environments. It is observed that there is a marginal loss in the tensile strength of polymeric samples when exposed to gamma radiation and thermal environments. PEI samples when subjected to alkaline corrosive environments show significant loss in the tensile strength. There is a significant decrease in the molecular weight of PEI under alkaline corrosive environments as seen from Gel Permeable Chromatography (GPC).","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"80"},"PeriodicalIF":0.0,"publicationDate":"2017-07-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43294791","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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