Pub Date : 2018-08-27DOI: 10.1504/IJNEST.2018.10015409
A. Agung
Safety analysis of a PWR fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA condition. The ATF being used is uranium silicide (U3Si2) and FCMF (Fully Ceramic Microencapsulated Fuel) with silicon carbide (SiC) and FeCrAl alloy as a cladding material. The objective of this research is to obtain dynamic characteristics of ATF-fuelled PWR at LB-LOCA condition. RELAP5-3D system code was used to model the reactor and simulate the transient. A safe shutdown of the reactor was assumed after a depressurisation following a double-ended guillotine breach in the main pipe. The results of simulations show that during LB-LOCA with partially functioning ECCS, the transient PCTs were far below the maximum allowable limit. The use of ATF could decrease the maximum transient PCT. It is shown that U3Si2 fuel with FeCrAl cladding has the minimum PCT transient and the shortest quench time to steady state condition after transient initiation.
{"title":"Prediction of peak cladding temperature in a three-loop pressurised water reactor with accident- tolerant fuel during loss-of-coolant accident","authors":"A. Agung","doi":"10.1504/IJNEST.2018.10015409","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10015409","url":null,"abstract":"Safety analysis of a PWR fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA condition. The ATF being used is uranium silicide (U3Si2) and FCMF (Fully Ceramic Microencapsulated Fuel) with silicon carbide (SiC) and FeCrAl alloy as a cladding material. The objective of this research is to obtain dynamic characteristics of ATF-fuelled PWR at LB-LOCA condition. RELAP5-3D system code was used to model the reactor and simulate the transient. A safe shutdown of the reactor was assumed after a depressurisation following a double-ended guillotine breach in the main pipe. The results of simulations show that during LB-LOCA with partially functioning ECCS, the transient PCTs were far below the maximum allowable limit. The use of ATF could decrease the maximum transient PCT. It is shown that U3Si2 fuel with FeCrAl cladding has the minimum PCT transient and the shortest quench time to steady state condition after transient initiation.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"196"},"PeriodicalIF":0.0,"publicationDate":"2018-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43147648","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 : 2018-06-26DOI: 10.1504/IJNEST.2018.10013859
S. Krovvidi, R. Punniyamoorthy, B. Sreedhar, S. Chandramouli, G. Padmakumar, S. Raghupathy, P. Selvaraj
Track roller type double row deep groove ball bearings made of SS-440C material are used in inclined fuel transfer machine (IFTM) of PFBR. The bearings are unlubricated, need to operate under sodium and subsequently in cover gas at 50°C. The number of rotations of the bearings during 40 years of the reactor operation is 1.15 million. This paper presents the experiment carried out to examine the performance of the bearings used in IFTM under simulated conditions as in reactor. The bearing was initially tested in air, argon and in sodium at 200°C. It is observed that the frictional torque value of the sodium wetted bearing in cover gas at 50°C is increasing due to solidified sodium sticking on the bearing. Wear in the bearing is significant after 1,065,000 rotations. This paper presents the scheme of testing of the bearing, details of the test facility, test results and discussion. The frequency of replacement of the bearings in IFTM of PFBR is established based on the experiment. This experiment gave better insight to use bearings for under sodium applications.
{"title":"Testing in sodium and qualification of the bearings used in inclined fuel transfer machine of prototype fast breeder reactor","authors":"S. Krovvidi, R. Punniyamoorthy, B. Sreedhar, S. Chandramouli, G. Padmakumar, S. Raghupathy, P. Selvaraj","doi":"10.1504/IJNEST.2018.10013859","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10013859","url":null,"abstract":"Track roller type double row deep groove ball bearings made of SS-440C material are used in inclined fuel transfer machine (IFTM) of PFBR. The bearings are unlubricated, need to operate under sodium and subsequently in cover gas at 50°C. The number of rotations of the bearings during 40 years of the reactor operation is 1.15 million. This paper presents the experiment carried out to examine the performance of the bearings used in IFTM under simulated conditions as in reactor. The bearing was initially tested in air, argon and in sodium at 200°C. It is observed that the frictional torque value of the sodium wetted bearing in cover gas at 50°C is increasing due to solidified sodium sticking on the bearing. Wear in the bearing is significant after 1,065,000 rotations. This paper presents the scheme of testing of the bearing, details of the test facility, test results and discussion. The frequency of replacement of the bearings in IFTM of PFBR is established based on the experiment. This experiment gave better insight to use bearings for under sodium applications.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"1"},"PeriodicalIF":0.0,"publicationDate":"2018-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46478850","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 : 2018-06-26DOI: 10.1504/IJNEST.2018.10013863
W. Menezes, H. A. Filho, R. Barros
A spectral nodal method for energy multi-group X, Y-geometry, discrete ordinates (SN) problems in non-multiplying medium is developed. This analytical coarse-mesh method is referred to as the multi-group spectral Green's function - constant nodal (SGF-CN) method. The SGF-CN method uses the multi-group SGF method for numerically solving the one-dimensional transverse-integrated SN nodal equations with constant approximations for the transverse leakage terms. As the energy-group transfer scattering source terms are treated analytically in the offered method, the only approximations occur in the group transverse leakage terms. Numerical results are given to illustrate the method's accuracy for coarse-mesh calculations.
{"title":"An analytical nodal method for energy multi-group discrete ordinates transport calculations in two-dimensional rectangular geometry","authors":"W. Menezes, H. A. Filho, R. Barros","doi":"10.1504/IJNEST.2018.10013863","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10013863","url":null,"abstract":"A spectral nodal method for energy multi-group X, Y-geometry, discrete ordinates (SN) problems in non-multiplying medium is developed. This analytical coarse-mesh method is referred to as the multi-group spectral Green's function - constant nodal (SGF-CN) method. The SGF-CN method uses the multi-group SGF method for numerically solving the one-dimensional transverse-integrated SN nodal equations with constant approximations for the transverse leakage terms. As the energy-group transfer scattering source terms are treated analytically in the offered method, the only approximations occur in the group transverse leakage terms. Numerical results are given to illustrate the method's accuracy for coarse-mesh calculations.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"66"},"PeriodicalIF":0.0,"publicationDate":"2018-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44697638","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 : 2018-06-26DOI: 10.1504/IJNEST.2018.10013864
K. S. Jassim, Shamil R. Sahib
Shell model calculations were performed to study the energy levels for the 25,26Mg and 27Al nuclei by employing the sdpfnow effective interaction with the large-scale sdpf model space by using the shell model code [email protected] for Windows. The electron scattering form factors for 19F nucleus also have been studied with and without effective charge on the sdpf-model space and Tassie model. The Harmonic Oscillator and Skyrme potentials have been used to calculate the wave functions of radial single-particle matrix elements. The level schemes are compared with the experimental data. Coulomb and magnetic form factors in the present work include the transitions from ground state (1/2+ 1/2) to the (7/2+ 1/2), (9/2+ 1/2), (3/2− 1/2), and (11/2− 1/2) states in 19F. Good agreements were obtained for all nuclei under study for energy levels and form factors comparing with the available experimental data.
{"title":"Large-scale shell model calculations of the 25,26Mg, 27Al and 19F nucleus","authors":"K. S. Jassim, Shamil R. Sahib","doi":"10.1504/IJNEST.2018.10013864","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10013864","url":null,"abstract":"Shell model calculations were performed to study the energy levels for the 25,26Mg and 27Al nuclei by employing the sdpfnow effective interaction with the large-scale sdpf model space by using the shell model code [email protected] for Windows. The electron scattering form factors for 19F nucleus also have been studied with and without effective charge on the sdpf-model space and Tassie model. The Harmonic Oscillator and Skyrme potentials have been used to calculate the wave functions of radial single-particle matrix elements. The level schemes are compared with the experimental data. Coulomb and magnetic form factors in the present work include the transitions from ground state (1/2+ 1/2) to the (7/2+ 1/2), (9/2+ 1/2), (3/2− 1/2), and (11/2− 1/2) states in 19F. Good agreements were obtained for all nuclei under study for energy levels and form factors comparing with the available experimental data.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"81"},"PeriodicalIF":0.0,"publicationDate":"2018-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46774907","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 : 2018-06-26DOI: 10.1504/IJNEST.2018.10013860
A. Hossain, A. Z. M. Salahuddin, M. Akbar
Threat modelling and assessments are the processes of gathering, organising and analysing existing or potential threats and deemed to have the capabilities to commit a malicious act. Potential adversaries who may attempt unauthorised removal of nuclear materials (NM) or other radioactive materials (RM) for which a physical protection system (PPS) is designed, and therefore must be assessed and prevented. In case of an undesired condition, the authorities have to carry out analytic activities to detect risky circumstances. Hence, in spite of the various methods for threat modelling, it is essential to systematically analyse these threats. Therefore, in this paper, a threat modelling technique by using fuzzy logic based intelligent approach is designed. The technique involves linking the relationship between input parameters of capability, intent, material and vulnerability and output parameter of threat level for nuclear and radioactive materials and their adaptation for the early forecast of irregular behaviour. For inputs overall capabilities 70%, overall likelihood 60%, and impact 60%, the output threat level is estimated as 76.5% for the domestic group deploying an RDD at an annual celebration. Results obtained from the study show the good performance of the developed model as compared to results considering single fuzzy inference system (SFIS).
{"title":"Threat modelling on nuclear and radioactive materials based on intelligent approach","authors":"A. Hossain, A. Z. M. Salahuddin, M. Akbar","doi":"10.1504/IJNEST.2018.10013860","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10013860","url":null,"abstract":"Threat modelling and assessments are the processes of gathering, organising and analysing existing or potential threats and deemed to have the capabilities to commit a malicious act. Potential adversaries who may attempt unauthorised removal of nuclear materials (NM) or other radioactive materials (RM) for which a physical protection system (PPS) is designed, and therefore must be assessed and prevented. In case of an undesired condition, the authorities have to carry out analytic activities to detect risky circumstances. Hence, in spite of the various methods for threat modelling, it is essential to systematically analyse these threats. Therefore, in this paper, a threat modelling technique by using fuzzy logic based intelligent approach is designed. The technique involves linking the relationship between input parameters of capability, intent, material and vulnerability and output parameter of threat level for nuclear and radioactive materials and their adaptation for the early forecast of irregular behaviour. For inputs overall capabilities 70%, overall likelihood 60%, and impact 60%, the output threat level is estimated as 76.5% for the domestic group deploying an RDD at an annual celebration. Results obtained from the study show the good performance of the developed model as compared to results considering single fuzzy inference system (SFIS).","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"19"},"PeriodicalIF":0.0,"publicationDate":"2018-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44363238","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 : 2018-06-26DOI: 10.1504/IJNEST.2018.10013861
M. M. El-din, R. Mahmoud, I. Eid, M. R. E. El-din, R. Rizk
The objective of this paper is to assess the radiation exposure resulting from radioactive patients injected with different activities of 2-[18F] fluoro-2-deoxy-D-glucose (18F-FDG) in Positron Emission Tomography/Computed Tomography (PET/CT) units. This objective is fulfilled by measuring the dose rates practically inside and outside PET/CT rooms around radioactive patients using a calibrated survey meter. Afterwards, the dose rates are estimated mathematically using Monte Carlo simulation model. The results show that the dose rates on patient's body surface decrease greatly with distance and it is recommended for PET/CT staff to stand at distances more than 1.5 m from radioactive patients if possible during direct contact. Also, it is found that the shielding thickness in the selected room dimensions is adequate and effective for the γ-radiation arising from radioactive patients. The practically measured dose rates around radioactive patients are quite similar to mathematically predicted results and slight differences may be attributed to the difference between the estimated 18F biological half life time and real biological half life time due different biological uptake or excretion time from one patient to another.
{"title":"Radiation dose rate assessment around patients in PET/CT units","authors":"M. M. El-din, R. Mahmoud, I. Eid, M. R. E. El-din, R. Rizk","doi":"10.1504/IJNEST.2018.10013861","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10013861","url":null,"abstract":"The objective of this paper is to assess the radiation exposure resulting from radioactive patients injected with different activities of 2-[18F] fluoro-2-deoxy-D-glucose (18F-FDG) in Positron Emission Tomography/Computed Tomography (PET/CT) units. This objective is fulfilled by measuring the dose rates practically inside and outside PET/CT rooms around radioactive patients using a calibrated survey meter. Afterwards, the dose rates are estimated mathematically using Monte Carlo simulation model. The results show that the dose rates on patient's body surface decrease greatly with distance and it is recommended for PET/CT staff to stand at distances more than 1.5 m from radioactive patients if possible during direct contact. Also, it is found that the shielding thickness in the selected room dimensions is adequate and effective for the γ-radiation arising from radioactive patients. The practically measured dose rates around radioactive patients are quite similar to mathematically predicted results and slight differences may be attributed to the difference between the estimated 18F biological half life time and real biological half life time due different biological uptake or excretion time from one patient to another.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"32"},"PeriodicalIF":0.0,"publicationDate":"2018-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42756199","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 : 2018-06-26DOI: 10.1504/IJNEST.2018.10013862
K. Chen, K. Ting, A. Nguyen, Li Wang, Y. Li, T. Kuo
Semi-elliptical underclad cracks resulting from the fabrication process of a reactor pressure vessel (RPV) were able to be detected by non-destructive testing method. Meanwhile, after long-term operation under severe conditions, such as high temperature, high pressure, and irradiation, the RPV becomes brittle and susceptible to damage, especially when subjected to pressurised thermal shocks (PTS). Therefore, the probabilistic fracture mechanics (PFM) analysis of RPV with the crack should be applied to evaluate the operation safety. To the best of the authors' knowledge, few studies or computer codes have applied PFM analysis for such cracks. Therefore, this study conducts PFM analysis for cracks by modifying the calculation procedure of FAVOR 12.1 computer code. The results show that during the lifetime of a nuclear power plant, such cracks will not threaten the RPV's safety. Additionally, three methods were proposed to improve FAVOR 12.1's ability to perform PFM analysis for axial through-clad cracking.
{"title":"Probabilistic fracture mechanics analysis of reactor pressure vessel with underclad and through-clad cracks under pressurised thermal shock transient","authors":"K. Chen, K. Ting, A. Nguyen, Li Wang, Y. Li, T. Kuo","doi":"10.1504/IJNEST.2018.10013862","DOIUrl":"https://doi.org/10.1504/IJNEST.2018.10013862","url":null,"abstract":"Semi-elliptical underclad cracks resulting from the fabrication process of a reactor pressure vessel (RPV) were able to be detected by non-destructive testing method. Meanwhile, after long-term operation under severe conditions, such as high temperature, high pressure, and irradiation, the RPV becomes brittle and susceptible to damage, especially when subjected to pressurised thermal shocks (PTS). Therefore, the probabilistic fracture mechanics (PFM) analysis of RPV with the crack should be applied to evaluate the operation safety. To the best of the authors' knowledge, few studies or computer codes have applied PFM analysis for such cracks. Therefore, this study conducts PFM analysis for cracks by modifying the calculation procedure of FAVOR 12.1 computer code. The results show that during the lifetime of a nuclear power plant, such cracks will not threaten the RPV's safety. Additionally, three methods were proposed to improve FAVOR 12.1's ability to perform PFM analysis for axial through-clad cracking.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"12 1","pages":"45"},"PeriodicalIF":0.0,"publicationDate":"2018-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47415170","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.2017.10009093
A. R. Kaladgi, A. Samee, M. Ramis
Liquid metals, such as sodium (Na), lead (Pb), and lead-bismuth (Pb-Bi) eutectic (e), are considered as potential coolants for the fast spectrum nuclear reactors of the next generation. So the main objective of this paper is to study the heat transfer and fluid flow characteristics of liquid metal coolants flowing over a nuclear fuel element having uniform volumetric energy generation. Stream function vorticity formulation method was used to solve the full Navier Stokes equations governing the flow. The energy equation was solved using central finite difference method. For the two-dimensional steady state heat conduction and stream-function equation, the discretisation was done in the form suitable to solve using 'line-by-line Gauss-Seidel' solution technique whereas the discretisation of vorticity transport and energy equations was done using Alternating Direction Implicit (ADI) scheme. After discretisation the systems of algebraic equations were solved using 'Thomas algorithm'. The complete work was done by writing a well-validated indigenous computer code using C-language. The parameters considered for the study were: aspect ratio of fuel element, Ar, conduction-convection parameter Ncc, total energy generation parameter Qt, and flow Reynolds number ReH. The results obtained can be used to minimise the maximum temperature in the fuel element (hot spots) and prevent its melting.
{"title":"Influence of Prandtl number on heat transfer of a flat vertical plate","authors":"A. R. Kaladgi, A. Samee, M. Ramis","doi":"10.1504/IJNEST.2017.10009093","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10009093","url":null,"abstract":"Liquid metals, such as sodium (Na), lead (Pb), and lead-bismuth (Pb-Bi) eutectic (e), are considered as potential coolants for the fast spectrum nuclear reactors of the next generation. So the main objective of this paper is to study the heat transfer and fluid flow characteristics of liquid metal coolants flowing over a nuclear fuel element having uniform volumetric energy generation. Stream function vorticity formulation method was used to solve the full Navier Stokes equations governing the flow. The energy equation was solved using central finite difference method. For the two-dimensional steady state heat conduction and stream-function equation, the discretisation was done in the form suitable to solve using 'line-by-line Gauss-Seidel' solution technique whereas the discretisation of vorticity transport and energy equations was done using Alternating Direction Implicit (ADI) scheme. After discretisation the systems of algebraic equations were solved using 'Thomas algorithm'. The complete work was done by writing a well-validated indigenous computer code using C-language. The parameters considered for the study were: aspect ratio of fuel element, Ar, conduction-convection parameter Ncc, total energy generation parameter Qt, and flow Reynolds number ReH. The results obtained can be used to minimise the maximum temperature in the fuel element (hot spots) and prevent its melting.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"272"},"PeriodicalIF":0.0,"publicationDate":"2017-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47509117","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.2017.10009091
Rashida Yasmeen, M. S. Mahmood
The kinetic parameters govern the transient behaviour of a nuclear reactor. Estimation of these parameters has great importance for the safe design and operation of a nuclear reactor. In order to understand the kinetic behaviour of TRIGA core, the influence of graphite dummy elements, absorber materials and beam ports on kinetic parameters has been studied. To do so, the effective delayed neutron fraction (βeff), prompt removal lifetime (e) and mean neutron generation time (Λ) have been calculated for an infinite system of LEU fuel cells. The investigation has been extended to four different cases of TRIGA reactor core. Monte Carlo simulation has been carried out to compute the kinetic parameters by Monte Carlo code MCNP5. The βeff value remains unchanged but the e and Λ values are found to be sensitive to the local perturbations of core components.
{"title":"Understanding kinetic behaviour of TRIGA reactor core","authors":"Rashida Yasmeen, M. S. Mahmood","doi":"10.1504/IJNEST.2017.10009091","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10009091","url":null,"abstract":"The kinetic parameters govern the transient behaviour of a nuclear reactor. Estimation of these parameters has great importance for the safe design and operation of a nuclear reactor. In order to understand the kinetic behaviour of TRIGA core, the influence of graphite dummy elements, absorber materials and beam ports on kinetic parameters has been studied. To do so, the effective delayed neutron fraction (βeff), prompt removal lifetime (e) and mean neutron generation time (Λ) have been calculated for an infinite system of LEU fuel cells. The investigation has been extended to four different cases of TRIGA reactor core. Monte Carlo simulation has been carried out to compute the kinetic parameters by Monte Carlo code MCNP5. The βeff value remains unchanged but the e and Λ values are found to be sensitive to the local perturbations of core components.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"265"},"PeriodicalIF":0.0,"publicationDate":"2017-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46577663","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.2017.10009088
L. Castro, R. Alfonso, Carlos R. García, J. Rosales, D. S. Dominguez
The High-Performance Light Water Reactor (HPLWR) is the European Supercritical Water-cooled Reactor design. In this paper, a thermal-hydraulic study of the HPLWR fuel assembly using CFD codes was carried out. The capability of the Reynolds Stress model of Speziale (SSG) and the k-ω Shear Stress Transport model (SST) for predicting the supercritical water heat transfer was evaluated. The axial temperature distributions of the fuel, cladding, coolant and moderator in the fuel assembly were obtained. Numerical results of the fuel temperature profiles were compared with that obtained by Waata (2006) and a good agreement was achieved. The cladding surface temperature profiles calculated with SSG and SST turbulence models are below the prescribed limits; however, hot spots in one sub-channel were found. The difference in the average thermal-hydraulic properties of the supercritical water calculated with SSG and SST was negligible. The fuel and cladding surface temperatures are higher when using the SST model.
{"title":"CFD analysis of thermal-hydraulic behaviour of the high performance light water reactor fuel assembly","authors":"L. Castro, R. Alfonso, Carlos R. García, J. Rosales, D. S. Dominguez","doi":"10.1504/IJNEST.2017.10009088","DOIUrl":"https://doi.org/10.1504/IJNEST.2017.10009088","url":null,"abstract":"The High-Performance Light Water Reactor (HPLWR) is the European Supercritical Water-cooled Reactor design. In this paper, a thermal-hydraulic study of the HPLWR fuel assembly using CFD codes was carried out. The capability of the Reynolds Stress model of Speziale (SSG) and the k-ω Shear Stress Transport model (SST) for predicting the supercritical water heat transfer was evaluated. The axial temperature distributions of the fuel, cladding, coolant and moderator in the fuel assembly were obtained. Numerical results of the fuel temperature profiles were compared with that obtained by Waata (2006) and a good agreement was achieved. The cladding surface temperature profiles calculated with SSG and SST turbulence models are below the prescribed limits; however, hot spots in one sub-channel were found. The difference in the average thermal-hydraulic properties of the supercritical water calculated with SSG and SST was negligible. The fuel and cladding surface temperatures are higher when using the SST model.","PeriodicalId":35144,"journal":{"name":"International Journal of Nuclear Energy Science and Technology","volume":"11 1","pages":"229"},"PeriodicalIF":0.0,"publicationDate":"2017-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49305615","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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