Dr. Md. Jahirul Haque Khan (original) (raw)

Papers by Dr. Md. Jahirul Haque Khan

International journal of science & technoledge, Jan 31, 2023

Calculation of LEU Fuel Burn-up and Core Life Time Estimation of BAEC TRIGA Research Reactor Usin... more Calculation of LEU Fuel Burn-up and Core Life Time Estimation of BAEC TRIGA Research Reactor Using 2D-TRIGLAV Code 1. Introduction The TRIGA Mark-II research reactor [1] has a maximum thermal neutron flux of 7.46x10 13 n/cm 2 /sec in the middle of the core and is light water-cooled. This graphite-reflected nuclear reactor is intended for continuous operation at a steady state power level of 3 MWt. The TRIGA reactor LEU fuel is made up of burnable poison Erbium, zirconium hydride (primary moderator), and 20 weight percent uranium enriched to 19.7% 235 U. Boron carbide (B4C) serves as the neutron absorber component of the control rods. 100 fuel elements, including five fuelled follower control rods, six control rods, one air follower control rod, 18 graphite dummy elements, one central thimble, one pneumatic transfer system irradiation endpoint, and various light waters make up the BTRR LEU core. All of these components were positioned, supported, and arranged in seven concentric hexagonal rings (A, B, C, D, E, F, and G) of a hexagonal lattice, as shown in figure 1, between the top and bottom grid plates. It was given the go-ahead to carry out a number of nuclear research projects like neutron activation analysis, thermal neutron radiography, and neutron diffraction scattering experiments, as well as to train workers and create radioisotopes for application in medicine, industry, and agriculture. The term 'subcritical' describes a system where the loss of neutrons is greater than the rate of neutron creation [2], and the neutron population gradually declines over time. Criticality is a nuclear term that relates to the balance of neutrons in the core. A system is said to be 'supercritical' when neutron production outpaces neutron loss, increasing the population of neutrons [2, 3]. When the neutron populations are stable, the production and loss of neutrons are perfectly balanced, and the nuclear system is in a critical condition [3]. By comparing the pace at which neutrons are created from fission and other sources to the rate at which they are lost through absorption, scattering, and leakage out of the nuclear reactor core, it is possible to determine the criticality of a system [4]. The neutron diffusion theory code is used to perform this analysis. The configuration of the initial core shape and neutron energy group constants for various homogenized regions of the core, along with the fission spectrum, are required. In this study, neutron group constants were obtained using the well-known 1-D neutron transport code WIMS-D/4 [5] and were utilized for the full core calculations with the TRIGLAV code [6] that is based on a 4-group timeindependent diffusion equation in two-dimensional cylindrical (r,) configuration. The neutron diffusion algorithm is

International journal of science & technoledge, Jan 31, 2023

EL 1 DE JUNIO EL ALBUM “SGTO. PEPPER’S LONELY HEARTS CLUB BAND” CUMPLIRA 50 ANOS, CONSIDERADO POR... more EL 1 DE JUNIO EL ALBUM “SGTO. PEPPER’S LONELY HEARTS CLUB BAND” CUMPLIRA 50 ANOS, CONSIDERADO POR LA REVISTA “ROLLING STONE” COMO EL MAS IMPORTANTE DE TODOS LOS TIEMPOS Y UNO TAN INFLUYENTE QUE, CON SUS 13 CANCIONES, UN SINFIN DE BANDAS APRENDIERON A TOCAR. PARA JULIO MUNOZ RUBIO, DEL CENTRO DE INVESTIGACIONES INTERDISCIPLINARIAS EN CIENCIAS Y HUMANIDADES (CEIICH), SIEMPRE TENEMOS QUE VER CON RECELO LISTADOS TAN TAJANTES, “AUNQUE EN ESTA OCASION DEBEMOS CONCEDER QUE HABLAMOS DE UNA OBRA MAESTRA DEL SIGLO XX QUE ROMPIO CON TODO LO QUE SE HABIA HECHO A LA FECHA”. LANZADO EL 1 DE JUNIO DE 1967 EN INGLATERRA –Y AL DIA SIGUIENTE EN ESTADOS UNIDOS–, EL “SARGENTO PIMIENTA” LLEGO EN EL MOMENTO PRECISO, PUES ES PRODUCTO DE UNA EPOCA EN LA QUE LA JUVENTUD TOMABA POR ASALTO LAS ESFERAS DE PODER Y LAS PERSONAS SE CUESTIONABAN TODO, RECORDO. EL CONTEXTO: VIETNAM EN GUERRA, LA GENTE SE CONGREGABA EN SAN FRANCISCO PARA DAR LUGAR AL LLAMADO VERANO DEL AMOR, EN CUBA SE CONSOLIDABA EL SOCIALISMO, LAS...

International Journal of Nuclear Energy Science and Technology

International Journal of Nuclear Energy Science and Technology

The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile a... more The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the newly generated continuous energy cross section data from JENDL-3.3 was performed against some well-known benchmark lattices using MCNP4C and the results were found to be in very good agreement with the experiment and other evaluations. For TRIGA analysis continuous energy cross section data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300K evaluations were used. Full S(α,β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH...

IEA-R1 reactor core has been modeled in COOLOD-N2 code with the aim to validate and verify the co... more IEA-R1 reactor core has been modeled in COOLOD-N2 code with the aim to validate and verify the code against the experiment data provided by International Atomic Energy Agency (IAEA) through coordinated research project (CRP). Fuel cladding temperatures at different locations of an instrumental fuel assembly (IFA) at different operating powers have been calculated under steady state condition by using the COOLOD-N2 code. The calculated temperatures are then verified against the temperatures recorded by thermo couples placed for those locations of IFA in 243 configuration of IEA-R1 reactor. Reasonable agreement has been found between the calculation and experiment for fuel clad temperature which in turn implies the validity of COOLOD-N2 code in the calculation ofcore thermal hydraulic parameters of research reactors under steady state condition.

Some parametric studies on pulsing mode for fresh core of TRIGA Mark II research reactor in AERE,... more Some parametric studies on pulsing mode for fresh core of TRIGA Mark II research reactor in AERE, Savar, have been carried out with coupled thermal-hydraulics code EUREKA-2/RR in association with neutronics code SRAC. At the beginning, role of some important parameters in pulsing like delayed neutron fraction (βeff) and reactivity insertion have been studied keeping prompt neutron life time (lp) fixed at 33.4 µ-sec. After a series of experiments, we found that the pulsing peak that is consistent with the Safety Analysis Report (SAR) is for the delayed neutron fraction (βeff) of 0.007 and reactivity insertion of 2$. Study has determined the pulsing peak of the fresh core for this particular condition to be 857.86 MW which is 852 MW according to SAR. Experiment also shows the pulsing peak increases with the increase of reactivity insertion whereas decreases with increase of delayed neutron fraction. With the utilization of the particular values of these parameters, pulsing parameters ...

Progress in Nuclear Energy, 2009

ABSTRACT The objective of this paper is to present the results of comparative study of integral p... more ABSTRACT The objective of this paper is to present the results of comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors. The nuclear data processing code NJOY'99 was deployed for the generation of the 69-group cross-section library from the basic evaluated nuclear data files JENDL-3.2 and JEF-2.2. TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project Stage-I. The inputs were the results of a detailed parametric study of the WIMS input options and also optimized for accuracy. The integral parameters (such as keff, ρ28, δ25, δ28, C∗) of five uranium-fuel thermal assemblies: TRX-1 and TRX-2 and BAPL-1, BAPL-2, and BAPL-3 were calculated with the help of WIMSD-5B code based on the generated 69-group cross-section library. The calculated results are compared with those of experiments and it is found that the obtained results between the two libraries are in good agreement with each other. Besides, the calculated integral parameters are also well consistent with the measured values, which reflect the validation of the generated 69-group cross-section library and this library thus obtained is necessary to meet up the nuclear data for neutronics calculation of TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh.

International Journal of Nuclear Energy, 2014

ABSTRACT EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the e... more ABSTRACT EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the experimental results provided by IAEA(International Atomic Energy Agency) obtained for a series of transient tests initiated by step insertion of different magnitudes of positive reactivity with varying degrees of different controlled parameters such as reactor initial power, coolant temperature and coolant flow condition. 20 out of 39 tests that fall under forced convection mode have been considered for the present simulation provided the reactor scram system is disabled. Peak power and peak clad temperature due to transient have been calculated and it was found that although peak clad temperature values agreed, the peak power values seem to underestimate the experimental values. Further study appears to be needed to identify the limitations in modeling or examining the effect of input parameters during modeling to obtain the better simulation results.

Annals of Nuclear Energy, 2013

ABSTRACT The aim of this study is to evaluate the kinetic parameters of 3 MW TRIGA Mark-II resear... more ABSTRACT The aim of this study is to evaluate the kinetic parameters of 3 MW TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety. The most important kinetic parameters of nuclear reactors are the effective delayed neutron fraction (βeff), the effective decay constant for ith family of delayed neutron precursor (λeff,i), the prompt neutron lifetime (lp) and the mean neutron generation time (Λ). These parameters are calculated using the 3-D diffusion code SRAC-CITATION of the comprehensive neutronics calculation code system SRAC2006 based on the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 in both cases. The calculated results of reactor kinetic parameters are compared to the available safety analysis report (SAR) values of 3 MW TRIGA Mark-II reactor by General Atomic as well as the MCNP5 values (numerically benchmark) based on the evaluated nuclear data library ENDL/B-VII.0. It was found that in most cases, the calculated results of kinetic parameters demonstrate a good agreement between the JENDL-3.3 and the ENDF/B-VII libraries as well as the SAR and the MCNP5 values respectively. Therefore, this study will be essential to improve the basic nuclear data of reactor kinetic parameters for safe operation of 3 MW TRIGA Mark-II research reactor.

Annals of Nuclear Energy, 2013

ABSTRACT The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has bee... more ABSTRACT The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor.

Annals of Nuclear Energy, 2009

The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and ... more The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.

Annals of Nuclear Energy, 2014

ABSTRACT The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the... more ABSTRACT The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of Keff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results.

Annals of Nuclear Energy, 2012

ABSTRACT EUREKA-2/RR code has been used for the analyses of reactivity insertion accident (RIA) a... more ABSTRACT EUREKA-2/RR code has been used for the analyses of reactivity insertion accident (RIA) and loss of flow accident (LOFA) of 3 MW TRIGA Mark-II research reactor of Bangladesh. Transient characteristics of different parameters such as core power, fuel temperature, clad temperature, departure from nucleate boiling ratio (DNBR) due to the different form and magnitude of reactivity insertion has been focused. It is found from the analysis that the magnitude of insertion reactivity and the reactor operating power during this insertion impose a total effect on the core safety. Also, transient effects on reactor were studied for 15% loss of flow of the primary coolant. Provided the scram system is available, the reactor is found to shutdown safely in both cases. From these two studies in series, it is seen that EUREKA-2/RR is well suited for the analyses of reactor safety parameters with good approximations.

Annals of Nuclear Energy, 2013

ABSTRACT The goal of this study is to present the validation study of the SRAC2006 code system ba... more ABSTRACT The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh.

The 3 MW TRIGA Mark II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been op... more The 3 MW TRIGA Mark II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since 1986 without any reshuffling or reloading yet. The key objective of this study was to calculate the core burnup lifetime at 700 MWd of the BAEC TRIGA Low Enrichment Uranium (LEU) fresh core and it is an important characteristic of the in-core fuel management. The core burnup lifetimes at 2 MW power condition was calculated using TRIGAP code. The calculated results were compared to the three dimensional MVP-BURN result and also the reference data of Safety Evaluation Report (NUREG-1282). It was found that the calculated core burnup lifetime shows a good agreement to the MVP-BURN result. By analyzing the results it might be concluded that the TRIGAP code shows a good performance for core burnup lifetime calculation of the LEU fresh core of 3 MW TRIGA MARK II research reactor, which reflects that the TRIGA model is simulated properly by TRIGAP code. Keywords— Fuel Element Burnup, TRIGAP, Core Burnup Lifetime, LEU Fuel, TRIGA, NUREG I. INTRODUCTION Burnup calculations are based upon the assumption that nuclide concentrations can be assumed constant when solving the neutron density distribution. They are formulated around two central equations in reactor physics, which are the neutron transport equation and the burnup equations. The TRIGA MARK II research reactor was commissioned at the Atomic Energy Research Establishment, Savar, Dhaka in 1986. The reactor was designed to implement the various fields of basic nuclear research activities like neutron scattering experiments neutron activation analysis and production of radio isotopes [1]. The reactor is a light water cooled; graphite reflected one, designed for continuous operation at a steady power level of 3000 kW (thermal). The fuels are of single type i.e. only LEU fuel (20 wt.% and 19.7% enriched uranium and 0.47 wt.% erbium) [2]. For the economic and efficient use of the reactor like increased isotope production, burnup calculations and in-core fuel managements are necessary. For this purpose individual fuel element burnup information is needed. This study aims at establishing the applied technological know-how for burnup analysis of the fuel elements loaded initially in the TRIGA core. The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information; burnup calculation is one of the most important parameter of them. The primary thrust of this study is on the burnup calculations and their analysis [3]. This study represents the results of the burnup calculations for TRIGA LEU fuel elements. The TRIGAP [4] computer code that has already been used successfully elsewhere for the analysis of TRIGA research reactor [5] was applied for this analysis. The analyses which have been performed during this study are: (i) determination of individual fuel element burnup in ring-wise, (ii) formulation of effective multiplication factor and excess reactivity and (iii) calculation of core burnup lifetime [1]. This study reflects that the TRIGAP code simulates the TRIGA model properly and this code is suitable for analysis of Core Burnup lifetime calculation of the LEU fresh core of TRIGA MARK II research reactor [6]. This study contributes valuable insight into the behaviour of the reactor and will ensure better utilization and operation of the reactor in future.

International journal of science & technoledge, Jan 31, 2023

Calculation of LEU Fuel Burn-up and Core Life Time Estimation of BAEC TRIGA Research Reactor Usin... more Calculation of LEU Fuel Burn-up and Core Life Time Estimation of BAEC TRIGA Research Reactor Using 2D-TRIGLAV Code 1. Introduction The TRIGA Mark-II research reactor [1] has a maximum thermal neutron flux of 7.46x10 13 n/cm 2 /sec in the middle of the core and is light water-cooled. This graphite-reflected nuclear reactor is intended for continuous operation at a steady state power level of 3 MWt. The TRIGA reactor LEU fuel is made up of burnable poison Erbium, zirconium hydride (primary moderator), and 20 weight percent uranium enriched to 19.7% 235 U. Boron carbide (B4C) serves as the neutron absorber component of the control rods. 100 fuel elements, including five fuelled follower control rods, six control rods, one air follower control rod, 18 graphite dummy elements, one central thimble, one pneumatic transfer system irradiation endpoint, and various light waters make up the BTRR LEU core. All of these components were positioned, supported, and arranged in seven concentric hexagonal rings (A, B, C, D, E, F, and G) of a hexagonal lattice, as shown in figure 1, between the top and bottom grid plates. It was given the go-ahead to carry out a number of nuclear research projects like neutron activation analysis, thermal neutron radiography, and neutron diffraction scattering experiments, as well as to train workers and create radioisotopes for application in medicine, industry, and agriculture. The term 'subcritical' describes a system where the loss of neutrons is greater than the rate of neutron creation [2], and the neutron population gradually declines over time. Criticality is a nuclear term that relates to the balance of neutrons in the core. A system is said to be 'supercritical' when neutron production outpaces neutron loss, increasing the population of neutrons [2, 3]. When the neutron populations are stable, the production and loss of neutrons are perfectly balanced, and the nuclear system is in a critical condition [3]. By comparing the pace at which neutrons are created from fission and other sources to the rate at which they are lost through absorption, scattering, and leakage out of the nuclear reactor core, it is possible to determine the criticality of a system [4]. The neutron diffusion theory code is used to perform this analysis. The configuration of the initial core shape and neutron energy group constants for various homogenized regions of the core, along with the fission spectrum, are required. In this study, neutron group constants were obtained using the well-known 1-D neutron transport code WIMS-D/4 [5] and were utilized for the full core calculations with the TRIGLAV code [6] that is based on a 4-group timeindependent diffusion equation in two-dimensional cylindrical (r,) configuration. The neutron diffusion algorithm is

International journal of science & technoledge, Jan 31, 2023

EL 1 DE JUNIO EL ALBUM “SGTO. PEPPER’S LONELY HEARTS CLUB BAND” CUMPLIRA 50 ANOS, CONSIDERADO POR... more EL 1 DE JUNIO EL ALBUM “SGTO. PEPPER’S LONELY HEARTS CLUB BAND” CUMPLIRA 50 ANOS, CONSIDERADO POR LA REVISTA “ROLLING STONE” COMO EL MAS IMPORTANTE DE TODOS LOS TIEMPOS Y UNO TAN INFLUYENTE QUE, CON SUS 13 CANCIONES, UN SINFIN DE BANDAS APRENDIERON A TOCAR. PARA JULIO MUNOZ RUBIO, DEL CENTRO DE INVESTIGACIONES INTERDISCIPLINARIAS EN CIENCIAS Y HUMANIDADES (CEIICH), SIEMPRE TENEMOS QUE VER CON RECELO LISTADOS TAN TAJANTES, “AUNQUE EN ESTA OCASION DEBEMOS CONCEDER QUE HABLAMOS DE UNA OBRA MAESTRA DEL SIGLO XX QUE ROMPIO CON TODO LO QUE SE HABIA HECHO A LA FECHA”. LANZADO EL 1 DE JUNIO DE 1967 EN INGLATERRA –Y AL DIA SIGUIENTE EN ESTADOS UNIDOS–, EL “SARGENTO PIMIENTA” LLEGO EN EL MOMENTO PRECISO, PUES ES PRODUCTO DE UNA EPOCA EN LA QUE LA JUVENTUD TOMABA POR ASALTO LAS ESFERAS DE PODER Y LAS PERSONAS SE CUESTIONABAN TODO, RECORDO. EL CONTEXTO: VIETNAM EN GUERRA, LA GENTE SE CONGREGABA EN SAN FRANCISCO PARA DAR LUGAR AL LLAMADO VERANO DEL AMOR, EN CUBA SE CONSOLIDABA EL SOCIALISMO, LAS...

International Journal of Nuclear Energy Science and Technology

International Journal of Nuclear Energy Science and Technology

The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile a... more The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the newly generated continuous energy cross section data from JENDL-3.3 was performed against some well-known benchmark lattices using MCNP4C and the results were found to be in very good agreement with the experiment and other evaluations. For TRIGA analysis continuous energy cross section data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300K evaluations were used. Full S(α,β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH...

IEA-R1 reactor core has been modeled in COOLOD-N2 code with the aim to validate and verify the co... more IEA-R1 reactor core has been modeled in COOLOD-N2 code with the aim to validate and verify the code against the experiment data provided by International Atomic Energy Agency (IAEA) through coordinated research project (CRP). Fuel cladding temperatures at different locations of an instrumental fuel assembly (IFA) at different operating powers have been calculated under steady state condition by using the COOLOD-N2 code. The calculated temperatures are then verified against the temperatures recorded by thermo couples placed for those locations of IFA in 243 configuration of IEA-R1 reactor. Reasonable agreement has been found between the calculation and experiment for fuel clad temperature which in turn implies the validity of COOLOD-N2 code in the calculation ofcore thermal hydraulic parameters of research reactors under steady state condition.

Some parametric studies on pulsing mode for fresh core of TRIGA Mark II research reactor in AERE,... more Some parametric studies on pulsing mode for fresh core of TRIGA Mark II research reactor in AERE, Savar, have been carried out with coupled thermal-hydraulics code EUREKA-2/RR in association with neutronics code SRAC. At the beginning, role of some important parameters in pulsing like delayed neutron fraction (βeff) and reactivity insertion have been studied keeping prompt neutron life time (lp) fixed at 33.4 µ-sec. After a series of experiments, we found that the pulsing peak that is consistent with the Safety Analysis Report (SAR) is for the delayed neutron fraction (βeff) of 0.007 and reactivity insertion of 2$. Study has determined the pulsing peak of the fresh core for this particular condition to be 857.86 MW which is 852 MW according to SAR. Experiment also shows the pulsing peak increases with the increase of reactivity insertion whereas decreases with increase of delayed neutron fraction. With the utilization of the particular values of these parameters, pulsing parameters ...

Progress in Nuclear Energy, 2009

ABSTRACT The objective of this paper is to present the results of comparative study of integral p... more ABSTRACT The objective of this paper is to present the results of comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors. The nuclear data processing code NJOY'99 was deployed for the generation of the 69-group cross-section library from the basic evaluated nuclear data files JENDL-3.2 and JEF-2.2. TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project Stage-I. The inputs were the results of a detailed parametric study of the WIMS input options and also optimized for accuracy. The integral parameters (such as keff, ρ28, δ25, δ28, C∗) of five uranium-fuel thermal assemblies: TRX-1 and TRX-2 and BAPL-1, BAPL-2, and BAPL-3 were calculated with the help of WIMSD-5B code based on the generated 69-group cross-section library. The calculated results are compared with those of experiments and it is found that the obtained results between the two libraries are in good agreement with each other. Besides, the calculated integral parameters are also well consistent with the measured values, which reflect the validation of the generated 69-group cross-section library and this library thus obtained is necessary to meet up the nuclear data for neutronics calculation of TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh.

International Journal of Nuclear Energy, 2014

ABSTRACT EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the e... more ABSTRACT EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the experimental results provided by IAEA(International Atomic Energy Agency) obtained for a series of transient tests initiated by step insertion of different magnitudes of positive reactivity with varying degrees of different controlled parameters such as reactor initial power, coolant temperature and coolant flow condition. 20 out of 39 tests that fall under forced convection mode have been considered for the present simulation provided the reactor scram system is disabled. Peak power and peak clad temperature due to transient have been calculated and it was found that although peak clad temperature values agreed, the peak power values seem to underestimate the experimental values. Further study appears to be needed to identify the limitations in modeling or examining the effect of input parameters during modeling to obtain the better simulation results.

Annals of Nuclear Energy, 2013

ABSTRACT The aim of this study is to evaluate the kinetic parameters of 3 MW TRIGA Mark-II resear... more ABSTRACT The aim of this study is to evaluate the kinetic parameters of 3 MW TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety. The most important kinetic parameters of nuclear reactors are the effective delayed neutron fraction (βeff), the effective decay constant for ith family of delayed neutron precursor (λeff,i), the prompt neutron lifetime (lp) and the mean neutron generation time (Λ). These parameters are calculated using the 3-D diffusion code SRAC-CITATION of the comprehensive neutronics calculation code system SRAC2006 based on the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 in both cases. The calculated results of reactor kinetic parameters are compared to the available safety analysis report (SAR) values of 3 MW TRIGA Mark-II reactor by General Atomic as well as the MCNP5 values (numerically benchmark) based on the evaluated nuclear data library ENDL/B-VII.0. It was found that in most cases, the calculated results of kinetic parameters demonstrate a good agreement between the JENDL-3.3 and the ENDF/B-VII libraries as well as the SAR and the MCNP5 values respectively. Therefore, this study will be essential to improve the basic nuclear data of reactor kinetic parameters for safe operation of 3 MW TRIGA Mark-II research reactor.

Annals of Nuclear Energy, 2013

ABSTRACT The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has bee... more ABSTRACT The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor.

Annals of Nuclear Energy, 2009

The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and ... more The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.

Annals of Nuclear Energy, 2014

ABSTRACT The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the... more ABSTRACT The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of Keff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results.

Annals of Nuclear Energy, 2012

ABSTRACT EUREKA-2/RR code has been used for the analyses of reactivity insertion accident (RIA) a... more ABSTRACT EUREKA-2/RR code has been used for the analyses of reactivity insertion accident (RIA) and loss of flow accident (LOFA) of 3 MW TRIGA Mark-II research reactor of Bangladesh. Transient characteristics of different parameters such as core power, fuel temperature, clad temperature, departure from nucleate boiling ratio (DNBR) due to the different form and magnitude of reactivity insertion has been focused. It is found from the analysis that the magnitude of insertion reactivity and the reactor operating power during this insertion impose a total effect on the core safety. Also, transient effects on reactor were studied for 15% loss of flow of the primary coolant. Provided the scram system is available, the reactor is found to shutdown safely in both cases. From these two studies in series, it is seen that EUREKA-2/RR is well suited for the analyses of reactor safety parameters with good approximations.

Annals of Nuclear Energy, 2013

ABSTRACT The goal of this study is to present the validation study of the SRAC2006 code system ba... more ABSTRACT The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh.

The 3 MW TRIGA Mark II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been op... more The 3 MW TRIGA Mark II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since 1986 without any reshuffling or reloading yet. The key objective of this study was to calculate the core burnup lifetime at 700 MWd of the BAEC TRIGA Low Enrichment Uranium (LEU) fresh core and it is an important characteristic of the in-core fuel management. The core burnup lifetimes at 2 MW power condition was calculated using TRIGAP code. The calculated results were compared to the three dimensional MVP-BURN result and also the reference data of Safety Evaluation Report (NUREG-1282). It was found that the calculated core burnup lifetime shows a good agreement to the MVP-BURN result. By analyzing the results it might be concluded that the TRIGAP code shows a good performance for core burnup lifetime calculation of the LEU fresh core of 3 MW TRIGA MARK II research reactor, which reflects that the TRIGA model is simulated properly by TRIGAP code. Keywords— Fuel Element Burnup, TRIGAP, Core Burnup Lifetime, LEU Fuel, TRIGA, NUREG I. INTRODUCTION Burnup calculations are based upon the assumption that nuclide concentrations can be assumed constant when solving the neutron density distribution. They are formulated around two central equations in reactor physics, which are the neutron transport equation and the burnup equations. The TRIGA MARK II research reactor was commissioned at the Atomic Energy Research Establishment, Savar, Dhaka in 1986. The reactor was designed to implement the various fields of basic nuclear research activities like neutron scattering experiments neutron activation analysis and production of radio isotopes [1]. The reactor is a light water cooled; graphite reflected one, designed for continuous operation at a steady power level of 3000 kW (thermal). The fuels are of single type i.e. only LEU fuel (20 wt.% and 19.7% enriched uranium and 0.47 wt.% erbium) [2]. For the economic and efficient use of the reactor like increased isotope production, burnup calculations and in-core fuel managements are necessary. For this purpose individual fuel element burnup information is needed. This study aims at establishing the applied technological know-how for burnup analysis of the fuel elements loaded initially in the TRIGA core. The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information; burnup calculation is one of the most important parameter of them. The primary thrust of this study is on the burnup calculations and their analysis [3]. This study represents the results of the burnup calculations for TRIGA LEU fuel elements. The TRIGAP [4] computer code that has already been used successfully elsewhere for the analysis of TRIGA research reactor [5] was applied for this analysis. The analyses which have been performed during this study are: (i) determination of individual fuel element burnup in ring-wise, (ii) formulation of effective multiplication factor and excess reactivity and (iii) calculation of core burnup lifetime [1]. This study reflects that the TRIGAP code simulates the TRIGA model properly and this code is suitable for analysis of Core Burnup lifetime calculation of the LEU fresh core of TRIGA MARK II research reactor [6]. This study contributes valuable insight into the behaviour of the reactor and will ensure better utilization and operation of the reactor in future.