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Nasir M Mirza - One of the best experts on this subject based on the ideXlab platform.
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parametric study of iodine 129 releases from nuclear fuel to fuel clad gap Primary Coolant in pwrs
Annals of Nuclear Energy, 2019Co-Authors: Rubina Nasir, Sikander M Mirza, Nasir M MirzaAbstract:Abstract A numerical methodology is developed to compute volatile fission fragment release from fuel to gap and then to the Primary Coolant under changing reactor conditions. Based on diffusion of fission fragments and mass balances in gap and in Coolant, a FORTRAN-77 based program DIFFP (Diffusion of Fission Products) has been developed to solve numerically the time & space dependent equations using source release rate respectively. For power shutdown effect of spiking have been modeled for given release rate. The predicted gas fractional release as a function of time using DIFFP program was in good agreement with the reported results. Then parametric studies for empirical diffusion constant (D′), Coolant purification rate (βP), and gas escape rate coefficient (ν) respectively were done to find the sensitivity of parameters. For operation time >150 days, the sensitivity increases from 4.5% to 20.4% with an increase D′ from 4.1 × 10−10 s−1 to 6.9 × 10−10 s−1 respectively and sensitivity is a nonlinear function of D′: sensitivity = ( - 62.20 ± 0.78 ) + ( 16.83 ± 0.37 ) D ′ D o ′ + ( - 0.703 ± 0.0398 ) D ′ D o ′ 2 where, 0.69 D ′ / D 0 ′ 1.17 and D ′ is the reference value (5.9 × 10−10 s−1) for the defective fuel rods. Sensitivity for gas escape rate coefficient (ν) was also determined by increasing it from 5% to 50% from its reference value (1.4 × 10−4 s−1) and results show that the Coolants activity becomes insensitive to changes in the gas escape rate coefficient and sensitivity value becomes less than 1% when effective full power days are larger than 150. If Coolant purification rate constant (βP) is parametrically changed from 5% to 30%, then the absolute sensitivity value increases from 4.8% to 23% respectively and for all cases, the absolute sensitivity remains independent of time (in terms of effective full power days) for a given value of βP and rises linearly with an increase in the value.
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parametric study of time dependent corrosion product activity due to 56mn 58co and 60co in the Primary Coolant circuit of a typical pressurized water reactor
Journal of Chemistry, 2015Co-Authors: Muhammad Rafique, Nasir M Mirza, Sikander M Mirza, Shahab Ahmed Abbasi, Kimberlee J Kearfott, Syed F NaeemAbstract:Results of a detailed study, based on the parametric analysis of activated corrosion products, in Primary Coolant of a typical pressurized water reactor (PWR) are presented. The parameters influencing time dependent buildup of corrosion product activity (CPA) in Primary Coolant loop of PWR were identified. The computer program CPAIR was used to accommodate for time dependent corrosion rates. The behaviors of 56Mn, 58Co, and 60Co were studied over the reactor operational time. During the course of normal operation of reactor, the CPA is dominated by 56Mn, while 58Co and 60Co are the predominant radionuclides after reactor shutdown. Parametric study suggests that the total CPA is most sensitive to ion-exchanger removal rates. For a removal rate of 300 cm3-s−1, the specific activity due to 56Mn has the maximum value of 3.552 × 104 Bq-m−3 after 1,000 hours of reactor operation. This value decreases drastically to 8.325 × 103 Bq-m−3 at removal rate of 900 cm3-s−1. Additionally, CPA due to 56Mn, 58Co, and 60Co shows strong dependence on removal rates from the core material surfaces. Variations in the values of radionuclide removal rates from piping surface and radionuclide removal rate from deposition on pipes showed only very small effects on CPA buildup.
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sensitivity analysis of fission product activity in Primary Coolant of typical pwrs
Progress in Nuclear Energy, 2011Co-Authors: Saeed Ehsan Awan, Sikander M Mirza, Nasir M MirzaAbstract:Abstract A model has been developed for static and dynamic activity analysis of the fission product activity (FPA) in Primary Coolant of typical pressurized water reactors (PWRs). It has been implemented in the FPCART based computer program FPCART-SA. For long steady power operation of reactor, the computed values of normalized static sensitivity have been compared with the corresponding values obtained by using the dynamic sensitivity analysis. The normalized sensitivity values for the reactor power ( P ), failed fuel fraction ( D ), Coolant leakage rate ( L ), total mass of Coolant ( m ) and the let-down flow rate ( Q ) have been calculated and the values: 1.0, 0.857, −2.0177 × 10 −6 , 2.349 × 10 −4 , −2.329 × 10 −4 have been found correspondingly for Kr-88 with the dominant value of FPA as 0.273 μCi/g.
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static and dynamic sensitivity analysis of corrosion product activity in Primary Coolant circuits of pressurized water reactors
Progress in Nuclear Energy, 2010Co-Authors: Sikander M Mirza, Muhammad Rafique, Farooq Ahmad, Nasir M MirzaAbstract:Abstract A CPAIR code ( Deeba et al., 1999 ) based model has been developed for static and dynamic sensitivity analysis of the corrosion product activity (CPA) in Primary circuits of pressurized water reactors (PWRs). This has been implemented in the Matlab program CPAIR-SA and carries out sensitivity analysis of CPA using both static as well as dynamic approaches. For steady state and constant corrosion rate operations, the predicted static value of the normalized sensitivity coefficients have been found in excellent agreement with the corresponding values found by the dynamic analysis. For saturated CPA values, the standard sensitivity coefficients have been found to yield dominant values for re-solution rate coefficients from scale in core and in piping K c , K p as −0.99652, 3.3882 × 10 −3 ; for deposition rate coefficients on core scale and on piping, ɛ c Q c , ɛ p Q p as 0.99798, −8.5652 × 10 −3 ; for removal rate coefficient by ion-exchanger ɛ I Q I as −1.9998; for volume of Primary Coolant, volume of core scale and volume of piping scale, V w , V c , V p as 1.0018, 0.99652 and −3.3882 × 10 −3 respectively.
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stochastic simulation of fission product activity in Primary Coolant due to fuel rod failures in typical pwrs under power transients
Journal of Nuclear Materials, 2008Co-Authors: Javed M Iqbal, Nasir M Mirza, Sikander M MirzaAbstract:Abstract During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the Primary Coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in Primary Coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in Primary Coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in Primary Coolant measured during EDITHMOX-1 experiments.
Sikander M Mirza - One of the best experts on this subject based on the ideXlab platform.
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parametric study of iodine 129 releases from nuclear fuel to fuel clad gap Primary Coolant in pwrs
Annals of Nuclear Energy, 2019Co-Authors: Rubina Nasir, Sikander M Mirza, Nasir M MirzaAbstract:Abstract A numerical methodology is developed to compute volatile fission fragment release from fuel to gap and then to the Primary Coolant under changing reactor conditions. Based on diffusion of fission fragments and mass balances in gap and in Coolant, a FORTRAN-77 based program DIFFP (Diffusion of Fission Products) has been developed to solve numerically the time & space dependent equations using source release rate respectively. For power shutdown effect of spiking have been modeled for given release rate. The predicted gas fractional release as a function of time using DIFFP program was in good agreement with the reported results. Then parametric studies for empirical diffusion constant (D′), Coolant purification rate (βP), and gas escape rate coefficient (ν) respectively were done to find the sensitivity of parameters. For operation time >150 days, the sensitivity increases from 4.5% to 20.4% with an increase D′ from 4.1 × 10−10 s−1 to 6.9 × 10−10 s−1 respectively and sensitivity is a nonlinear function of D′: sensitivity = ( - 62.20 ± 0.78 ) + ( 16.83 ± 0.37 ) D ′ D o ′ + ( - 0.703 ± 0.0398 ) D ′ D o ′ 2 where, 0.69 D ′ / D 0 ′ 1.17 and D ′ is the reference value (5.9 × 10−10 s−1) for the defective fuel rods. Sensitivity for gas escape rate coefficient (ν) was also determined by increasing it from 5% to 50% from its reference value (1.4 × 10−4 s−1) and results show that the Coolants activity becomes insensitive to changes in the gas escape rate coefficient and sensitivity value becomes less than 1% when effective full power days are larger than 150. If Coolant purification rate constant (βP) is parametrically changed from 5% to 30%, then the absolute sensitivity value increases from 4.8% to 23% respectively and for all cases, the absolute sensitivity remains independent of time (in terms of effective full power days) for a given value of βP and rises linearly with an increase in the value.
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parametric study of time dependent corrosion product activity due to 56mn 58co and 60co in the Primary Coolant circuit of a typical pressurized water reactor
Journal of Chemistry, 2015Co-Authors: Muhammad Rafique, Nasir M Mirza, Sikander M Mirza, Shahab Ahmed Abbasi, Kimberlee J Kearfott, Syed F NaeemAbstract:Results of a detailed study, based on the parametric analysis of activated corrosion products, in Primary Coolant of a typical pressurized water reactor (PWR) are presented. The parameters influencing time dependent buildup of corrosion product activity (CPA) in Primary Coolant loop of PWR were identified. The computer program CPAIR was used to accommodate for time dependent corrosion rates. The behaviors of 56Mn, 58Co, and 60Co were studied over the reactor operational time. During the course of normal operation of reactor, the CPA is dominated by 56Mn, while 58Co and 60Co are the predominant radionuclides after reactor shutdown. Parametric study suggests that the total CPA is most sensitive to ion-exchanger removal rates. For a removal rate of 300 cm3-s−1, the specific activity due to 56Mn has the maximum value of 3.552 × 104 Bq-m−3 after 1,000 hours of reactor operation. This value decreases drastically to 8.325 × 103 Bq-m−3 at removal rate of 900 cm3-s−1. Additionally, CPA due to 56Mn, 58Co, and 60Co shows strong dependence on removal rates from the core material surfaces. Variations in the values of radionuclide removal rates from piping surface and radionuclide removal rate from deposition on pipes showed only very small effects on CPA buildup.
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sensitivity analysis of fission product activity in Primary Coolant of typical pwrs
Progress in Nuclear Energy, 2011Co-Authors: Saeed Ehsan Awan, Sikander M Mirza, Nasir M MirzaAbstract:Abstract A model has been developed for static and dynamic activity analysis of the fission product activity (FPA) in Primary Coolant of typical pressurized water reactors (PWRs). It has been implemented in the FPCART based computer program FPCART-SA. For long steady power operation of reactor, the computed values of normalized static sensitivity have been compared with the corresponding values obtained by using the dynamic sensitivity analysis. The normalized sensitivity values for the reactor power ( P ), failed fuel fraction ( D ), Coolant leakage rate ( L ), total mass of Coolant ( m ) and the let-down flow rate ( Q ) have been calculated and the values: 1.0, 0.857, −2.0177 × 10 −6 , 2.349 × 10 −4 , −2.329 × 10 −4 have been found correspondingly for Kr-88 with the dominant value of FPA as 0.273 μCi/g.
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static and dynamic sensitivity analysis of corrosion product activity in Primary Coolant circuits of pressurized water reactors
Progress in Nuclear Energy, 2010Co-Authors: Sikander M Mirza, Muhammad Rafique, Farooq Ahmad, Nasir M MirzaAbstract:Abstract A CPAIR code ( Deeba et al., 1999 ) based model has been developed for static and dynamic sensitivity analysis of the corrosion product activity (CPA) in Primary circuits of pressurized water reactors (PWRs). This has been implemented in the Matlab program CPAIR-SA and carries out sensitivity analysis of CPA using both static as well as dynamic approaches. For steady state and constant corrosion rate operations, the predicted static value of the normalized sensitivity coefficients have been found in excellent agreement with the corresponding values found by the dynamic analysis. For saturated CPA values, the standard sensitivity coefficients have been found to yield dominant values for re-solution rate coefficients from scale in core and in piping K c , K p as −0.99652, 3.3882 × 10 −3 ; for deposition rate coefficients on core scale and on piping, ɛ c Q c , ɛ p Q p as 0.99798, −8.5652 × 10 −3 ; for removal rate coefficient by ion-exchanger ɛ I Q I as −1.9998; for volume of Primary Coolant, volume of core scale and volume of piping scale, V w , V c , V p as 1.0018, 0.99652 and −3.3882 × 10 −3 respectively.
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stochastic simulation of fission product activity in Primary Coolant due to fuel rod failures in typical pwrs under power transients
Journal of Nuclear Materials, 2008Co-Authors: Javed M Iqbal, Nasir M Mirza, Sikander M MirzaAbstract:Abstract During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the Primary Coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in Primary Coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in Primary Coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in Primary Coolant measured during EDITHMOX-1 experiments.
Mei-ya Wang - One of the best experts on this subject based on the ideXlab platform.
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predicted water chemistry in the Primary Coolant circuit of a supercritical water reactor
Nuclear Science and Engineering, 2013Co-Authors: Mei-ya Wang, Tsungkuang Yeh, Hongming Liu, Min LeeAbstract:AbstractAmong the six types of Generation IV reactors, the supercritical water reactor (SCWR) is the only one that adopts light water as the reactor Coolant. Different from the boiling, two-phase Coolant in the core of a traditional boiling water reactor (BWR), the Coolant in an SCWR would remain in one phase throughout the entire Primary Coolant circuit (PCC) due to its much higher operating temperature (>374°C) and pressure (>22.1 MPa). For a conventional BWR, the Coolant is relatively oxidizing due to the presence of hydrogen peroxide and oxygen, directly or indirectly produced via water radiolysis. This outcome eventually leads to degradation of structural materials, primarily stress corrosion cracking. In an SCWR, the solubility of oxygen in the reactor Coolant is extremely high. In the absence of the gas stripping effect in a single-phase Coolant, worse degradation phenomena are expected to appear in the structural and core components. To ensure proper designs of the structural components and suitab...
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water chemistry in the Primary Coolant circuit of a boiling water reactor during startup operations
Nuclear Science and Engineering, 2013Co-Authors: Mei-ya WangAbstract:The Coolant in a boiling water reactor (BWR) usually contains a relatively high level of dissolved oxygen from intrusion of atmospheric air during a cold shutdown. Accordingly, the structural materials in the Primary Coolant circuit (PCC) of a BWR could be exposed to a strongly oxidizing environment for a short period of time during a subsequent startup operation. Due to limitedmeasurable water chemistry data, a well-developedcomputer code DEMACE was used in the current studyto investigate the variations in redox species concentration and in electrochemical corrosion potential (ECP) of components in the PCC of a domestic BWR during startup operations. Our analyses indicated that the dissolved hydrogen level in the reactor Coolant at a low power level without steam generation in the core was higher than that at a power level with a minor amount of steam generated in the core. The dissolved oxygen concentrations in the reactor Coolant would be relatively high and more than 1000 ppb during startup operations at power levels greater than 2.5%. In the meantime, the concentrations of hydrogen peroxide could be more than 2000 ppb at the core outlet region during startup operations, which rendered a strongly oxidizing Coolant environment. The electrochemical corrosion potentials of structural componentsin the PCC of the analyzed BWR generally followed the concentration trend of H2O2. It was predicted that the Coolant environment in a BWR during a plant startup could be highly oxidizing, and the structural components would therefore suffer from a more serious corrosion problem than that under operations at the rated power level.
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impact of power uprate on the water chemistry in the Primary Coolant circuit of a boiling water reactor operating under a fixed core flow rate
Journal of Nuclear Science and Technology, 2008Co-Authors: Mei-ya WangAbstract:The approach of power uprate has been adopted by the utilities of light water reactors over the past few decades in order to increase the power generation efficiency of a nuclear reactor. Upon a power uprate, the power density of a nuclear reactor would change immediately, followed by water chemistry variations due to the enhanced radiolysis of water in the core and near-core regions. For commercial boiling water reactors (BWRs), it is currently a common practice to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal hydrogen injection rate may require a proper adjustment after a power uprate is practiced in a BWR. A DEMACE computer code was used in the current study to investigate the impact of various power uprate levels on major radiolytic species concentrations and the electrochemical corrosion potential (ECP) behavior of components in the Primary Coolant circuit of a domestic BWR operating under either normal water chemistry or HWC. The results of our analysis indicated that the chemical species concentrations and ECP did not vary monotonically with increases in reactor power level at a fixed feedwater hydrogen concentration. In particular, the upper plenum and upper downcomer regions exhibited uniquely higher ECPs at a 102% power level than at the other evaluated power levels. The impact of power uprate on the water chemistry in the Primary Coolant circuit of a BWR is expected to vary from location to location and eventually from plant to plant due to different degrees of radiolysis and physical dimensions.
W Peng - One of the best experts on this subject based on the ideXlab platform.
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experimental investigation of 14c in the Primary Coolant of the 10 mw high temperature gas cooled reactor
Radiocarbon, 2019Co-Authors: Feng Xie, W Peng, Jianzhu Cao, X Feng, Liqiang Wei, Jiejuan Tong, K SunAbstract:The very high temperature reactor (VHTR) is a development of the high-temperature gas-cooled reactors (HTGRs) and one of the six proposed Generation IV reactor concept candidates. The 10 MW high temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China. A sampling system for the measurement of carbon-14 (14C) was established in the helium purification system of the HTR-10 Primary loop, which could sample 14C from the Coolant at three locations. The results showed that activity concentration of 14C in the HTR-10 Primary Coolant was 1.2(1) × 102 Bq/m3 (STP). The production mechanisms, distribution characteristics, reduction routes, and release types of 14C in HTR-10 were analyzed and discussed. A theoretical model was built to calculate the amount of 14C in the core of HTR-10 and its concentration in the Primary Coolant. The activation reaction of 13C has been identified to be the dominant 14C source in the core, whereas in the Primary Coolant, it is the activation of 14N. These results can supplement important information for the source term analysis of 14C in HTR-10 and promote the study of 14C in HTGRs.
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experimental investigation of 14 c in the Primary Coolant of the 10 mw high temperature gas cooled reactor
Radiocarbon, 2019Co-Authors: W Peng, X Feng, Jiejuan Tong, F LiAbstract:The very high temperature reactor (VHTR) is a development of the high-temperature gas-cooled reactors (HTGRs) and one of the six proposed Generation IV reactor concept candidates. The 10 MW high temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China. A sampling system for the measurement of carbon-14 ( 14 C) was established in the helium purification system of the HTR-10 Primary loop, which could sample 14 C from the Coolant at three locations. The results showed that activity concentration of 14 C in the HTR-10 Primary Coolant was 1.2(1) × 10 2 Bq/m 3 (STP). The production mechanisms, distribution characteristics, reduction routes, and release types of 14 C in HTR-10 were analyzed and discussed. A theoretical model was built to calculate the amount of 14 C in the core of HTR-10 and its concentration in the Primary Coolant. The activation reaction of 13 C has been identified to be the dominant 14 C source in the core, whereas in the Primary Coolant, it is the activation of 14 N. These results can supplement important information for the source term analysis of 14 C in HTR-10 and promote the study of 14 C in HTGRs.
K Sun - One of the best experts on this subject based on the ideXlab platform.
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experimental investigation of 14c in the Primary Coolant of the 10 mw high temperature gas cooled reactor
Radiocarbon, 2019Co-Authors: Feng Xie, W Peng, Jianzhu Cao, X Feng, Liqiang Wei, Jiejuan Tong, K SunAbstract:The very high temperature reactor (VHTR) is a development of the high-temperature gas-cooled reactors (HTGRs) and one of the six proposed Generation IV reactor concept candidates. The 10 MW high temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China. A sampling system for the measurement of carbon-14 (14C) was established in the helium purification system of the HTR-10 Primary loop, which could sample 14C from the Coolant at three locations. The results showed that activity concentration of 14C in the HTR-10 Primary Coolant was 1.2(1) × 102 Bq/m3 (STP). The production mechanisms, distribution characteristics, reduction routes, and release types of 14C in HTR-10 were analyzed and discussed. A theoretical model was built to calculate the amount of 14C in the core of HTR-10 and its concentration in the Primary Coolant. The activation reaction of 13C has been identified to be the dominant 14C source in the core, whereas in the Primary Coolant, it is the activation of 14N. These results can supplement important information for the source term analysis of 14C in HTR-10 and promote the study of 14C in HTGRs.