Pub Date : 2024-10-28DOI: 10.1016/j.pnucene.2024.105515
Helical petal-shaped fuel rods have the characteristics of increased heat transfer area and self-supporting positioning, which makes them have great potential for application in small modular reactors. By employing the multi-scale coupled numerical model, the flow and heat transfer characteristics of the petal-shaped bundle in hexagonal arrangement were obtained under natural circulation conditions. The results indicated that the spatial structure of flow channel exhibited a centrally symmetric distribution. Consistent flow and heat transfer behaviors were obtained at symmetric positions. Additionally, fluid viscosity exerted the most significant influence on flow resistance coefficient. Meanwhile, vortices that develop in the opposite direction resulted in flow losses, which induced variations in the resistance coefficient along the channel. Finally, the applicability of existing flow and heat transfer correlations was evaluated under natural circulation conditions. This study has provided significant theoretical guidance for engineering application of the petal-shaped fuel rods.
{"title":"Numerical investigation on flow and heat transfer of petal-shaped fuel bundle in hexagonal arrangement under natural circulation conditions","authors":"","doi":"10.1016/j.pnucene.2024.105515","DOIUrl":"10.1016/j.pnucene.2024.105515","url":null,"abstract":"<div><div>Helical petal-shaped fuel rods have the characteristics of increased heat transfer area and self-supporting positioning, which makes them have great potential for application in small modular reactors. By employing the multi-scale coupled numerical model, the flow and heat transfer characteristics of the petal-shaped bundle in hexagonal arrangement were obtained under natural circulation conditions. The results indicated that the spatial structure of flow channel exhibited a centrally symmetric distribution. Consistent flow and heat transfer behaviors were obtained at symmetric positions. Additionally, fluid viscosity exerted the most significant influence on flow resistance coefficient. Meanwhile, vortices that develop in the opposite direction resulted in flow losses, which induced variations in the resistance coefficient along the channel. Finally, the applicability of existing flow and heat transfer correlations was evaluated under natural circulation conditions. This study has provided significant theoretical guidance for engineering application of the petal-shaped fuel rods.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530082","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-26DOI: 10.1016/j.pnucene.2024.105514
The TOPAZ-II thermionic space reactor system, which was designed by the Soviet Union, is characterized by its high degree of nonlinearity and positive temperature reactivity feedback. The thermionic space reactor exhibits characteristics of high inertia and significant delay in controlling its electrical power and outlet temperature. The simple PID controller is difficult to achieve good performance. To carry out controller design for thermionic space reactor, the simulation platform for thermionic space reactor is developed based on coupling between reactor system thermal-hydraulic code RESYS and control system simulator in this study. After that, based on the cascade control strategy and gain-scheduling, the reactor thermal controller, electric power controller, outlet temperature controller applicable for the full power range is designed with transfer function model and frequency domain analysis method. To validate the nonlinear electric power controller performance, the continuous minor step disturbances, major step disturbances, and ramp variation of electric power setpoint is simulated. The performance of outlet temperature controller is verified with the simulation result of step and ramp variation of outlet temperature setpoint. Thereafter, the start-up process of the thermionic space reactor TOPAZ-II is simulated and analyzed. The simulation result reveals that the controller designed in this paper can overcome the nonlinearity of the thermionic space reactor system and has good performance throughout the entire power range. Compared to traditional simple PID controller, the cascade controller has better performance and can achieve good control performance even in situations where simple PID controllers cannot function properly.
{"title":"Non-linear cascade control with gain-scheduling and startup control strategy study for thermionic space reactor TOPAZ-II","authors":"","doi":"10.1016/j.pnucene.2024.105514","DOIUrl":"10.1016/j.pnucene.2024.105514","url":null,"abstract":"<div><div>The TOPAZ-II thermionic space reactor system, which was designed by the Soviet Union, is characterized by its high degree of nonlinearity and positive temperature reactivity feedback. The thermionic space reactor exhibits characteristics of high inertia and significant delay in controlling its electrical power and outlet temperature. The simple PID controller is difficult to achieve good performance. To carry out controller design for thermionic space reactor, the simulation platform for thermionic space reactor is developed based on coupling between reactor system thermal-hydraulic code RESYS and control system simulator in this study. After that, based on the cascade control strategy and gain-scheduling, the reactor thermal controller, electric power controller, outlet temperature controller applicable for the full power range is designed with transfer function model and frequency domain analysis method. To validate the nonlinear electric power controller performance, the continuous minor step disturbances, major step disturbances, and ramp variation of electric power setpoint is simulated. The performance of outlet temperature controller is verified with the simulation result of step and ramp variation of outlet temperature setpoint. Thereafter, the start-up process of the thermionic space reactor TOPAZ-II is simulated and analyzed. The simulation result reveals that the controller designed in this paper can overcome the nonlinearity of the thermionic space reactor system and has good performance throughout the entire power range. Compared to traditional simple PID controller, the cascade controller has better performance and can achieve good control performance even in situations where simple PID controllers cannot function properly.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530081","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-22DOI: 10.1016/j.pnucene.2024.105502
Radiation shielding materials are key components to suppress the hazardous effects of ionizing radiation, especially energetic gamma rays and penetrative neutrons. This review includes the most ancient building material, concrete, in its different composition obtained by introducing a variety of additives, aggregates and nanomaterials in its conventional form. The objective of present work is to critically review and compare the variety of concretes, reported through various experimental and computational methods, so that the best composition among diverse concrete categories can be highlighted. For this purpose the essential shielding parameters namely mass attenuation coefficient (MAC) and half value layer (HVL) at useful gamma energies of 662, 1173 and 1332 keV have been compared graphically. The protective shielding concretes against neutrons have also been studied through a plot corresponding to Fast Neutron Removal Cross-sections (FNRCS) data of different concretes. Lastly, shielding competency of granite and pyroclastic rock samples for light and heavy charged particles have been included by taking into consideration the interaction parameters namely mass stopping power and projected range. Apart from this, numerous advanced applications of radiation shielding concretes, proper utilization of different forms of waste in concrete mix and few shortcomings of concrete specimens are also listed in this review paper. From the comparative plots of various concretes it is concluded that marble based concretes are best for gamma ray attenuation and nanomaterials based compositions are top if lesser thickness is to employ for attenuation. On the basis of acquired knowledge from literature, the present work will highlight the future perspectives of concretes as shielding materials and would be quite helpful for the selection of appropriate compositions by the community interacting directly or indirectly with ionizing radiations.
{"title":"Efficacy of advanced concretes for attenuation of ionizing radiations: A comprehensive review and comparison","authors":"","doi":"10.1016/j.pnucene.2024.105502","DOIUrl":"10.1016/j.pnucene.2024.105502","url":null,"abstract":"<div><div>Radiation shielding materials are key components to suppress the hazardous effects of ionizing radiation, especially energetic gamma rays and penetrative neutrons. This review includes the most ancient building material, concrete, in its different composition obtained by introducing a variety of additives, aggregates and nanomaterials in its conventional form. The objective of present work is to critically review and compare the variety of concretes, reported through various experimental and computational methods, so that the best composition among diverse concrete categories can be highlighted. For this purpose the essential shielding parameters namely mass attenuation coefficient (MAC) and half value layer (HVL) at useful gamma energies of 662, 1173 and 1332 keV have been compared graphically. The protective shielding concretes against neutrons have also been studied through a plot corresponding to Fast Neutron Removal Cross-sections (FNRCS) data of different concretes. Lastly, shielding competency of granite and pyroclastic rock samples for light and heavy charged particles have been included by taking into consideration the interaction parameters namely mass stopping power and projected range. Apart from this, numerous advanced applications of radiation shielding concretes, proper utilization of different forms of waste in concrete mix and few shortcomings of concrete specimens are also listed in this review paper. From the comparative plots of various concretes it is concluded that marble based concretes are best for gamma ray attenuation and nanomaterials based compositions are top if lesser thickness is to employ for attenuation. On the basis of acquired knowledge from literature, the present work will highlight the future perspectives of concretes as shielding materials and would be quite helpful for the selection of appropriate compositions by the community interacting directly or indirectly with ionizing radiations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530075","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-22DOI: 10.1016/j.pnucene.2024.105505
The time-consuming issue of transport calculations is prominent in the burnup calculation of nuclear reactor. Multi-Group Cross Section (MGXS) method is an acceleration technique developed based on the characteristics of Monte Carlo simulation, which can significantly reduce the computation time required to solve a single group cross section in transportation calculations. The effectiveness of the method has been verified in the test calculations of water reactor pins. However, liquid molten salt reactors (MSRs) exhibit significant differences from conventional water reactors in terms of neutron energy spectra and fuel cycle mode. The effectiveness of the MGXS method in MSR burnup simulations remains to be validated, and targeted adjustments are required during its application. In this study, OpenMC and ORIGEN2 are coupled to develop an accelerated calculation method for MSR burnup simulations based on the MGXS approach. The reasonable grouping structure of the MGXS method is explored, and the performance of different grouping structures is tested. Results show that the transport calculation can be accelerated by an average factor of 2.4 for a single burnup zone by using MGXS method and the acceleration effect is generally independent of the grouping structure adopted. The nuclide mass bias compared to the traditional direct solution can be reduced to approximately 1% when the fuel burnup is 250MWd/kg for the LEU loading scheme with the 10000 groups structure. For the TRU loading scheme, the mass bias compared to the traditional direct solution of important nuclides (such as U-233, U-235, Pu-239 and so on) can be controlled below 0.5% at a burnup of 230 MW d/kg. The results indicate that the grouping strategy proposed in this study can achieve the adaptation of MGXS to MSRs, and the 10000 groups structure adopted in the study exhibits good accuracy.
{"title":"Calculation acceleration for fuel cycle simulation of molten salt reactor based on multi-group cross section method","authors":"","doi":"10.1016/j.pnucene.2024.105505","DOIUrl":"10.1016/j.pnucene.2024.105505","url":null,"abstract":"<div><div>The time-consuming issue of transport calculations is prominent in the burnup calculation of nuclear reactor. Multi-Group Cross Section (MGXS) method is an acceleration technique developed based on the characteristics of Monte Carlo simulation, which can significantly reduce the computation time required to solve a single group cross section in transportation calculations. The effectiveness of the method has been verified in the test calculations of water reactor pins. However, liquid molten salt reactors (MSRs) exhibit significant differences from conventional water reactors in terms of neutron energy spectra and fuel cycle mode. The effectiveness of the MGXS method in MSR burnup simulations remains to be validated, and targeted adjustments are required during its application. In this study, OpenMC and ORIGEN2 are coupled to develop an accelerated calculation method for MSR burnup simulations based on the MGXS approach. The reasonable grouping structure of the MGXS method is explored, and the performance of different grouping structures is tested. Results show that the transport calculation can be accelerated by an average factor of 2.4 for a single burnup zone by using MGXS method and the acceleration effect is generally independent of the grouping structure adopted. The nuclide mass bias compared to the traditional direct solution can be reduced to approximately 1% when the fuel burnup is 250MWd/kg for the LEU loading scheme with the 10000 groups structure. For the TRU loading scheme, the mass bias compared to the traditional direct solution of important nuclides (such as U-233, U-235, Pu-239 and so on) can be controlled below 0.5% at a burnup of 230 MW d/kg. The results indicate that the grouping strategy proposed in this study can achieve the adaptation of MGXS to MSRs, and the 10000 groups structure adopted in the study exhibits good accuracy.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530079","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-22DOI: 10.1016/j.pnucene.2024.105508
The paper considers the effect of sodium diethyldithiocarbamate (SDDC) addition on the immobilization of technetium in a Portland cement matrix. The leaching process was evaluated in a model solution that simulated the conditions of the future radioactive waste (RW) storage site at the Yeniseisky site. The results demonstrated that the addition of 1.0 wt % SDDC to the cement composite increased the incorporation of immobilized technetium into the cement matrix by 35% in comparison to the blank sample. Furthermore, the retention of technetium in the cement matrix was observed to be enhanced, with a retention of up to 90% observed on the 200th day of the experiment. Resulting materials fulfill the necessary technical characteristics for use as a matrix and engineered safety barrier in the concept of deep RW disposal. The addition of SDDC into cement results in microbial respiratory activity decreasing. In order to evaluate the mechanisms of technetium immobilization, the structure of the technetium compound with diethyldithiocarbamate (DDC) [Tc(C5H10NS2)4]TcO4 is described for the first time. This compound contains a Tc(V) atom which coordinates four diethyldithiocarbamate C5H10NS2 moieties through eight sulfur atoms to form a complex cation. The addition of SDDC in cement is presumed to result in a cascade of disproportionation reactions and the formation of stable Tc(IV) compounds, as it was evidenced by XANES spectroscopy.
{"title":"Modification of Portland cement matrix with diethyldithiocarbamate for technetium immobilization","authors":"","doi":"10.1016/j.pnucene.2024.105508","DOIUrl":"10.1016/j.pnucene.2024.105508","url":null,"abstract":"<div><div>The paper considers the effect of sodium diethyldithiocarbamate (SDDC) addition on the immobilization of technetium in a Portland cement matrix. The leaching process was evaluated in a model solution that simulated the conditions of the future radioactive waste (RW) storage site at the Yeniseisky site. The results demonstrated that the addition of 1.0 wt % SDDC to the cement composite increased the incorporation of immobilized technetium into the cement matrix by 35% in comparison to the blank sample. Furthermore, the retention of technetium in the cement matrix was observed to be enhanced, with a retention of up to 90% observed on the 200th day of the experiment. Resulting materials fulfill the necessary technical characteristics for use as a matrix and engineered safety barrier in the concept of deep RW disposal. The addition of SDDC into cement results in microbial respiratory activity decreasing. In order to evaluate the mechanisms of technetium immobilization, the structure of the technetium compound with diethyldithiocarbamate (DDC) [Tc(C<sub>5</sub>H<sub>10</sub>NS<sub>2</sub>)<sub>4</sub>]TcO<sub>4</sub> is described for the first time. This compound contains a Tc(V) atom which coordinates four diethyldithiocarbamate C<sub>5</sub>H<sub>10</sub>NS<sub>2</sub> moieties through eight sulfur atoms to form a complex cation. The addition of SDDC in cement is presumed to result in a cascade of disproportionation reactions and the formation of stable Tc(IV) compounds, as it was evidenced by XANES spectroscopy.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530080","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-21DOI: 10.1016/j.pnucene.2024.105506
A liquid-fueled molten salt reactor (MSR) can reach a deep burnup based on online reprocessing and continuously refueling, which requires significantly different burnup calculation methods for MSRs compared with those for the traditional reactors. To address the unique burnup features and consider the fidelity of isotopic evolution in an MSR, a fuel depletion code ThorMCB is developed based on the OpenMC coupled with a specific depletion code, MODEC. Furthermore, to lower the computational cost of acquiring the equilibrium state through the time evolution step by step for an MSR, an equilibrium burnup calculation code ThorMCB-eq based on the OpenMC and MODEC is developed, which can obtain the equilibrium burnup efficiently. A single fuel lattice of MSR and an a Molten Salt Fast Reactor (MSFR) benchmark are applied for verifying the correctness of the ThorMCB and ThorMCB-eq codes. Compared with a neutron transport calculation code KENO-VI coupled with MODEC, the maximum deviation of the dominant heavy nuclides (HNs) at equilibrium state by ThorMCB is less than 10%, and that of the total mass of fission products (FPs) is less than 3%. For the MSFR benchmark, the neutronic parameters including temperature reactivity coefficient, the mass evolution of main HNs and FPs and breeding ratio (BR) from ThorMCB agree with the references. The equilibrium behavior can be quickly obtained with ThorMCB-eq, and the relative mass deviations of most nuclides keep around 2% in comparison with the results of step-by-step burnup evolution with ThorMCB. Furthermore, the same fuel contents and micro one-group cross sections at equilibrium are obtained with two different types of start-up fuels and a constant power density and fuel reprocessing scheme. In conclusion, the verified results indicate that ThorMCB and ThorMCB-eq can both provide reliable simulation for depletion evolution and equilibrium burnup for MSR fuel cycle.
{"title":"Development of fuel depletion code for molten salt reactor with very deep burnup","authors":"","doi":"10.1016/j.pnucene.2024.105506","DOIUrl":"10.1016/j.pnucene.2024.105506","url":null,"abstract":"<div><div>A liquid-fueled molten salt reactor (MSR) can reach a deep burnup based on online reprocessing and continuously refueling, which requires significantly different burnup calculation methods for MSRs compared with those for the traditional reactors. To address the unique burnup features and consider the fidelity of isotopic evolution in an MSR, a fuel depletion code ThorMCB is developed based on the OpenMC coupled with a specific depletion code, MODEC. Furthermore, to lower the computational cost of acquiring the equilibrium state through the time evolution step by step for an MSR, an equilibrium burnup calculation code ThorMCB-eq based on the OpenMC and MODEC is developed, which can obtain the equilibrium burnup efficiently. A single fuel lattice of MSR and an a Molten Salt Fast Reactor (MSFR) benchmark are applied for verifying the correctness of the ThorMCB and ThorMCB-eq codes. Compared with a neutron transport calculation code KENO-VI coupled with MODEC, the maximum deviation of the dominant heavy nuclides (HNs) at equilibrium state by ThorMCB is less than 10%, and that of the total mass of fission products (FPs) is less than 3%. For the MSFR benchmark, the neutronic parameters including temperature reactivity coefficient, the mass evolution of main HNs and FPs and breeding ratio (BR) from ThorMCB agree with the references. The equilibrium behavior can be quickly obtained with ThorMCB-eq, and the relative mass deviations of most nuclides keep around 2% in comparison with the results of step-by-step burnup evolution with ThorMCB. Furthermore, the same fuel contents and micro one-group cross sections at equilibrium are obtained with two different types of start-up fuels and a constant power density and fuel reprocessing scheme. In conclusion, the verified results indicate that ThorMCB and ThorMCB-eq can both provide reliable simulation for depletion evolution and equilibrium burnup for MSR fuel cycle.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530078","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-19DOI: 10.1016/j.pnucene.2024.105489
The heating source of the nuclear power steam supply installation is PWR secondary steam, but the secondary steam is potentially radioactive, and the radioactive material may migrate to the industrial steam, which goes through the leakage of the heat exchanger. Compared with the conventional industrial steam, the steam supplied by the nuclear power has potential radioactivity. Nuclear power is first introduced to produce the industrial steam, in order to analyze the specific activity and radiation effects of the industrial steam on the staffs at the user side, an analytical model of industrial steam radioactivity is developed based on the technological process of industrial steam under operating conditions, and then the exposure pathways of the staff at the user side from the industrial steam is analyzed. The specific activity of the industrial steam and the exposure dose of the staff is evaluated. According to the evaluation results, it can be seen that the dose to staff caused by the industrial steam is far smaller than the radiation impact on the public caused by the nuclear power plant, and can meet the requirements of relevant regulations and standards. Therefore from the radiation safety point of view, the industrial steam supplied by the nuclear power steam supply installation is safe and acceptable for industrial use.
{"title":"Research on the radioactive safety of industrial steam supplied by PWR nuclear power plant","authors":"","doi":"10.1016/j.pnucene.2024.105489","DOIUrl":"10.1016/j.pnucene.2024.105489","url":null,"abstract":"<div><div>The heating source of the nuclear power steam supply installation is PWR secondary steam, but the secondary steam is potentially radioactive, and the radioactive material may migrate to the industrial steam, which goes through the leakage of the heat exchanger. Compared with the conventional industrial steam, the steam supplied by the nuclear power has potential radioactivity. Nuclear power is first introduced to produce the industrial steam, in order to analyze the specific activity and radiation effects of the industrial steam on the staffs at the user side, an analytical model of industrial steam radioactivity is developed based on the technological process of industrial steam under operating conditions, and then the exposure pathways of the staff at the user side from the industrial steam is analyzed. The specific activity of the industrial steam and the exposure dose of the staff is evaluated. According to the evaluation results, it can be seen that the dose to staff caused by the industrial steam is far smaller than the radiation impact on the public caused by the nuclear power plant, and can meet the requirements of relevant regulations and standards. Therefore from the radiation safety point of view, the industrial steam supplied by the nuclear power steam supply installation is safe and acceptable for industrial use.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530077","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-18DOI: 10.1016/j.pnucene.2024.105491
Several nuclear reactors use ice condensers to condense steam in the case of a loss of coolant accident. These ice condensers have many problems that could be alleviated by using another material. The effects of low-dose neutron and gamma radiation on the thermal properties of polyethylene wax (PEW) were investigated for this purpose. PEW was irradiated in the Missouri University of Science and Technology Research Reactor (MSTR), the University of Missouri Cyclotron (MUC) and the University of Missouri Research Reactor (MURR) up to doses equivalent to 10 months in a nuclear power reactor's containment structure. The melting temperature and latent heat of fusion were determined using differential scanning calorimetry (DSC). Changes in the molecular bonds was determined using Raman spectroscopy. It was found that there was not a significant change in the thermal properties nor bonding over the investigated doses. This suggests that organic PCMs could be reliable alternatives to ice in nuclear reactor containment applications. The measured melting peak was found to be significantly wider expected by the suppliers' description. The ramifications of wide melting peaks are discussed in the context of reactor accident analysis and further experiments are suggested.
{"title":"Neutron and gamma radiation effects on thermal storage properties of polyethylene wax","authors":"","doi":"10.1016/j.pnucene.2024.105491","DOIUrl":"10.1016/j.pnucene.2024.105491","url":null,"abstract":"<div><div>Several nuclear reactors use ice condensers to condense steam in the case of a loss of coolant accident. These ice condensers have many problems that could be alleviated by using another material. The effects of low-dose neutron and gamma radiation on the thermal properties of polyethylene wax (PEW) were investigated for this purpose. PEW was irradiated in the Missouri University of Science and Technology Research Reactor (MSTR), the University of Missouri Cyclotron (MUC) and the University of Missouri Research Reactor (MURR) up to doses equivalent to 10 months in a nuclear power reactor's containment structure. The melting temperature and latent heat of fusion were determined using differential scanning calorimetry (DSC). Changes in the molecular bonds was determined using Raman spectroscopy. It was found that there was not a significant change in the thermal properties nor bonding over the investigated doses. This suggests that organic PCMs could be reliable alternatives to ice in nuclear reactor containment applications. The measured melting peak was found to be significantly wider expected by the suppliers' description. The ramifications of wide melting peaks are discussed in the context of reactor accident analysis and further experiments are suggested.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142445671","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-18DOI: 10.1016/j.pnucene.2024.105507
Pressurized Water Reactors (PWRs) management presents the core reloading pattern optimization as a significant problem that contributes to the improvement of reactor productivity and optimal fuel utilization with considering safety conditions. In present study, Multi-Objective Elitist Teaching-Learning-Based Optimization (MO-ETLBO) technique is suggested to cope up the multi-objective reloading optimization problem of the Chashma Nuclear Power Generating Station (CNPGS) unit-3 core. A multivariable objective function is designed to evaluate the quality of each loading pattern while maximizing critical boron concentration (CBC), minimizing the power peaking factor (PPF), and optimally enhancing the cycle length while ensuring adequate safety margins and design limits. It has been found that the equilibrium cycle can be extended to 16.07 extended full power days (EFPDs) while maintaining the PPF and CBC within design limits. To validate the effectiveness of TLBO, the optimized loading pattern of the equilibrium core was evaluated using the deterministic computer code DONJON5, for neutronic parameter analysis. The results show that the algorithm proposed in this study is a promising approach for reloading pattern optimization in CNPGS unit-3, offering potential improvements in reactor cycle length while ensuring safety and enhancing overall performance.
{"title":"Generating optimal reloading patterns for CNPGS Unit-3 core using multi-objective elitist teaching-learning-based optimizer","authors":"","doi":"10.1016/j.pnucene.2024.105507","DOIUrl":"10.1016/j.pnucene.2024.105507","url":null,"abstract":"<div><div>Pressurized Water Reactors (PWRs) management presents the core reloading pattern optimization as a significant problem that contributes to the improvement of reactor productivity and optimal fuel utilization with considering safety conditions. In present study, Multi-Objective Elitist Teaching-Learning-Based Optimization (MO-ETLBO) technique is suggested to cope up the multi-objective reloading optimization problem of the Chashma Nuclear Power Generating Station (CNPGS) unit-3 core. A multivariable objective function is designed to evaluate the quality of each loading pattern while maximizing critical boron concentration (CBC), minimizing the power peaking factor (PPF), and optimally enhancing the cycle length while ensuring adequate safety margins and design limits. It has been found that the equilibrium cycle can be extended to 16.07 extended full power days (EFPDs) while maintaining the PPF and CBC within design limits. To validate the effectiveness of TLBO, the optimized loading pattern of the equilibrium core was evaluated using the deterministic computer code DONJON5, for neutronic parameter analysis. The results show that the algorithm proposed in this study is a promising approach for reloading pattern optimization in CNPGS unit-3, offering potential improvements in reactor cycle length while ensuring safety and enhancing overall performance.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530076","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-17DOI: 10.1016/j.pnucene.2024.105501
Condition monitoring is essential in industrial processes to ensure safe and efficient operations. Sensor signals, which accurately reflect the state of industrial systems, play a central role in this monitoring. However, the harsh conditions in many industrial environments, especially in nuclear power plants, increase the likelihood of sensor failures. Condition monitoring systems detect anomalies by reconstructing input data, with high reconstruction errors indicating the presence of anomalies. The Multivariate State Estimation Technique (MSET) is a widely used nonlinear, non-parametric model for condition monitoring. Traditional nonlinear models assume that training and test data come from the same distribution. This assumption can lead to significant errors when the model encounters anomalies, making it challenging to detect and reconstruct sensor states. To address these challenges, this paper introduces a self-correcting anomaly diagnosis model. Unlike traditional methods, this model establishes a dedicated data structure to store normal sensor patterns and generates a dynamic memory matrix that adapts to changes in industrial processes; The proposed method combines penalized offset projection with multi-scale estimation to mitigate the impact of anomalies on estimation results. Additionally, a variable correlation analysis method is developed to optimize input feature selection for the model. The new approach self-corrects anomalous data in a transformed signal space, achieving accurate reconstruction of sensor states. The model's performance is validated using real sensor data from a nuclear power plant system. Results demonstrate that the proposed model significantly enhances signal reconstruction and anomaly detection capabilities, even under more severe simulated conditions. Compared to traditional nonlinear models, the new method improves the metric for reducing anomaly interference by an order of magnitude. However, we did not change the calculation method of the higher-order kernel in the original method, which still faces the problem of matrix inversion.
{"title":"Research on sensor condition monitoring and signal reconstruction based on self-correcting anomaly diagnosis model","authors":"","doi":"10.1016/j.pnucene.2024.105501","DOIUrl":"10.1016/j.pnucene.2024.105501","url":null,"abstract":"<div><div>Condition monitoring is essential in industrial processes to ensure safe and efficient operations. Sensor signals, which accurately reflect the state of industrial systems, play a central role in this monitoring. However, the harsh conditions in many industrial environments, especially in nuclear power plants, increase the likelihood of sensor failures. Condition monitoring systems detect anomalies by reconstructing input data, with high reconstruction errors indicating the presence of anomalies. The Multivariate State Estimation Technique (MSET) is a widely used nonlinear, non-parametric model for condition monitoring. Traditional nonlinear models assume that training and test data come from the same distribution. This assumption can lead to significant errors when the model encounters anomalies, making it challenging to detect and reconstruct sensor states. To address these challenges, this paper introduces a self-correcting anomaly diagnosis model. Unlike traditional methods, this model establishes a dedicated data structure to store normal sensor patterns and generates a dynamic memory matrix that adapts to changes in industrial processes; The proposed method combines penalized offset projection with multi-scale estimation to mitigate the impact of anomalies on estimation results. Additionally, a variable correlation analysis method is developed to optimize input feature selection for the model. The new approach self-corrects anomalous data in a transformed signal space, achieving accurate reconstruction of sensor states. The model's performance is validated using real sensor data from a nuclear power plant system. Results demonstrate that the proposed model significantly enhances signal reconstruction and anomaly detection capabilities, even under more severe simulated conditions. Compared to traditional nonlinear models, the new method improves the metric for reducing anomaly interference by an order of magnitude. However, we did not change the calculation method of the higher-order kernel in the original method, which still faces the problem of matrix inversion.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142445670","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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