Pub Date : 2024-11-27DOI: 10.1016/j.nucengdes.2024.113720
Jacob A. Hirschhorn, Mustafa K. Jaradat, Ryan T. Sweet, Paul A. Demkowicz, Paolo Balestra, Gerhard Strydom
Uranium oxycarbide (UCO)-bearing tri-structural isotropic (TRISO) particle fuels are expected to be used in numerous U.S. commercial reactor applications within the next decade. In this work, we reviewed historical particle fuel transient experiments to identify gaps in TRISO fuel performance transient testing. A BISON–Griffin modeling framework was then developed to conduct preliminary TRISO transient analyses and begin to address these gaps. The framework was demonstrated using limiting-case transient conditions from a prototypic high-temperature gas-cooled reactor (HTGR). It was then applied to develop a matrix of experiments that could be performed in the Transient Reactor Test Facility (TREAT) to (1) evaluate UCO-fueled particle performance at moderate and high heat rates, (2) assess whether historical testing involving UO-fueled particles is applicable to modern UCO-fueled particles, (3) deconvolute the impacts of temperature and heat rate on particle transient response, and (4) collect the data needed for fuel performance model validation and/or further development.
{"title":"Development and demonstration of a BISON–Griffin modeling framework for the design of targeted TRISO transient experiments in the Transient Reactor Test Facility","authors":"Jacob A. Hirschhorn, Mustafa K. Jaradat, Ryan T. Sweet, Paul A. Demkowicz, Paolo Balestra, Gerhard Strydom","doi":"10.1016/j.nucengdes.2024.113720","DOIUrl":"10.1016/j.nucengdes.2024.113720","url":null,"abstract":"<div><div>Uranium oxycarbide (UCO)-bearing tri-structural isotropic (TRISO) particle fuels are expected to be used in numerous U.S. commercial reactor applications within the next decade. In this work, we reviewed historical particle fuel transient experiments to identify gaps in TRISO fuel performance transient testing. A BISON–Griffin modeling framework was then developed to conduct preliminary TRISO transient analyses and begin to address these gaps. The framework was demonstrated using limiting-case transient conditions from a prototypic high-temperature gas-cooled reactor (HTGR). It was then applied to develop a matrix of experiments that could be performed in the Transient Reactor Test Facility (TREAT) to (1) evaluate UCO-fueled particle performance at moderate and high heat rates, (2) assess whether historical testing involving UO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>-fueled particles is applicable to modern UCO-fueled particles, (3) deconvolute the impacts of temperature and heat rate on particle transient response, and (4) collect the data needed for fuel performance model validation and/or further development.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113720"},"PeriodicalIF":1.9,"publicationDate":"2024-11-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142722111","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 : 2024-11-26DOI: 10.1016/j.nucengdes.2024.113733
Sayed Saeed Mustafa
Following the nuclear disaster at Fukushima in Japan in 2011, there is a growing search for novel cladding materials that can displace Zirconium in light water reactors. In this paper, niobium (Nb), titanium (Ti), vanadium (Va) and Inconel-600 alloy are selected as possible innovative cladding materials that have high melting points and resist corrosion. MCNPX code was used to simulate these cladding materials in a standard pressurized water reactor assembly. The impact of these materials on the reactor safety aspects was discussed in terms of the depletion calculations at the unit cell and assembly levels. The included reactor safety aspects in this work are effective multiplication factor (Keff), cycle length, relative fission power, reactivity coefficients, reactivity worth, fission products and actinides, neutron spectrum, spectral index, radial power distribution and peaking factor. For each proposed cladding material, the study focused on determining the required thickness (at constant enrichment) and evaluating the suitable enrichment (at constant cladding thickness) to obtain the same cycle length of zirconium. The simulation depicted that the lowest decrease of cycle length was observed for niobium which contributed to reducing the Zirconium cycle length by 15%. Meanwhile, the high absorbing cladding materials such as Ti, Va and Inconel-600 reduced the Zirconium cycle length by 29%, 32% and 40%, respectively. Enhanced negativity of fuel temperature coefficient (FTC), moderator temperature coefficient (MTC) and void reactivity coefficient are noticed for Ti, Va, and Inconel-600 at the BOL. On the other hand, Zr and Nb provide the most negativity of reactivity coefficients at the MOL and EOL owing to the low inventory of Pu-239 and fission products. The control rod worth values of Zr and Nb are larger than those of Ti, Va and Inconel-600 throughout the fuel depletion thanks to the softening of neutron spectrum in the case of Zr and Nb. In terms of minimizing the radioactive waste, Nb offers the second lowest inventory of fission products and actinides after zirconium. Finally, the peaking factors for Inconel-600, Va and Ti are slightly higher than those for Zr and Nb. As a consequence, the power distribution is more controllable in the cases of Zr and Nb.
2011 年日本福岛核灾难发生后，人们越来越多地寻求能在轻水反应堆中取代锆的新型包壳材料。本文选择了铌 (Nb)、钛 (Ti)、钒 (Va) 和 Inconel-600 合金作为可能的创新包壳材料，这些材料具有高熔点和耐腐蚀性。MCNPX 代码用于模拟这些包层材料在标准压水堆组件中的应用。这些材料对反应堆安全方面的影响在单元和组件层面的损耗计算中进行了讨论。这项工作包括的反应堆安全方面包括有效倍增因子（Keff）、循环长度、相对裂变功率、反应系数、反应值、裂变产物和锕系元素、中子谱、光谱指数、径向功率分布和峰值因数。对于每种拟议的包层材料，研究的重点是确定所需的厚度（在富集度不变的情况下），并评估合适的富集度（在包层厚度不变的情况下），以获得与锆相同的循环长度。模拟结果表明，铌的周期长度减少最少，使锆的周期长度减少了 15%。同时，高吸收包层材料（如钛、钒和铬镍铁合金-600）的锆循环时间分别缩短了 29%、32% 和 40%。在 BOL 处，Ti、Va 和 Inconel-600 的燃料温度系数（FTC）、慢化剂温度系数（MTC）和空隙反应系数的负值都有所提高。另一方面，由于钚 239 和裂变产物的存量较低，Zr 和 Nb 在 MOL 和 EOL 时的反应系数负值最大。在整个燃料耗尽过程中，Zr 和 Nb 的控制棒值要大于 Ti、Va 和 Inconel-600 的控制棒值，这要归功于 Zr 和 Nb 中子谱的软化。在减少放射性废物方面，铌的裂变产物和锕系元素存量仅次于锆。最后，Inconel-600、Va 和 Ti 的峰值系数略高于 Zr 和 Nb。因此，Zr 和 Nb 的功率分布更容易控制。
{"title":"The possibility of utilizing novel cladding materials instead of zirconium in light water reactors","authors":"Sayed Saeed Mustafa","doi":"10.1016/j.nucengdes.2024.113733","DOIUrl":"10.1016/j.nucengdes.2024.113733","url":null,"abstract":"<div><div>Following the nuclear disaster at Fukushima in Japan in 2011, there is a growing search for novel cladding materials that can displace Zirconium in light water reactors. In this paper, niobium (Nb), titanium (Ti), vanadium (Va) and Inconel-600 alloy are selected as possible innovative cladding materials that have high melting points and resist corrosion. MCNPX code was used to simulate these cladding materials in a standard pressurized water reactor assembly. The impact of these materials on the reactor safety aspects was discussed in terms of the depletion calculations at the unit cell and assembly levels. The included reactor safety aspects in this work are effective multiplication factor (K<strong><em><sub>eff</sub></em></strong>), cycle length, relative fission power, reactivity coefficients, reactivity worth, fission products and actinides, neutron spectrum, spectral index, radial power distribution and peaking factor. For each proposed cladding material, the study focused on determining the required thickness (at constant enrichment) and evaluating the suitable enrichment (at constant cladding thickness) to obtain the same cycle length of zirconium. The simulation depicted that the lowest decrease of cycle length was observed for niobium which contributed to reducing the Zirconium cycle length by 15%. Meanwhile, the high absorbing cladding materials such as Ti, Va and Inconel-600 reduced the Zirconium cycle length by 29%, 32% and 40%, respectively. Enhanced negativity of fuel temperature coefficient (FTC), moderator temperature coefficient (MTC) and void reactivity coefficient are noticed for Ti, Va, and Inconel-600 at the BOL. On the other hand, Zr and Nb provide the most negativity of reactivity coefficients at the MOL and EOL owing to the low inventory of Pu-239 and fission products. The control rod worth values of Zr and Nb are larger than those of Ti, Va and Inconel-600 throughout the fuel depletion thanks to the softening of neutron spectrum in the case of Zr and Nb. In terms of minimizing the radioactive waste, Nb offers the second lowest inventory of fission products and actinides after zirconium. Finally, the peaking factors for Inconel-600, Va and Ti are slightly higher than those for Zr and Nb. As a consequence, the power distribution is more controllable in the cases of Zr and Nb.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113733"},"PeriodicalIF":1.9,"publicationDate":"2024-11-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701177","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 : 2024-11-26DOI: 10.1016/j.nucengdes.2024.113670
L. Tiborcz , S. Beck
Severe accident analysis is an important component in ensuring high standards of safety in nuclear power plants. Since the accident in the Fukushima-Daichi NPP even more attention has been paid to this highly complicated and complex field. Numerical simulation tools are widely used to analyse postulated accident sequences, including severe accidents, as well as the evaluation of their possible radioactive impact on the environment. The system code package AC2 developed by GRS can simulate the whole reactor in detail, both the core region with the RCS, as well as the containment, starting from normal operational conditions up to severe accidents including core melting, making it a highly valuable tool. At the same time, such tools and their models are developed based on a limited number of experiments and available data, particularly models related to severe accident phenomena. Therefore, it is of great interest to be able to evaluate their accuracy and uncertainty on their respective application fields. In this paper an approach developed and tested on the Phébus FPT1 experiment is applied to a reactor scenario in a generic German PWR to assess source term related uncertainties. A full scale AC2 simulation (ATHLET-CD/COCOSYS) is carried out for a medium break LOCA with station blackout in a generic German PWR and used as a best estimate case for the BEPU analysis. The uncertainty and sensitivity analysis focuses on source term related phenomena and figures of merit. Altogether over 80 uncertain input parameters directly related to the modelling of fission product behaviour are considered. In addition to the 95/95 tolerance limits, sensitivity measures (Spearman Rank Correlation Coefficient) are derived to further analyse the dependency of the simulation results on different input parameters.
{"title":"Uncertainty and sensitivity analysis of a postulated severe accident in a generic German PWR with the system code AC2","authors":"L. Tiborcz , S. Beck","doi":"10.1016/j.nucengdes.2024.113670","DOIUrl":"10.1016/j.nucengdes.2024.113670","url":null,"abstract":"<div><div>Severe accident analysis is an important component in ensuring high standards of safety in nuclear power plants. Since the accident in the Fukushima-Daichi NPP even more attention has been paid to this highly complicated and complex field. Numerical simulation tools are widely used to analyse postulated accident sequences, including severe accidents, as well as the evaluation of their possible radioactive impact on the environment. The system code package AC<sup>2</sup> developed by GRS can simulate the whole reactor in detail, both the core region with the RCS, as well as the containment, starting from normal operational conditions up to severe accidents including core melting, making it a highly valuable tool. At the same time, such tools and their models are developed based on a limited number of experiments and available data, particularly models related to severe accident phenomena. Therefore, it is of great interest to be able to evaluate their accuracy and uncertainty on their respective application fields. In this paper an approach developed and tested on the Phébus FPT1 experiment is applied to a reactor scenario in a generic German PWR to assess source term related uncertainties. A full scale AC<sup>2</sup> simulation (ATHLET-CD/COCOSYS) is carried out for a medium break LOCA with station blackout in a generic German PWR and used as a best estimate case for the BEPU analysis. The uncertainty and sensitivity analysis focuses on source term related phenomena and figures of merit. Altogether over 80 uncertain input parameters directly related to the modelling of fission product behaviour are considered. In addition to the 95/95 tolerance limits, sensitivity measures (Spearman Rank Correlation Coefficient) are derived to further analyse the dependency of the simulation results on different input parameters.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113670"},"PeriodicalIF":1.9,"publicationDate":"2024-11-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142722700","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-26DOI: 10.1016/j.nucengdes.2024.113728
Namgyu Park
Corrected nonhomogeneous solution is presented for the original paper (Park et al., 2024), which provided correlations between random variables using both homogeneous and nonhomogeneous solutions for the fuel assembly with random geometric boundary conditions. The author made the mistake of applying incorrect mixed boundary conditions to obtain the nonhomogeneous solution. This work shows that the conclusions of the original paper are still valid even when considering the modified nonhomogeneous solution.
{"title":"Correction for nonhomogeneous solution of fuel assembly with random geometric boundary conditions","authors":"Namgyu Park","doi":"10.1016/j.nucengdes.2024.113728","DOIUrl":"10.1016/j.nucengdes.2024.113728","url":null,"abstract":"<div><div>Corrected nonhomogeneous solution is presented for the original paper (<span><span>Park et al., 2024</span></span>), which provided correlations between random variables using both homogeneous and nonhomogeneous solutions for the fuel assembly with random geometric boundary conditions. The author made the mistake of applying incorrect mixed boundary conditions to obtain the nonhomogeneous solution. This work shows that the conclusions of the original paper are still valid even when considering the modified nonhomogeneous solution.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113728"},"PeriodicalIF":1.9,"publicationDate":"2024-11-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701178","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 : 2024-11-26DOI: 10.1016/j.nucengdes.2024.113710
Yunlong Zhou, Yiwen Ran, Qichao Liu, Shibo Zhang
Accurate prediction of void fraction of gas–liquid two-phase flow under fluctuating vibration is crucial for the safe and stable operation of floating nuclear power plants. The void fraction characteristics of gas–liquid two-phase flow in horizontal pipe under different vibration conditions are studied experimentally. The results showed that the void fraction of bubbly flow and intermittent flow varies considerably under fluctuating vibration, whereas changes in stratified flow and annular flow are less pronounced. Generally speaking, the void fraction first increases and then decreases with the increase of pipe diameter, while the increase of vibration frequency and amplitude cause a nonlinear variation in the void fraction. Evaluation of void fraction calculation models for stationary pipes reveals that existing models have significant prediction errors for bubbly flow and intermittent flow void fractions. By considering the effects of pipe diameter and vibration parameters, the Froude number of liquid phase is introduced to develop a void fraction calculation model for bubbly flow and intermittent flow. The Mean Absolute Relative Difference (MARD) of new established model is 10.66% and 12.06%. This significantly improved the prediction accuracy of the void fraction under fluctuating vibration.
{"title":"Investigation on void fraction of gas–liquid two-phase flow in horizontal pipe under fluctuating vibration","authors":"Yunlong Zhou, Yiwen Ran, Qichao Liu, Shibo Zhang","doi":"10.1016/j.nucengdes.2024.113710","DOIUrl":"10.1016/j.nucengdes.2024.113710","url":null,"abstract":"<div><div>Accurate prediction of void fraction of gas–liquid two-phase flow under fluctuating vibration is crucial for the safe and stable operation of floating nuclear power plants. The void fraction characteristics of gas–liquid two-phase flow in horizontal pipe under different vibration conditions are studied experimentally. The results showed that the void fraction of bubbly flow and intermittent flow varies considerably under fluctuating vibration, whereas changes in stratified flow and annular flow are less pronounced. Generally speaking, the void fraction first increases and then decreases with the increase of pipe diameter, while the increase of vibration frequency and amplitude cause a nonlinear variation in the void fraction. Evaluation of void fraction calculation models for stationary pipes reveals that existing models have significant prediction errors for bubbly flow and intermittent flow void fractions. By considering the effects of pipe diameter and vibration parameters, the Froude number of liquid phase is introduced to develop a void fraction calculation model for bubbly flow and intermittent flow. The Mean Absolute Relative Difference (MARD) of new established model is 10.66% and 12.06%. This significantly improved the prediction accuracy of the void fraction under fluctuating vibration.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113710"},"PeriodicalIF":1.9,"publicationDate":"2024-11-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701235","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}
Sodium-cooled fast reactors (SFR) are included in the list of fourth-generation nuclear energy systems, the so-called generation IV (GEN-IV). It is the only technology of GEN-IV that possesses the significant practical experience of design, construction, and operation of high-power reactors. Due to the high temperature of the coolant at the BN-600 and BN-800 reactor core outlets, the steam generator produces steam with higher enthalpy, and this significantly increases the efficiency of cogeneration at SFR power plants in comparison with pressured water reactor (PWR) or boiling water reactor (BWR) units. The primary nuclear fuel effectiveness rises due to cogeneration with higher efficiency of a sodium-cooled fast reactor nuclear power plant (NPP) in comparison with thermal reactors. In addition, it is necessary to reduce the reactor refueling outages to improve the nuclear power plant utilization factor.
It is possible to reduce the duration of the reactor refueling by the route optimization of the refueling machine movement. The prevailing in the world thermal neutron reactors with water cooling have a refueling machine that points the certain fuel assembly using two coordinates. The fast neutron reactors use sodium as a coolant; thus, there is a problem with its violent reaction with water and air oxygen. Therefore, it is necessary to exclude the contact of sodium and surrounding air during the refueling. To achieve this, the system of refueling machine pointing is used, consisting of two or three eccentrically located rotating plugs with hydraulic locks. The article presents the results of the creation of mathematical models of refueling machine movement and fuel assembly gripping. The time-optimal algorithms for the operation of refueling machines with three rotating plugs are proposed. The use of the new algorithm allows to reduce the time of movement of the grip of the refueling machine by 30–37%.
{"title":"Optimization of the route of the refueling machine to reduce the refueling time: Case study of the BN-800 reactor","authors":"O.L. Tashlykov, A.N. Sesekin, V.A. Klimova, K.A. Mahmoud","doi":"10.1016/j.nucengdes.2024.113725","DOIUrl":"10.1016/j.nucengdes.2024.113725","url":null,"abstract":"<div><div>Sodium-cooled fast reactors (SFR) are included in the list of fourth-generation nuclear energy systems, the so-called generation IV (GEN-IV). It is the only technology of GEN-IV that possesses the significant practical experience of design, construction, and operation of high-power reactors. Due to the high temperature of the coolant at the BN-600 and BN-800 reactor core outlets, the steam generator produces steam with higher enthalpy, and this significantly increases the efficiency of cogeneration at SFR power plants in comparison with pressured water reactor (PWR) or boiling water reactor (BWR) units. The primary nuclear fuel effectiveness rises due to cogeneration with higher efficiency of a sodium-cooled fast reactor nuclear power plant (NPP) in comparison with thermal reactors. In addition, it is necessary to reduce the reactor refueling outages to improve the nuclear power plant utilization factor.</div><div>It is possible to reduce the duration of the reactor refueling by the route optimization of the refueling machine movement. The prevailing in the world thermal neutron reactors with water cooling have a refueling machine that points the certain fuel assembly using two coordinates. The fast neutron reactors use sodium as a coolant; thus, there is a problem with its violent reaction with water and air oxygen. Therefore, it is necessary to exclude the contact of sodium and surrounding air during the refueling. To achieve this, the system of refueling machine pointing is used, consisting of two or three eccentrically located rotating plugs with hydraulic locks. The article presents the results of the creation of mathematical models of refueling machine movement and fuel assembly gripping. The time-optimal algorithms for the operation of refueling machines with three rotating plugs are proposed. The use of the new algorithm allows to reduce the time of movement of the grip of the refueling machine by 30–37%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113725"},"PeriodicalIF":1.9,"publicationDate":"2024-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701233","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}
The Plutonium Fuel Fabrication Facility is currently in the decommissioning phase, with glovebox dismantling operations ongoing since 2010. During conventional glovebox dismantling operations, the glovebox to be dismantled is enclosed within plastic tents to contain contamination. The glovebox is then dismantled by workers wearing air-fed suits with thermal or mechanical cutting tools, which typically generate dross or sparks in the form of radioactive aerosols during cutting. Despite the longevity and meticulous organization of this manual method, the workload remains considerable, while the allowable working time is limited. In addition, the potential for inhalation exposure to plutonium is elevated in the event of an accident given the contamination of the work area. To overcome disadvantages associated with conventional glovebox dismantling methods, new methods are currently being developed. The primary objective is to reduce the reliance on operation based on air-fed suits and enhance worker safety by introducing remote equipment and a new floor-reinforcing panel. Another objective is to suppress waste generation by reusing all equipment on multiple occasions which is achieved by developing a containment system that have a large open port with a pallet for the storage and reuse of equipment for successive operations. Furthermore, a glove operation compartment is designed and tested for the manual handling of dismantled materials as an additional strategy to reduce work based on air-fed suits and mitigate secondary waste generation.
{"title":"Development of plutonium fuel facility decommissioning technology to accelerate glovebox dismantling and reduce air-fed suits based operations","authors":"Masato Yoshida, Satoshi Iguchi, Hiroshi Hirano, Akihiro Kitamura","doi":"10.1016/j.nucengdes.2024.113691","DOIUrl":"10.1016/j.nucengdes.2024.113691","url":null,"abstract":"<div><div>The Plutonium Fuel Fabrication Facility is currently in the decommissioning phase, with glovebox dismantling operations ongoing since 2010. During conventional glovebox dismantling operations, the glovebox to be dismantled is enclosed within plastic tents to contain contamination. The glovebox is then dismantled by workers wearing air-fed suits with thermal or mechanical cutting tools, which typically generate dross or sparks in the form of radioactive aerosols during cutting. Despite the longevity and meticulous organization of this manual method, the workload remains considerable, while the allowable working time is limited. In addition, the potential for inhalation exposure to plutonium is elevated in the event of an accident given the contamination of the work area. To overcome disadvantages associated with conventional glovebox dismantling methods, new methods are currently being developed. The primary objective is to reduce the reliance on operation based on air-fed suits and enhance worker safety by introducing remote equipment and a new floor-reinforcing panel. Another objective is to suppress waste generation by reusing all equipment on multiple occasions which is achieved by developing a containment system that have a large open port with a pallet for the storage and reuse of equipment for successive operations. Furthermore, a glove operation compartment is designed and tested for the manual handling of dismantled materials as an additional strategy to reduce work based on air-fed suits and mitigate secondary waste generation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113691"},"PeriodicalIF":1.9,"publicationDate":"2024-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701234","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 : 2024-11-23DOI: 10.1016/j.nucengdes.2024.113726
Wen-Yu Wang , Bo-Shuan You
Dry storage of spent nuclear fuel (SNF) and the cooling of fuel assemblies are essential for the nuclear industry. SNF will be stored in concrete dry casks and an indoor dry storage facility at the Chinshan Nuclear Power Plant. In this study, Ansys Fluent is used to simulate the symmetric boundaries of thermal cases and the ventilation characteristics of eight vertical dry storage casks for SNF as cylinders in the studied dry storage facility. Incompressible ideal gas is adopted in the simulation, and the low Reynolds k-ε turbulence model is used. The following parameters are analyzed in this study: (i) different heat loads, (ii) different arrangements, (iii) accident conditions, and (iv) installation of an active ventilation system at the outlet position of the dry storage systems (DSSs) in the facility. The results show that the DSS heat load, half-blockages, and arrangements significantly influence the temperature distribution in the facility. Installing an active ventilation system at the outlet position affects the staff and significantly decreases the temperature distribution in the facility. It is recommended that the rotational fan speed be set at 600 to 750 rpm for future designs. The results provide future design guidelines for indoor dry storage facilities.
{"title":"Thermal assessments for design and active ventilation system of indoor dry storage facility of Chinshan Nuclear Power Plant","authors":"Wen-Yu Wang , Bo-Shuan You","doi":"10.1016/j.nucengdes.2024.113726","DOIUrl":"10.1016/j.nucengdes.2024.113726","url":null,"abstract":"<div><div>Dry storage of spent nuclear fuel (SNF) and the cooling of fuel assemblies are essential for the nuclear industry. SNF will be stored in concrete dry casks and an indoor dry storage facility at the Chinshan Nuclear Power Plant. In this study, Ansys Fluent is used to simulate the symmetric boundaries of thermal cases and the ventilation characteristics of eight vertical dry storage casks for SNF as cylinders in the studied dry storage facility. Incompressible ideal gas is adopted in the simulation, and the low Reynolds k-ε turbulence model is used. The following parameters are analyzed in this study: (i) different heat loads, (ii) different arrangements, (iii) accident conditions, and (iv) installation of an active ventilation system at the outlet position of the dry storage systems (DSSs) in the facility. The results show that the DSS heat load, half-blockages, and arrangements significantly influence the temperature distribution in the facility. Installing an active ventilation system at the outlet position affects the staff and significantly decreases the temperature distribution in the facility. It is recommended that the rotational fan speed be set at 600 to 750 rpm for future designs. The results provide future design guidelines for indoor dry storage facilities.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113726"},"PeriodicalIF":1.9,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701232","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 : 2024-11-23DOI: 10.1016/j.nucengdes.2024.113662
Landon Brockmeyer , Nadish Saini , Adrian Tentner , Jun Fang , Elia Merzari
The research literature on Computational Fluid Dynamics (CFD) of coolant flow through rod bundles with spacer-grids and mixing vanes is replete, ranging from high fidelity Large Eddy Simulation (LES)/Direct Numerical Simulation (DNS) simulations to Reynolds-Averaged Navier–Stokes (RANS) modeled studies. The mixing of flow between subchannels and the pressure drop through the bundle are fundamental quantities useful for comparing and evaluating CFD methods. Less commonly observed and compared are the forces exerted onto the structure by the fluid. The present study seeks to evaluate the use of RANS simulations for predicting the structural response to fluid flow. Wall resolved RANS simulations are benchmarked against LES simulations of fluid flow at a Reynolds number of 15,000 through a 3 × 3 fuel rod bundle with a simple spacer grid. Velocity line-plots are compared showing good agreement between RANS and LES results, ascertaining that the former is capable of capturing the essential time-averaged velocity profile. Additionally, the distribution of forces on the spacer grid and fuel rods are collected as a function of time and space. The RANS methods are evaluated using the frequency and magnitude of the fluctuating forces on various portions of the structure as compared to LES. The power spectral density evaluation of the models reveal underprediction of force amplitude on the rod walls by RANS and also discrepancy in the prediction of high frequency spectra, especially in the immediate vicinity of spacer-grid structure, which may be attributed to the lack of random turbulence fluctuation or insufficient modeling of small-scale eddies in RANS simulation.
从高保真大涡模拟（LES）/直接数值模拟（DNS）模拟到雷诺平均纳维-斯托克斯（RANS）模型研究，有关冷却剂流经带间隔栅和混合叶片的棒束的计算流体动力学（CFD）研究文献非常丰富。子通道之间的流动混合和通过管束的压降是用于比较和评估 CFD 方法的基本量。流体对结构施加的作用力较少被观察和比较。本研究旨在评估使用 RANS 模拟预测结构对流体流动的响应。壁面解析 RANS 模拟与雷诺数为 15,000 的流体流经 3 × 3 燃料棒束的 LES 模拟进行了对比，燃料棒束采用简单的间隔网格。速度线图比较显示，RANS 和 LES 的结果非常一致，证明前者能够捕捉到重要的时间平均速度曲线。此外，还收集了间隔网格和燃料棒上的力分布作为时间和空间的函数。与 LES 相比，RANS 方法利用结构各部分受力波动的频率和大小进行评估。对模型的功率谱密度评估显示，RANS 对燃料棒壁上的力振幅预测不足，对高频谱的预测也存在差异，特别是在间隔栅结构附近，这可能归因于 RANS 模拟中缺乏随机湍流波动或对小尺度涡流的建模不足。
{"title":"Evaluation of RANS vs. LES simulation of fluid flow through 3 × 3 rod bundle with a simple spacer grid as a precursor to coupled fluid–structure interaction simulations","authors":"Landon Brockmeyer , Nadish Saini , Adrian Tentner , Jun Fang , Elia Merzari","doi":"10.1016/j.nucengdes.2024.113662","DOIUrl":"10.1016/j.nucengdes.2024.113662","url":null,"abstract":"<div><div>The research literature on Computational Fluid Dynamics (CFD) of coolant flow through rod bundles with spacer-grids and mixing vanes is replete, ranging from high fidelity Large Eddy Simulation (LES)/Direct Numerical Simulation (DNS) simulations to Reynolds-Averaged Navier–Stokes (RANS) modeled studies. The mixing of flow between subchannels and the pressure drop through the bundle are fundamental quantities useful for comparing and evaluating CFD methods. Less commonly observed and compared are the forces exerted onto the structure by the fluid. The present study seeks to evaluate the use of RANS simulations for predicting the structural response to fluid flow. Wall resolved RANS simulations are benchmarked against LES simulations of fluid flow at a Reynolds number of 15,000 through a 3 × 3 fuel rod bundle with a simple spacer grid. Velocity line-plots are compared showing good agreement between RANS and LES results, ascertaining that the former is capable of capturing the essential time-averaged velocity profile. Additionally, the distribution of forces on the spacer grid and fuel rods are collected as a function of time and space. The RANS methods are evaluated using the frequency and magnitude of the fluctuating forces on various portions of the structure as compared to LES. The power spectral density evaluation of the models reveal underprediction of force amplitude on the rod walls by RANS and also discrepancy in the prediction of high frequency spectra, especially in the immediate vicinity of spacer-grid structure, which may be attributed to the lack of random turbulence fluctuation or insufficient modeling of small-scale eddies in RANS simulation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113662"},"PeriodicalIF":1.9,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142700654","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}
With recent developments in nuclear technology, the safety versus nuclear radiations with negative environmental influences is of great importance. Heavyweight concrete (HWC) is an effective absorbent material, capable of providing adequate shielding versus nuclear radiations because of its acceptable structural characteristics. However, the role of aggregate type in the shielding and fracture characteristics of HWC has not been explored comprehensively. On the other hand, nanosilica is one of the reactive pozzolans which is employed for improvement of concrete properties. Thus, in this investigation, the influences of aggregate type (magnetite and hematite) and nanosilica on the mechanical, fracture and shielding features of HWC were studied. Four different cement replacements by nanosilica (0, 2, 4 and 6 %) were used to evaluate its influence on the properties of HWC. The results depicted that the fracture energies increase 18.8 and 16.8 % for heavyweight magnetite and hematite concretes with increasing nanosilica up to 6 wt% (wt.%) of cement, respectively. Furthermore, characteristic length declines from 385.6 to 364.8 mm and from 562.6 to 522.9 mm for heavyweight magnetite and hematite concretes with increasing nanosilica up to 6 wt% of cement, respectively. The results also showed that application of magnetite aggregates in HWC can more effectively shield against nuclear radiations than hematite ones which this issue becomes more obvious with increasing nanosilica content to 4 wt% of cement.
{"title":"Effects of nanosilica and aggregate type on the mechanical, fracture and shielding features of heavyweight concrete","authors":"Mohsen Ghorbani , Morteza Biklaryan , Morteza Hosseinali Beygi , Omid Lotfi-Omran","doi":"10.1016/j.nucengdes.2024.113713","DOIUrl":"10.1016/j.nucengdes.2024.113713","url":null,"abstract":"<div><div>With recent developments in nuclear technology, the safety versus nuclear radiations with negative environmental influences is of great importance. Heavyweight concrete (HWC) is an effective absorbent material, capable of providing adequate shielding versus nuclear radiations because of its acceptable structural characteristics. However, the role of aggregate type in the shielding and fracture characteristics of HWC has not been explored comprehensively. On the other hand, nanosilica is one of the reactive pozzolans which is employed for improvement of concrete properties. Thus, in this investigation, the influences of aggregate type (magnetite and hematite) and nanosilica on the mechanical, fracture and shielding features of HWC were studied. Four different cement replacements by nanosilica (0, 2, 4 and 6 %) were used to evaluate its influence on the properties of HWC. The results depicted that the fracture energies increase 18.8 and 16.8 % for heavyweight magnetite and hematite concretes with increasing nanosilica up to 6 wt% (wt.%) of cement, respectively. Furthermore, characteristic length declines from 385.6 to 364.8 mm and from 562.6 to 522.9 mm for heavyweight magnetite and hematite concretes with increasing nanosilica up to 6 wt% of cement, respectively. The results also showed that application of magnetite aggregates in HWC can more effectively shield against nuclear radiations than hematite ones which this issue becomes more obvious with increasing nanosilica content to 4 wt% of cement.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113713"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701230","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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