Pub Date : 2024-08-08DOI: 10.1016/j.net.2024.08.014
Karyoung Choi, Minseok Kim, Kibeom Son, Gyunyoung Heo
Deep geological disposal, which is a permanent method that isolates a storage facility in rocks 200 to 1000 m deep underground, plays a role in safely disposing spent nuclear fuel to prevent excessive radioactive species from leaking out and its safety should be primarily proved. Risk for various radionuclide leakage scenarios should be calculated to be used as a safety indicator. One of the scenarios that significantly affects the safety of repository is the occurrence of external events such as an earthquake scenario. Several occurrences of such external disasters are expected over a long period of time which the safety functions of the deep disposal site must be maintained. Rather than performing a conservative single event evaluation, we expected that performing a more realistic evaluation, such as considering the number of events, is preferrable for design and operation optimization, particularly in an early project phase. This paper suggests a need for assessing multi-event scenarios and the methodology for simulation of several external disasters using earthquakes as an example. The methodology and algorithm of simulating multiple earthquakes and assessing risk distribution under those events at repository site is introduced in GoldSim software, and its results are expected to be used at simulating other event scenarios.
{"title":"Methodology for risk assessment of multi-event scenarios on radioactive waste repository","authors":"Karyoung Choi, Minseok Kim, Kibeom Son, Gyunyoung Heo","doi":"10.1016/j.net.2024.08.014","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.014","url":null,"abstract":"Deep geological disposal, which is a permanent method that isolates a storage facility in rocks 200 to 1000 m deep underground, plays a role in safely disposing spent nuclear fuel to prevent excessive radioactive species from leaking out and its safety should be primarily proved. Risk for various radionuclide leakage scenarios should be calculated to be used as a safety indicator. One of the scenarios that significantly affects the safety of repository is the occurrence of external events such as an earthquake scenario. Several occurrences of such external disasters are expected over a long period of time which the safety functions of the deep disposal site must be maintained. Rather than performing a conservative single event evaluation, we expected that performing a more realistic evaluation, such as considering the number of events, is preferrable for design and operation optimization, particularly in an early project phase. This paper suggests a need for assessing multi-event scenarios and the methodology for simulation of several external disasters using earthquakes as an example. The methodology and algorithm of simulating multiple earthquakes and assessing risk distribution under those events at repository site is introduced in GoldSim software, and its results are expected to be used at simulating other event scenarios.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"7 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207265","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 : 2024-08-08DOI: 10.1016/j.net.2024.08.012
Junhyuk Jang, Minsoo Lee, Gha-Young Kim, Seok Yoon
Copper oxidation at low temperatures below 140 °C and its effects on corrosive behavior in aerobic groundwater are investigated to estimate the intactness of canisters at early stages of disposal. The Cu coupon surface is covered by fine particles that form thin oxide layers after 30 d of oxidation; a thin CuO layer of thickness <100 nm is formed after oxidation at 40 °C; after oxidation at 140 °C, the CuO surface changes to a CuO layer of thickness <500 nm. The thickness of the Cu surface oxidized at 90 °C is between those of the surfaces oxidized at 40 and 140 °C. All Cu coupons exhibit similar current densities ranging from 0.77 to 1.87 μA cm, although the corrosion potential of the Cu coupon layered with CuO is higher than that of the others. Long-term oxidation tests for 406 d reveal no significant changes in the Cu surface at temperatures below 90 °C, indicating no significant change in the electrochemical behavior. The results of this study suggest that the storage of canisters at temperatures below 90 °C has no significant effect on the degradation of canister performance in long-term disposal.
研究了铜在低于 140 °C 的低温下的氧化及其对好氧地下水中腐蚀行为的影响，以估计早期处置阶段滤毒罐的完好性。铜券表面被细颗粒覆盖，氧化 30 d 后形成薄氧化层；40 °C 氧化后形成厚度小于 100 nm 的薄 CuO 层；140 °C 氧化后，CuO 表面变为厚度小于 500 nm 的 CuO 层。在 90 °C 下氧化的铜表面厚度介于在 40 °C 和 140 °C 下氧化的表面厚度之间。所有的铜试样都显示出类似的电流密度，范围在 0.77 到 1.87 μA cm 之间，但层状 CuO 铜试样的腐蚀电位高于其他试样。406 d 的长期氧化测试显示，在低于 90 °C 的温度下，铜表面没有发生明显变化，这表明电化学行为没有发生重大变化。这项研究的结果表明，在低于 90 °C 的温度下储存滤毒罐对滤毒罐在长期处置过程中的性能退化没有明显影响。
{"title":"Corrosive behavior of copper canisters under air and aerobic groundwater at early stages of deep geological disposal","authors":"Junhyuk Jang, Minsoo Lee, Gha-Young Kim, Seok Yoon","doi":"10.1016/j.net.2024.08.012","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.012","url":null,"abstract":"Copper oxidation at low temperatures below 140 °C and its effects on corrosive behavior in aerobic groundwater are investigated to estimate the intactness of canisters at early stages of disposal. The Cu coupon surface is covered by fine particles that form thin oxide layers after 30 d of oxidation; a thin CuO layer of thickness <100 nm is formed after oxidation at 40 °C; after oxidation at 140 °C, the CuO surface changes to a CuO layer of thickness <500 nm. The thickness of the Cu surface oxidized at 90 °C is between those of the surfaces oxidized at 40 and 140 °C. All Cu coupons exhibit similar current densities ranging from 0.77 to 1.87 μA cm, although the corrosion potential of the Cu coupon layered with CuO is higher than that of the others. Long-term oxidation tests for 406 d reveal no significant changes in the Cu surface at temperatures below 90 °C, indicating no significant change in the electrochemical behavior. The results of this study suggest that the storage of canisters at temperatures below 90 °C has no significant effect on the degradation of canister performance in long-term disposal.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"15 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207267","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 : 2024-08-08DOI: 10.1016/j.net.2024.08.015
M.I. Sayyed, K.A. Mahmoud, Taha A. Hanafy
This work presents a study of the effect of replacing lead dioxide and gadolinium (III) oxide with boron trioxide on the physical, mechanical, and radiation shielding properties for the BO-NaO-ZnO-PbO-GdO glass systems. The Archimedes method confirms that the increase in the PbO+GdO concentration within the fabricated glass system in the range from 16 to 22 mol.% increases the fabricated glass samples' density from 4.052 to 4.408 g/cm. Additionally, the Makishima-Mackenzie model was utilized to investigate the influence of PbO+GdO on the mechanical properties of the investigated glass samples. The increase in the substituting of PbO+GdO decreases the fabricated glass samples' mechanical properties and micro-hardness. Furthermore, the Monte Carlo simulation method was applied for the estimation of the impact of PbO+GdO concentration on the fabricated samples' radiation shielding parameters. The increase in the concentration of PbO+GdO with range of (16, 18, 20 and 22) leads to increase the linear attenuation coefficient (LAC) to 8.014–11.517 cm at 0.06 MeV, 0.381–0.423 cm at 0.6 MeV, 0.133–0.149 cm at 5 MeV, and 0.132–0.154 cm at 15 MeV with the same order, respectively. Therefore, the introduction of PbO+GdO concentration enhances the fabricated glass samples' radiation shielding properties to be suitable for γ-ray shielding applications.
{"title":"A new protective glass material against gamma ray: Thorough analysis to determine the impact of adding gadolinium (III) oxide","authors":"M.I. Sayyed, K.A. Mahmoud, Taha A. Hanafy","doi":"10.1016/j.net.2024.08.015","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.015","url":null,"abstract":"This work presents a study of the effect of replacing lead dioxide and gadolinium (III) oxide with boron trioxide on the physical, mechanical, and radiation shielding properties for the BO-NaO-ZnO-PbO-GdO glass systems. The Archimedes method confirms that the increase in the PbO+GdO concentration within the fabricated glass system in the range from 16 to 22 mol.% increases the fabricated glass samples' density from 4.052 to 4.408 g/cm. Additionally, the Makishima-Mackenzie model was utilized to investigate the influence of PbO+GdO on the mechanical properties of the investigated glass samples. The increase in the substituting of PbO+GdO decreases the fabricated glass samples' mechanical properties and micro-hardness. Furthermore, the Monte Carlo simulation method was applied for the estimation of the impact of PbO+GdO concentration on the fabricated samples' radiation shielding parameters. The increase in the concentration of PbO+GdO with range of (16, 18, 20 and 22) leads to increase the linear attenuation coefficient (LAC) to 8.014–11.517 cm at 0.06 MeV, 0.381–0.423 cm at 0.6 MeV, 0.133–0.149 cm at 5 MeV, and 0.132–0.154 cm at 15 MeV with the same order, respectively. Therefore, the introduction of PbO+GdO concentration enhances the fabricated glass samples' radiation shielding properties to be suitable for γ-ray shielding applications.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"23 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207266","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 : 2024-08-06DOI: 10.1016/j.net.2024.08.005
Andrew K. Gillespie, Cuikun Lin, Django Jones, Sandeep Puri, R.V. Duncan
Materials used to study nuclear fusion can retain atmospheric helium unless pretreated before an experiment. Understanding helium outgassing is important for accurate diagnostics in experiments surrounding nuclear fusion. The presence of helium is often cited as the primary evidence that a nuclear reaction has occurred, so it is imperative that known sources of helium are mitigated prior to proceeding with novel nuclear experiments. It is also necessary to ensure hermeticity when transferring gas aliquots from an experiment to a mass spectrometer. In this article, we present studies of helium leak rates in systems used in novel nuclear experiments. We also present studies of helium retention in materials subjected to various heating profiles and atmospheric concentrations. Without pretreatment, 12-inch lengths of both 3/8” diameter tubes and 1/2″ diameter stainless-steel 316 tubing yielded an average areal outgassing amount of 0.64 pmol/cm. If pretreatment is impractical, then the results may be scaled based on the tubing length necessary for constructing custom experimental equipment. It also may reabsorb He from the atmosphere in time. These studies also demonstrate that it is necessary to pretreat most materials prior to performing experiments where the presence of He is being used as an indicator for novel nuclear reactions.
{"title":"Exploring helium retention from technical materials: Development and investigation","authors":"Andrew K. Gillespie, Cuikun Lin, Django Jones, Sandeep Puri, R.V. Duncan","doi":"10.1016/j.net.2024.08.005","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.005","url":null,"abstract":"Materials used to study nuclear fusion can retain atmospheric helium unless pretreated before an experiment. Understanding helium outgassing is important for accurate diagnostics in experiments surrounding nuclear fusion. The presence of helium is often cited as the primary evidence that a nuclear reaction has occurred, so it is imperative that known sources of helium are mitigated prior to proceeding with novel nuclear experiments. It is also necessary to ensure hermeticity when transferring gas aliquots from an experiment to a mass spectrometer. In this article, we present studies of helium leak rates in systems used in novel nuclear experiments. We also present studies of helium retention in materials subjected to various heating profiles and atmospheric concentrations. Without pretreatment, 12-inch lengths of both 3/8” diameter tubes and 1/2″ diameter stainless-steel 316 tubing yielded an average areal outgassing amount of 0.64 pmol/cm. If pretreatment is impractical, then the results may be scaled based on the tubing length necessary for constructing custom experimental equipment. It also may reabsorb He from the atmosphere in time. These studies also demonstrate that it is necessary to pretreat most materials prior to performing experiments where the presence of He is being used as an indicator for novel nuclear reactions.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"15 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207269","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 : 2024-08-06DOI: 10.1016/j.net.2024.08.011
Mumtaz Ali, Ahmed Samour, Suhaib Ahmed Soomro, Waqar Khalid, Turgut Tursoy
Nuclear energy is considered an effective means of enhancing environmental sustainability. Considering this point, this study aims to explore the impact of nuclear energy, financial globalization, technological innovation, and economic growth on ecological sustainability in the top-10 nuclear energy-consuming economies from 1995 to 2020. The load capacity factor is used as a novel proxy for ecological sustainability, explaining how human actions affect ecological sustainability and how nature compensates for human-induced damage. The study employs a novel non-parametric MMQR approach to obtain coefficients across heterogeneous quantiles. The MMQR estimation findings indicate that: (i) nuclear and renewable energy consumption and financial globalization promote environmental sustainability by increasing the LCF; (ii) economic growth degrades ecological sustainability by decreasing the LCF; and (iii) the results from Granger causality suggest a causal link among economic growth, technological innovation, nuclear energy, and LCF. The study recommends that the governments of the top-10 nuclear energy-consuming countries facilitate more investment in green technologies and green energy to achieve environmental sustainability.
{"title":"A step towards a sustainable environment in top-10 nuclear energy consumer countries: The role of financial globalization and nuclear energy","authors":"Mumtaz Ali, Ahmed Samour, Suhaib Ahmed Soomro, Waqar Khalid, Turgut Tursoy","doi":"10.1016/j.net.2024.08.011","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.011","url":null,"abstract":"Nuclear energy is considered an effective means of enhancing environmental sustainability. Considering this point, this study aims to explore the impact of nuclear energy, financial globalization, technological innovation, and economic growth on ecological sustainability in the top-10 nuclear energy-consuming economies from 1995 to 2020. The load capacity factor is used as a novel proxy for ecological sustainability, explaining how human actions affect ecological sustainability and how nature compensates for human-induced damage. The study employs a novel non-parametric MMQR approach to obtain coefficients across heterogeneous quantiles. The MMQR estimation findings indicate that: (i) nuclear and renewable energy consumption and financial globalization promote environmental sustainability by increasing the LCF; (ii) economic growth degrades ecological sustainability by decreasing the LCF; and (iii) the results from Granger causality suggest a causal link among economic growth, technological innovation, nuclear energy, and LCF. The study recommends that the governments of the top-10 nuclear energy-consuming countries facilitate more investment in green technologies and green energy to achieve environmental sustainability.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"37 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207268","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 : 2024-08-05DOI: 10.1016/j.net.2024.08.009
Hyun Chul Lee, Jungjin Bang, Kiwan Jang, Min Beom Heo, Doo Yong Lee, Daewoo Kim, Chang-Ha Lee
After the severe accident at the Fukushima nuclear power plant, there has been heightened interest in the development of severe accident countermeasures. As part of these efforts, a passive filtration system capable of passive operation during power loss is under development.
{"title":"Exothermic characteristics evaluation of the Pt coated α-Al2O3 catalyst in bed for passive airborne radioactive material reduction system under forced flow conditions","authors":"Hyun Chul Lee, Jungjin Bang, Kiwan Jang, Min Beom Heo, Doo Yong Lee, Daewoo Kim, Chang-Ha Lee","doi":"10.1016/j.net.2024.08.009","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.009","url":null,"abstract":"After the severe accident at the Fukushima nuclear power plant, there has been heightened interest in the development of severe accident countermeasures. As part of these efforts, a passive filtration system capable of passive operation during power loss is under development.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"1 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207270","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 : 2024-08-05DOI: 10.1016/j.net.2024.08.007
Qing Zhao, Shuo Meng, Longcheng Liu
In this work, a new Time Domain Random Walk (TDRW) algorithm is proposed to estimate the tracer distribution profile within the rock matrix. The development of the new algorithm stems from the statistical properties of the analytical solution to a single fracture-matrix system, in which the particle position at a certain time is calculated and recorded. With the position of each particle determined, the resulting distribution will then provide an estimate of the tracer distribution profile directly. In addition, the newly developed algorithm can readily be extended to a case of more complicated fracture-matrix system, in which an arbitrary injection boundary condition may also be used. To verify the accuracy and applicability of the new algorithm, three benchmark simulations are made, in which the results of different approaches are found to be identical. Nevertheless, the new algorithm has a higher computational efficiency, due to its lower calculation demand.
{"title":"Tracer transport in fractured porous media: Distribution of tracer concentration within the rock matrix and the implementation of time domain random walk algorithm","authors":"Qing Zhao, Shuo Meng, Longcheng Liu","doi":"10.1016/j.net.2024.08.007","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.007","url":null,"abstract":"In this work, a new Time Domain Random Walk (TDRW) algorithm is proposed to estimate the tracer distribution profile within the rock matrix. The development of the new algorithm stems from the statistical properties of the analytical solution to a single fracture-matrix system, in which the particle position at a certain time is calculated and recorded. With the position of each particle determined, the resulting distribution will then provide an estimate of the tracer distribution profile directly. In addition, the newly developed algorithm can readily be extended to a case of more complicated fracture-matrix system, in which an arbitrary injection boundary condition may also be used. To verify the accuracy and applicability of the new algorithm, three benchmark simulations are made, in which the results of different approaches are found to be identical. Nevertheless, the new algorithm has a higher computational efficiency, due to its lower calculation demand.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"53 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207271","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 : 2024-08-05DOI: 10.1016/j.net.2024.08.010
Hae Min Park, Jong Hyuk Lee, Chiwoong Choi, Kwi-Seok Ha, Byung Hyun You, Jaeseok Heo, Kyung Doo Kim, Sung Won Bae, Seung Wook Lee, Dong Hyuk Lee, Sang Ik Lee, Chan Eok Park, Bub Dong Chung, Kwang Won Seul
Recent safety issues such as cladding oxidation and fuel fragmentation, relocation and dispersal (FFRD) make the loss-of-coolant accident (LOCA) acceptance criteria more difficult to be satisfied. To obtain the adequate safety margin and more economical use of the nuclear power plant, re-classification of LOCAs for Korean operating PWRs is now under consideration to exclude the large break LOCA (LBLOCA) from the design basis accidents (DBAs). Therefore, the intermediate break LOCA (IBLOCA) might become the limiting break size of concern in the LOCA analysis. To accomplish this reform of LOCA classification, an extensive understanding of IBLOCA is crucial, and the applicability of safety analysis code should be confirmed. For this purpose, IBLOCA PIRT was developed. The PIRT panel was organized and the general process of the PIRT development established by Wilson and Boyack (1998) was adopted to develop the IBLOCA PIRT. Based on IBLOCA analyses using the safety and performance analyzing code (SPACE), the PIRT panel defined the temporal phases, systems, components and possible phenomena during an IBLOCA. For all possible phenomena, the relative importance and the knowledge level were determined for each component and each phase via discussions among PIRT panel members. For phenomena having relatively high importance, a code validation matrix was also developed. The results of the IBLOCA PIRT will be used to improve the SPACE code for IBLOCA application and resolve future regulatory issues.
{"title":"Development of a phenomena identification and ranking table (PIRT) for intermediate break loss-of-coolant accident in PWRs","authors":"Hae Min Park, Jong Hyuk Lee, Chiwoong Choi, Kwi-Seok Ha, Byung Hyun You, Jaeseok Heo, Kyung Doo Kim, Sung Won Bae, Seung Wook Lee, Dong Hyuk Lee, Sang Ik Lee, Chan Eok Park, Bub Dong Chung, Kwang Won Seul","doi":"10.1016/j.net.2024.08.010","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.010","url":null,"abstract":"Recent safety issues such as cladding oxidation and fuel fragmentation, relocation and dispersal (FFRD) make the loss-of-coolant accident (LOCA) acceptance criteria more difficult to be satisfied. To obtain the adequate safety margin and more economical use of the nuclear power plant, re-classification of LOCAs for Korean operating PWRs is now under consideration to exclude the large break LOCA (LBLOCA) from the design basis accidents (DBAs). Therefore, the intermediate break LOCA (IBLOCA) might become the limiting break size of concern in the LOCA analysis. To accomplish this reform of LOCA classification, an extensive understanding of IBLOCA is crucial, and the applicability of safety analysis code should be confirmed. For this purpose, IBLOCA PIRT was developed. The PIRT panel was organized and the general process of the PIRT development established by Wilson and Boyack (1998) was adopted to develop the IBLOCA PIRT. Based on IBLOCA analyses using the safety and performance analyzing code (SPACE), the PIRT panel defined the temporal phases, systems, components and possible phenomena during an IBLOCA. For all possible phenomena, the relative importance and the knowledge level were determined for each component and each phase via discussions among PIRT panel members. For phenomena having relatively high importance, a code validation matrix was also developed. The results of the IBLOCA PIRT will be used to improve the SPACE code for IBLOCA application and resolve future regulatory issues.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"31 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207275","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 : 2024-08-05DOI: 10.1016/j.net.2024.08.006
Tao Zhou, Yong Zhu, Shengnan Tang
Optimizing the hydraulic components of nuclear main pump (NMP) and conducting performance verification is crucial. Due to the large size of the real NMP, the strict requirements of the operation and the high test-cost, there are many difficulties in the real test. The mixed flow NMP is taken as the research object, and the CAP1400 NMP is selected as the prototype pump (PP). The model pumps (MPs) with varying scales are established based on the similarity conversion algorithm (SCA). Then, the influence of different scales on the hydraulic performance and internal flow field is investigated and compared. It is demonstrated that the predicted value of head is 4 % higher than the design value at the design operating point, and the maximum efficiency point is close to the design operating point. In the range of full flow conditions, the head, hydraulic efficiency, impeller efficiency, guide vane energy loss, internal flow field, and vorticity distribution of PP and MPs are basically consistent with the trend of flow rate variations. The PP and MPs conform to the SCA. The hydraulic design and performance optimization of NMP are achieved by using the model proportional scaling approach.
{"title":"Scale effect for hydraulic model of a mixed flow nuclear main pump","authors":"Tao Zhou, Yong Zhu, Shengnan Tang","doi":"10.1016/j.net.2024.08.006","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.006","url":null,"abstract":"Optimizing the hydraulic components of nuclear main pump (NMP) and conducting performance verification is crucial. Due to the large size of the real NMP, the strict requirements of the operation and the high test-cost, there are many difficulties in the real test. The mixed flow NMP is taken as the research object, and the CAP1400 NMP is selected as the prototype pump (PP). The model pumps (MPs) with varying scales are established based on the similarity conversion algorithm (SCA). Then, the influence of different scales on the hydraulic performance and internal flow field is investigated and compared. It is demonstrated that the predicted value of head is 4 % higher than the design value at the design operating point, and the maximum efficiency point is close to the design operating point. In the range of full flow conditions, the head, hydraulic efficiency, impeller efficiency, guide vane energy loss, internal flow field, and vorticity distribution of PP and MPs are basically consistent with the trend of flow rate variations. The PP and MPs conform to the SCA. The hydraulic design and performance optimization of NMP are achieved by using the model proportional scaling approach.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"7 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207272","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 : 2024-08-03DOI: 10.1016/j.net.2024.08.004
Maryam Al Huwayz, Aljawhara H. Almuqrin, F.F. Alharbi, M.I. Sayyed, B. Albarzan
This research focuses on the preparation of a new glass system designed specifically for applications in radiation shielding materials. These glasses are based on the general formula (56-x)BO–10TiO–8BaO–27ZnO-(x-1)NdO, where x takes the values of 2, 4, 6 and 8 mol%. For the examination of the designed glasses' radiation attenuation performance, Phy-X software was used, which is a useful approach for predicting the linear attenuation coefficient (LAC), the half value layer (HVL), and effective atomic number. The LAC decreases from 1.489 cm to 0.551 cm for the glass with x = 1 mol%, while the glass with 7 mol% NdO saw a decrease in the LAC from 2.483 cm to 0.718 cm. Introducing NdO increases the glasses' LAC, suggesting enhanced radiation shielding performance. Also, NdO addition influences the HVL within the glasses, with higher content reducing the HVL. At 0.122 MeV, the HVL and tenth value layer (TVL) are 0.456 and 1.546 cm, respectively. At 0.245 MeV, the TVL is about 3.32 times higher than the HVL. The lowest mean free path (MFP) is found at 0.122 MeV, which varies between 0.672 cm for Nd1 and 0.403 cm for Nd4.
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