The MORE framework, developed by the Nuclear Power Institute of China (NPIC), is a new coupling framework that integrates multiple powerful functionalities for handling input and output data from simulation codes. SHARK, also developed by NPIC, is an advanced whole core transport code that incorporates constructive solid geometry (CSG), the subgroup method, and the 2D/1D transport method. TH1D, another code developed by NPIC, is a single-channel thermal–hydraulic code that utilizes the single-channel model and one-dimensional heat conduction equation. Both SHARK and TH1D are encapsulated within the MORE framework. The coupling system generates interfaces for coupling which are controlled by a supervisor. Finally, validation of the coupling system is conducted using the VERA benchmark, with numerical results demonstrating that the MORE framework is suitable for accurate whole core coupling calculations.
由中国核动力研究院（NPIC）开发的 MORE 框架是一个新的耦合框架，集成了多种强大功能，用于处理来自仿真代码的输入和输出数据。同样由 NPIC 开发的 SHARK 是一种先进的全核输运代码，它集成了构造实体几何（CSG）、子群法和 2D/1D 输运法。TH1D 是 NPIC 开发的另一种代码，是一种单通道热流体力学代码，采用单通道模型和一维热传导方程。SHARK 和 TH1D 都封装在 MORE 框架内。耦合系统生成的耦合界面由监控器控制。最后，使用 VERA 基准对耦合系统进行了验证，数值结果表明 MORE 框架适用于精确的全核心耦合计算。
{"title":"Research and application of the neutronics and thermal–hydraulic coupling based on the MORE framework","authors":"Bo Wang , Zeyi Xie , Dayu Huang , Wenbo Zhao , Hongbo Zhang , Zhang Chen , Wei Zeng , Wenbin Wu","doi":"10.1016/j.anucene.2024.111067","DOIUrl":"10.1016/j.anucene.2024.111067","url":null,"abstract":"<div><div>The MORE framework, developed by the Nuclear Power Institute of China (NPIC), is a new coupling framework that integrates multiple powerful functionalities for handling input and output data from simulation codes. SHARK, also developed by NPIC, is an advanced whole core transport code that incorporates constructive solid geometry (CSG), the subgroup method, and the 2D/1D transport method. TH1D, another code developed by NPIC, is a single-channel thermal–hydraulic code that utilizes the single-channel model and one-dimensional heat conduction equation. Both SHARK and TH1D are encapsulated within the MORE framework. The coupling system generates interfaces for coupling which are controlled by a supervisor. Finally, validation of the coupling system is conducted using the VERA benchmark, with numerical results demonstrating that the MORE framework is suitable for accurate whole core coupling calculations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111067"},"PeriodicalIF":1.9,"publicationDate":"2024-11-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704033","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}
This study focused on synthesizing glasses with the formula (70-x-y)B2O3-20BaO-10TiO2-xZnO-yPbO (where x = y = 5, 10, 15, and 20 mol %) using the melt-quenching approach and examined the impact of ZnO and PbO substitution on the mechanical, optical, and radiation shielding properties of these glasses. The mechanical moduli and micro-hardness decreased with higher ZnO and PbO concentrations, indicating reduced rigidity and elasticity. Optical analyses demonstrated that as ZnO and PbO contents increased, the absorption edge shifted to longer wavelengths, and the optical band gap (Eg) decreased. Furthermore, Urbach energy (EU) values increased from 0.212 to 0.431 eV, indicating a higher degree of structural disorder. Radiation shielding studies revealed that glasses with higher concentrations of ZnO and PbO demonstrated enhanced shielding capabilities. Among the glasses prepared, the sample with the composition 30B2O3 + 20BaO + 10TiO2 + 20ZnO + 20PbO exhibited the best performance, characterized by a high mass attenuation coefficient (MAC) and linear attenuation coefficient (LAC).
本研究的重点是采用熔淬法合成式为（70-x-y）B2O3-20BaO-10TiO2-xZnO-yPbO（其中 x = y = 5、10、15 和 20 mol %）的玻璃，并考察 ZnO 和 PbO 替代对这些玻璃的机械、光学和辐射屏蔽性能的影响。氧化锌和氧化铅浓度越高，机械模量和微硬度越低，表明刚性和弹性降低。光学分析表明，随着氧化锌和氧化铅含量的增加，吸收边缘向更长的波长移动，光带隙（Eg）减小。此外，厄巴赫能（EU）值从 0.212 eV 增加到 0.431 eV，表明结构紊乱程度更高。辐射屏蔽研究表明，氧化锌和氧化铅浓度较高的玻璃具有更强的屏蔽能力。在制备的玻璃中，成分为 30B2O3 + 20BaO + 10TiO2 + 20ZnO + 20PbO 的样品性能最佳，具有较高的质量衰减系数（MAC）和线性衰减系数（LAC）。
{"title":"Tailoring glass characteristics: Unveiling the impact of PbO and ZnO in Titanium-Barium borate glasses for advanced radiation protection","authors":"Jaber Alyami , Yas Al-Hadeethi , Othman A. Fallatah , Shrikant Biradar , M.I. Sayyed , Fahad Almutairi","doi":"10.1016/j.anucene.2024.111069","DOIUrl":"10.1016/j.anucene.2024.111069","url":null,"abstract":"<div><div>This study focused on synthesizing glasses with the formula (70-x-y)B<sub>2</sub>O<sub>3</sub>-20BaO-10TiO<sub>2</sub>-xZnO-yPbO (where x = y = 5, 10, 15, and 20 mol %) using the melt-quenching approach and examined the impact of ZnO and PbO substitution on the mechanical, optical, and radiation shielding properties of these glasses. The mechanical moduli and micro-hardness decreased with higher ZnO and PbO concentrations, indicating reduced rigidity and elasticity. Optical analyses demonstrated that as ZnO and PbO contents increased, the absorption edge shifted to longer wavelengths, and the optical band gap (E<sub>g</sub>) decreased. Furthermore, Urbach energy (E<sub>U</sub>) values increased from 0.212 to 0.431 eV, indicating a higher degree of structural disorder. Radiation shielding studies revealed that glasses with higher concentrations of ZnO and PbO demonstrated enhanced shielding capabilities. Among the glasses prepared, the sample with the composition 30B<sub>2</sub>O<sub>3</sub> + 20BaO + 10TiO<sub>2</sub> + 20ZnO + 20PbO exhibited the best performance, characterized by a high mass attenuation coefficient (MAC) and linear attenuation coefficient (LAC).</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111069"},"PeriodicalIF":1.9,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704032","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.anucene.2024.111026
Byoung-Uhn Bae, Jae Bong Lee, Yu-Sun Park, Seok Cho, Kyoung-Ho Kang
Considering the importance of the pressure build-up depending on the mass / energy (M/E) release from a reactor coolant system (RCS), the ATLAS-CUBE integral effect test facility was utilized to simulate the thermal–hydraulic interaction between a RCS and a containment. To investigate the effect of the break size on the pressure / temperature (P/T) build-up in a containment, this study focused on the integral effect tests for an intermediate-break loss-of-coolant accident (IBLOCA) and a small-break loss-of-coolant accident (SBLOCA). As the test results, the decrease of the coolant water level in the RCS according to the cold leg break induced the core heat-up and the reactor core was cooled down after the safety injection to the RCS. The P/T transient of a containment could be highly dependent not only on the break size, but also on the two-phase flow characteristics and the initial temperature of the steam-gas mixture in a containment.
{"title":"Investigation of the RCS-containment integral effect test on intermediate and small break loss-of-coolant accident (LOCA) transients","authors":"Byoung-Uhn Bae, Jae Bong Lee, Yu-Sun Park, Seok Cho, Kyoung-Ho Kang","doi":"10.1016/j.anucene.2024.111026","DOIUrl":"10.1016/j.anucene.2024.111026","url":null,"abstract":"<div><div>Considering the importance of the pressure build-up depending on the mass / energy (M/E) release from a reactor coolant system (RCS), the ATLAS-CUBE integral effect test facility was utilized to simulate the thermal–hydraulic interaction between a RCS and a containment. To investigate the effect of the break size on the pressure / temperature (P/T) build-up in a containment, this study focused on the integral effect tests for an intermediate-break loss-of-coolant accident (IBLOCA) and a small-break loss-of-coolant accident (SBLOCA). As the test results, the decrease of the coolant water level in the RCS according to the cold leg break induced the core heat-up and the reactor core was cooled down after the safety injection to the RCS. The P/T transient of a containment could be highly dependent not only on the break size, but also on the two-phase flow characteristics and the initial temperature of the steam-gas mixture in a containment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111026"},"PeriodicalIF":1.9,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704030","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 : 2024-11-23DOI: 10.1016/j.anucene.2024.111074
Sohail Ahmad Raza, Liangzhi Cao, Yongping Wang, Yuxuan Wu, Haoyong Li, M. Hashim
The safety of High-Temperature Gas-cooled Reactors (HTGRs) critically depends on understanding the radionuclide inventory and Fission Products (FPs) release behavior, which are fundamental for radiological protection and source term determination in reactor licensing. This study presents a novel method that combines well-established codes (ORIGEN2.2, NECP–MCX, V.S.O.P. (99/11), and STACY) to perform coupled calculations for neutronics, thermal hydraulics, fuel depletion, and fission product releases. An elaborate simulation code, Fission Products Inventory and Release Rate Calculation System (FIRCS) has been developed to track several fictitious tracer pebbles across a user-defined grid. The concept of mock tracers is introduced for equilibrium core and release scenarios. Neutron flux and fuel temperature distributions are derived from the Multiphysics code VSOP. ORIGEN2.2 then simulates flux irradiation at each grid point, utilizing burnup-dependent neutron cross-section libraries generated by NECP–MCX for each core pass. The code tracks radionuclides, temperatures, and Particle Failure Fraction (PFF) for the entire flow history of each tracer. This data is used to calculate release rates for individual tracers in STACY. In HTGR cyclic simulation, these tracers are sequentially introduced into the core with each cycle and a recirculation matrix is computed based on the quantity, pass number, and position of tracers in the core. The matrix is used to retrieve the Concentration and Release Rate (CRR) of radionuclides from these tracers which is then utilized to calculate CRR for the entire core. The estimate converges towards accurate estimates as the number of tracers increases. Thermal decay power, discharge inventory, and photon emission spectra are also calculated for spent fuel. Over a period of 50 days, the accumulated decay power for 40,000 spent fuel pebbles is determined to be 27.4 kW. This work delves deeper into the methodological details and its first application to a 250 MW(t) HTR-PM design. Results are presented for the equilibrium core, including radionuclide inventory and release rates of key fission products. Iodine-131, Cesium-137, Strontium-90, and Silver-110 m have activities of 2.5 × 1017 Bq, 2 × 1016 Bq, 1.6 × 1016 Bq, and 3.5 × 1014 Bq, respectively. Among these radionuclides, Iodine-131 exhibits the highest release rate, followed by Cesium-137, Silver-110 m, and Strontium-90. The calculations in this study have been validated against published data, demonstrating the reliability of the results presented in this work. The application of this methodology to a 250 MW(t) HTR-PM design demonstrates its potential for informing future core design decisions and safety assessments in HTGR development.
{"title":"Development of radionuclide inventory and fission product release calculation model and its application to HTR-PM","authors":"Sohail Ahmad Raza, Liangzhi Cao, Yongping Wang, Yuxuan Wu, Haoyong Li, M. Hashim","doi":"10.1016/j.anucene.2024.111074","DOIUrl":"10.1016/j.anucene.2024.111074","url":null,"abstract":"<div><div>The safety of High-Temperature Gas-cooled Reactors (HTGRs) critically depends on understanding the radionuclide inventory and Fission Products (FPs) release behavior, which are fundamental for radiological protection and source term determination in reactor licensing. This study presents a novel method that combines well-established codes (ORIGEN2.2, NECP–MCX, V.S.O.P. (99/11), and STACY) to perform coupled calculations for neutronics, thermal hydraulics, fuel depletion, and fission product releases. An elaborate simulation code, <strong>F</strong>ission Products <strong>I</strong>nventory and <strong>R</strong>elease Rate <strong>C</strong>alculation <strong>S</strong>ystem (FIRCS) has been developed to track several fictitious tracer pebbles across a user-defined grid. The concept of mock tracers is introduced for equilibrium core and release scenarios. Neutron flux and fuel temperature distributions are derived from the Multiphysics code VSOP. ORIGEN2.2 then simulates flux irradiation at each grid point, utilizing burnup-dependent neutron cross-section libraries generated by NECP–MCX for each core pass. The code tracks radionuclides, temperatures, and Particle Failure Fraction (PFF) for the entire flow history of each tracer. This data is used to calculate release rates for individual tracers in STACY. In HTGR cyclic simulation, these tracers are sequentially introduced into the core with each cycle and a recirculation matrix is computed based on the quantity, pass number, and position of tracers in the core. The matrix is used to retrieve the Concentration and Release Rate (CRR) of radionuclides from these tracers which is then utilized to calculate CRR for the entire core. The estimate converges towards accurate estimates as the number of tracers increases. Thermal decay power, discharge inventory, and photon emission spectra are also calculated for spent fuel. Over a period of 50 days, the accumulated decay power for 40,000 spent fuel pebbles is determined to be 27.4 kW. This work delves deeper into the methodological details and its first application to a 250 MW(t) HTR-PM design. Results are presented for the equilibrium core, including radionuclide inventory and release rates of key fission products. Iodine-131, Cesium-137, Strontium-90, and Silver-110 m have activities of 2.5 × 10<sup>17</sup> Bq, 2 × 10<sup>16</sup> Bq, 1.6 × 10<sup>16</sup> Bq, and 3.5 × 10<sup>14</sup> Bq, respectively. Among these radionuclides, Iodine-131 exhibits the highest release rate, followed by Cesium-137, Silver-110 m, and Strontium-90. The calculations in this study have been validated against published data, demonstrating the reliability of the results presented in this work. The application of this methodology to a 250 MW(t) HTR-PM design demonstrates its potential for informing future core design decisions and safety assessments in HTGR development.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111074"},"PeriodicalIF":1.9,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704031","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.anucene.2024.111046
Liliane Basso Barichello , Imre Pázsit
The theoretical basis of multiplicity counting of nuclear safeguards lies in the calculation of the factorial moments of the number of neutrons emitted from the item. While the traditional method to derive these moments uses the so-called point model in which the spatial transport of neutrons in the item is neglected, the theoretical framework has recently been re-derived in a one-speed transport model, which is inherently of the backward (adjoint) type. The arising integral equations for the moments were solved numerically with a collision number type (iterated kernel or Neumann-series) expansion. In this paper, we show that effective methods of analytical character, originally developed for direct (forward-type) transport problems, can be associated with the solution of the adjoint-type moment equations. The theory is described, and quantitative results are given for selected representative cases. The accuracy and computational speed of the method is investigated and compared favourably with those of the collision number expansion method. The quantitative results also lend some new insight into the properties of statistics of the multiplicative process for the exiting neutrons.
{"title":"Calculation of the multiplicity moments in nuclear safeguards with forward transport theory","authors":"Liliane Basso Barichello , Imre Pázsit","doi":"10.1016/j.anucene.2024.111046","DOIUrl":"10.1016/j.anucene.2024.111046","url":null,"abstract":"<div><div>The theoretical basis of multiplicity counting of nuclear safeguards lies in the calculation of the factorial moments of the number of neutrons emitted from the item. While the traditional method to derive these moments uses the so-called point model in which the spatial transport of neutrons in the item is neglected, the theoretical framework has recently been re-derived in a one-speed transport model, which is inherently of the backward (adjoint) type. The arising integral equations for the moments were solved numerically with a collision number type (iterated kernel or Neumann-series) expansion. In this paper, we show that effective methods of analytical character, originally developed for direct (forward-type) transport problems, can be associated with the solution of the adjoint-type moment equations. The theory is described, and quantitative results are given for selected representative cases. The accuracy and computational speed of the method is investigated and compared favourably with those of the collision number expansion method. The quantitative results also lend some new insight into the properties of statistics of the multiplicative process for the exiting neutrons.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111046"},"PeriodicalIF":1.9,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704029","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 : 2024-11-22DOI: 10.1016/j.anucene.2024.111071
Ayhan Kara, Emil Mammadzada
This study investigates the impact of different fuel cladding materials on the performance and safety of VVER-1200 reactors using the Serpent 2 Monte Carlo code. Cladding materials evaluated include Zircaloy, 304SS, 310SS, FeCrAl, APMT, TiC, ZrC, and SiC. Key parameters assessed are fuel performance, neutronic behavior, infinite multiplication factor (kinf), and radioactive fission product levels. Results indicate that Zircaloy, ZrC, and SiC claddings retain criticality longer (kinf > 1) with favorable neutron flux and fission neutron production. TiC, however, loses criticality early and generates high neutron poisons and fission products. Steel alloys (304SS, 310SS), APMT, and FeCrAl demonstrate moderate performance affecting reactor criticality and neutron flux. Overall, Zircaloy is identified as the most effective cladding, balancing criticality, minimizing plutonium buildup, and reducing radioactive fission products, with ZrC and SiC as close competitors.
本研究使用 Serpent 2 Monte Carlo 代码研究了不同燃料包壳材料对 VVER-1200 反应堆性能和安全性的影响。评估的包层材料包括锆合金、304SS、310SS、铁铬铝、APMT、TiC、ZrC 和 SiC。评估的主要参数包括燃料性能、中子行为、无限倍增因子（kinf）和放射性裂变产物水平。结果表明，锆合金、碳化锆和碳化硅包壳保持临界状态的时间更长（kinf >1），中子通量和裂变中子产生量也更多。然而，TiC 很早就失去临界状态，并产生大量中子毒物和裂变产物。钢合金（304SS、310SS）、APMT 和 FeCrAl 在影响反应堆临界度和中子通量方面的性能适中。总体而言，锆合金被认为是最有效的包层，它能平衡临界状态、最大限度地减少钚的积累并减少放射性裂变产物，而 ZrC 和 SiC 则是最接近的竞争对手。
{"title":"The effects of different fuel cladding materials on neutronic behavior and fuel depletion performance in the VVER-1200 reactor","authors":"Ayhan Kara, Emil Mammadzada","doi":"10.1016/j.anucene.2024.111071","DOIUrl":"10.1016/j.anucene.2024.111071","url":null,"abstract":"<div><div>This study investigates the impact of different fuel cladding materials on the performance and safety of VVER-1200 reactors using the Serpent 2 Monte Carlo code. Cladding materials evaluated include Zircaloy, 304SS, 310SS, FeCrAl, APMT, TiC, ZrC, and SiC. Key parameters assessed are fuel performance, neutronic behavior, infinite multiplication factor (kinf), and radioactive fission product levels. Results indicate that Zircaloy, ZrC, and SiC claddings retain criticality longer (kinf > 1) with favorable neutron flux and fission neutron production. TiC, however, loses criticality early and generates high neutron poisons and fission products. Steel alloys (304SS, 310SS), APMT, and FeCrAl demonstrate moderate performance affecting reactor criticality and neutron flux. Overall, Zircaloy is identified as the most effective cladding, balancing criticality, minimizing plutonium buildup, and reducing radioactive fission products, with ZrC and SiC as close competitors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111071"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704102","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 : 2024-11-22DOI: 10.1016/j.anucene.2024.111061
R. Dagan , T. König , M. Herm , F. Alvarez , E. Dorval , S. Häkkinen , E. Vlassopoulos , A. Shama , A. Smaizys , P. Schillebeeckx
A characterization of the spent nuclear fuel (SNF) for its radionuclide (RN) inventory is vital for various back-end stages of the nuclear fuel cycle. It concerns both the fuel and the metallic (i.e., cladding and structural material) components of the spent fuel assemblies, where different calculation approaches and methods should be deployed for their characterization. This study concentrates on fuel traces and other impurities within the cladding. During the operating cycles, the Zircaloy cladding is exposed to a considerable amount of irradiation. The impact of the exposure should be checked to assure the integrity of the cladding and thus the safety of the stored spent fuel. Within the work package “Spent Nuclear Fuel Characterization and Evolution until Disposal” (SFC) of the EURAD project, dedicated samples were produced, irradiated and the radionuclide inventory of the cladding was analysed and compared. In parallel a blind test was performed, in which different partners used different codes to simulate the irradiation quantity. The blind test showed good agreement between most of the codes, in particular in view of the small amount of the evolved fuel traces. Furthermore, the presence of actinides, caused by precipitation of uranium on the inner surface of the cladding during manufacturing, was found to be negligible in comparison to precipitation of traces of fuel pellets on the cladding during operation. The good agreement between the simulating codes enables to depict further the initial amount of alloying elements of the cladding material itself in a better manner. In particular specific isotopes of cobalt, nickel and iron, which are directly connected to the unique properties of each cladding material can be better identified based on the accurate measuring techniques used in this study.
{"title":"Investigation of nuclide inventory of cladding material irradiated in the Goesgen PWR core","authors":"R. Dagan , T. König , M. Herm , F. Alvarez , E. Dorval , S. Häkkinen , E. Vlassopoulos , A. Shama , A. Smaizys , P. Schillebeeckx","doi":"10.1016/j.anucene.2024.111061","DOIUrl":"10.1016/j.anucene.2024.111061","url":null,"abstract":"<div><div>A characterization of the spent nuclear fuel (SNF) for its radionuclide (RN) inventory is vital for various back-end stages of the nuclear fuel cycle. It concerns both the fuel and the metallic (i.e., cladding and structural material) components of the spent fuel assemblies, where different calculation approaches and methods should be deployed for their characterization. This study concentrates on fuel traces and other impurities within the cladding. During the operating cycles, the Zircaloy cladding is exposed to a considerable amount of irradiation. The impact of the exposure should be checked to assure the integrity of the cladding and thus the safety of the stored spent fuel. Within the work package “Spent Nuclear Fuel Characterization and Evolution until Disposal” (SFC) of the EURAD project, dedicated samples were produced, irradiated and the radionuclide inventory of the cladding was analysed and compared. In parallel a blind test was performed, in which different partners used different codes to simulate the irradiation quantity. The blind test showed good agreement between most of the codes, in particular in view of the small amount of the evolved fuel traces. Furthermore, the presence of actinides, caused by precipitation of uranium on the inner surface of the cladding during manufacturing, was found to be negligible in comparison to precipitation of traces of fuel pellets on the cladding during operation. The good agreement between the simulating codes enables to depict further the initial amount of alloying elements of the cladding material itself in a better manner. In particular specific isotopes of cobalt, nickel and iron, which are directly connected to the unique properties of each cladding material can be better identified based on the accurate measuring techniques used in this study.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111061"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704101","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 : 2024-11-22DOI: 10.1016/j.anucene.2024.111066
Y.M. Chen , D.W. Wu , T.C. Wang , M. Lee
The sequence Core Damage Frequency of a pressurized water reactor has been quantified in three initiating events that are Medium Break Loss of Coolant Accident, Small Break Loss of Coolant Accident and Steam Generator Tube Rupture, following a realistic methodology called risk informed safety margin characteristic. The surrogate plant analyzed in the study is a typical pressurized water reactor. The plant adopted two Westinghouse Three-Loop Pressurized Water Reactors with rated thermal power of 2,830 MWt. The phenomenon identification and ranking table is applied for uncertainty analysis. The mitigation actions as described in plant specific Probabilistic Risk Assessment include cooldown and depressurization, emergency cooldown and depressurization, high head safety injection, high head safety recirculation, low head safety recirculation and Refueling Water Storage Tank replenishment. These mitigation actions are analyzed by thermal hydraulic system analysis code RELAP5-3D to determine the successfulness of the actions. The uncertainties of input parameters of the plant conditions are included, and the time of mitigation action executed is treated as one of the input uncertainties. The results of realistic methodology show a decrease in Core Damage Frequency for all three analyzed events in comparison with conventional methodology. The differences between three initiating events are also discussed.
{"title":"Application of risk informed safety margin characterization to the analysis of a pressurized water reactor","authors":"Y.M. Chen , D.W. Wu , T.C. Wang , M. Lee","doi":"10.1016/j.anucene.2024.111066","DOIUrl":"10.1016/j.anucene.2024.111066","url":null,"abstract":"<div><div>The sequence Core Damage Frequency of a pressurized water reactor has been quantified in three initiating events that are Medium Break Loss of Coolant Accident, Small Break Loss of Coolant Accident and Steam Generator Tube Rupture, following a realistic methodology called risk informed safety margin characteristic. The surrogate plant analyzed in the study is a typical pressurized water reactor. The plant adopted two Westinghouse Three-Loop Pressurized Water Reactors with rated thermal power of 2,830 MWt. The phenomenon identification and ranking table is applied for uncertainty analysis. The mitigation actions as described in plant specific Probabilistic Risk Assessment include cooldown and depressurization, emergency cooldown and depressurization, high head safety injection, high head safety recirculation, low head safety recirculation and Refueling Water Storage Tank replenishment. These mitigation actions are analyzed by thermal hydraulic system analysis code RELAP5-3D to determine the successfulness of the actions. The uncertainties of input parameters of the plant conditions are included, and the time of mitigation action executed is treated as one of the input uncertainties. The results of realistic methodology show a decrease in Core Damage Frequency for all three analyzed events in comparison with conventional methodology. The differences between three initiating events are also discussed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111066"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704035","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 : 2024-11-21DOI: 10.1016/j.anucene.2024.111032
Xue-ying Huang, Hong Xia, Wen-zhe Yin, Yong-kuo Liu
The steam generator (SG) is a critical component of the steam power conversion system in nuclear power plants. The heat transfer tubes of steam generators are susceptible to mechanical and chemical damage due to prolonged exposure to high-temperature, high-pressure environments, and high-radiation media. Timely detection of abnormal states and accurate assessment of the breach degree in the heat transfer tubes are crucial for enhancing the economic and operational safety of nuclear power plants. This study focuses on simulating the normal state of steam generator heat transfer tubes and different degrees of abnormal states using a simulator, while collecting characteristic parameters that can be monitored by sensors. In order to improve the fidelity of the simulated signals to real-world engineering signals, in cases where the breach degree is significant, the reactor undergoes an emergency shutdown, resulting in a smaller amount of effective signal data collected for larger breach degrees. To address these issues, this paper employs the Synthetic Minority Oversampling Technique (SMOTE) to expand the capacity of small sample data. Additionally, to mitigate the impact of high-dimensional feature parameters on subsequent condition monitoring and breach degree assessment, a Denoised AutoEncoder (DAE) is employed to reduce the dimensionality of the feature parameters. The One-Class Support Vector Machine (One-Class SVM) is then utilized to monitor the condition of the steam generator heat transfer tubes. When an abnormality is detected in the heat transfer tubes, a Bi-directional Long Short-Term Memory (Bi-LSTM) model is used to evaluate the magnitude of the tube leakage. The experimental results demonstrate that the developed system achieves a high monitoring accuracy and provides a good assessment of the fault degree.
{"title":"Condition monitoring and breakage assessment of steam generator heat transfer tubes in nuclear power plants","authors":"Xue-ying Huang, Hong Xia, Wen-zhe Yin, Yong-kuo Liu","doi":"10.1016/j.anucene.2024.111032","DOIUrl":"10.1016/j.anucene.2024.111032","url":null,"abstract":"<div><div>The steam generator (SG) is a critical component of the steam power conversion system in nuclear power plants. The heat transfer tubes of steam generators are susceptible to mechanical and chemical damage due to prolonged exposure to high-temperature, high-pressure environments, and high-radiation media. Timely detection of abnormal states and accurate assessment of the breach degree in the heat transfer tubes are crucial for enhancing the economic and operational safety of nuclear power plants. This study focuses on simulating the normal state of steam generator heat transfer tubes and different degrees of abnormal states using a simulator, while collecting characteristic parameters that can be monitored by sensors. In order to improve the fidelity of the simulated signals to real-world engineering signals, in cases where the breach degree is significant, the reactor undergoes an emergency shutdown, resulting in a smaller amount of effective signal data collected for larger breach degrees. To address these issues, this paper employs the Synthetic Minority Oversampling Technique (SMOTE) to expand the capacity of small sample data. Additionally, to mitigate the impact of high-dimensional feature parameters on subsequent condition monitoring and breach degree assessment, a Denoised AutoEncoder (DAE) is employed to reduce the dimensionality of the feature parameters. The One-Class Support Vector Machine (One-Class SVM) is then utilized to monitor the condition of the steam generator heat transfer tubes. When an abnormality is detected in the heat transfer tubes, a Bi-directional Long Short-Term Memory (Bi-LSTM) model is used to evaluate the magnitude of the tube leakage. The experimental results demonstrate that the developed system achieves a high monitoring accuracy and provides a good assessment of the fault degree.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111032"},"PeriodicalIF":1.9,"publicationDate":"2024-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704100","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 : 2024-11-21DOI: 10.1016/j.anucene.2024.111051
Shuai Qin , Qian Zhang , Kai Wang , Dong Huang , Song Li , Yuechao Liang
The Double Heterogeneous (DH) system, where fuel particles are randomly dispersed in the non-fissile matrix, is challenging for the reactor physics calculation. The Sanchez-Pomraning method accurately handles the DH system, but integrating it into existing reactor physics code requires code development. This study adopts the Sanchez-Pomraning coupled Ultra-Fine-Group (SP-UFG) slowing-down calculation to generate the heterogeneous Resonance Integral (RI) for DH system treatment with simple volume homogenization. Fully Ceramic Micro-encapsulated (FCM) fuel pin-cells and plates with varying configurations are calculated for verification. Effective cross-sections (XSs) and keff calculated by the heterogeneous RI are compared with SP-UFG results. Results show that the maximum bias of XSs and keff caused by the XS biases are less than 5% and 200 pcm, respectively. The maximum bias of keff when compared with Monte Carlo calculated results is −213 pcm, demonstrating that only considering the DH effect in the resonance energy region is acceptable.
{"title":"Research on application of heterogeneous resonance Integral for double heterogeneous system","authors":"Shuai Qin , Qian Zhang , Kai Wang , Dong Huang , Song Li , Yuechao Liang","doi":"10.1016/j.anucene.2024.111051","DOIUrl":"10.1016/j.anucene.2024.111051","url":null,"abstract":"<div><div>The Double Heterogeneous (DH) system, where fuel particles are randomly dispersed in the non-fissile matrix, is challenging for the reactor physics calculation. The Sanchez-Pomraning method accurately handles the DH system, but integrating it into existing reactor physics code requires code development. This study adopts the Sanchez-Pomraning coupled Ultra-Fine-Group (SP-UFG) slowing-down calculation to generate the heterogeneous Resonance Integral (RI) for DH system treatment with simple volume homogenization. Fully Ceramic Micro-encapsulated (FCM) fuel pin-cells and plates with varying configurations are calculated for verification. Effective cross-sections (XSs) and <em>k</em><sub>eff</sub> calculated by the heterogeneous RI are compared with SP-UFG results. Results show that the maximum bias of XSs and <em>k</em><sub>eff</sub> caused by the XS biases are less than 5% and 200 pcm, respectively. The maximum bias of <em>k</em><sub>eff</sub> when compared with Monte Carlo calculated results is −213 pcm, demonstrating that only considering the DH effect in the resonance energy region is acceptable.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111051"},"PeriodicalIF":1.9,"publicationDate":"2024-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704098","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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