Annals of Nuclear Energy最新文献

{"title":"Application of data partitioned Kriging algorithm with GPU acceleration in real-time and refined reconstruction of three-dimensional radiation fields","authors":"Ningbiao Xiao,&nbsp;Jinsen Guo,&nbsp;Zijia Kuang,&nbsp;Wei Wang","doi":"10.1016/j.anucene.2024.111047","DOIUrl":"10.1016/j.anucene.2024.111047","url":null,"abstract":"<div><div>Accurately assessing personal radiation doses in real radiation environments like nuclear power plants requires precise and real-time reconstruction of three-dimensional radiation fields. The Kriging algorithm, known for its accuracy in spatial interpolation, provides a promising approach for this task. However, its computational demands can be significant, especially in real-time scenarios. To address this, we enhance the Kriging algorithm with GPU acceleration and data partitioning strategies, enabling efficient and accurate reconstruction of three-dimensional nuclear radiation fields. Using Fluka software for Monte Carlo simulations, we generated a virtual radiation field of dimensions 5 m <span><math><mo>×</mo></math></span> 5 m <span><math><mo>×</mo></math></span> 5 m for a single-source, unshielded scenario, and a field of dimensions 20 m <span><math><mo>×</mo></math></span> 6 m <span><math><mo>×</mo></math></span> 8 m for a multi-source, shielded scenario. Using the simulated data, we compared the prediction accuracy of the improved algorithm with the conventional Kriging algorithm and further explored factors influencing the acceleration ratio of the improved algorithm. The results indicate that the GPU-accelerated and data-partitioned Kriging algorithm achieves nearly identical accuracy compared to the traditional method. In the single-source, unshielded scenario, with more than 343 known (measurement) points and predicting <span><math><mrow><mn>95</mn><mo>×</mo><mn>95</mn><mo>×</mo><mn>95</mn><mo>=</mo></mrow></math></span> 857,375 points, the prediction accuracy remains above 92.25%. In the multi-source, shielded scenario, with more than 8000 known (measurement) points and predicting <span><math><mrow><mn>95</mn><mo>×</mo><mn>95</mn><mo>×</mo><mn>95</mn><mo>=</mo></mrow></math></span> 857,375 points, the prediction accuracy remains above 91.17%. The acceleration performance of the improved algorithm is consistent across both scenarios, with the acceleration ratio increasing as the number of known and predicted points grows, reaching approximately 20 for smaller datasets and up to 93 for larger datasets. Additionally, the acceleration effect of the improved algorithm varies with data partition size, initially increasing and then decreasing as the partition size increases. When the number of known points is 512 and the number of predicted points is 884,736, the optimal partition size lies between 80,000 and 90,000, resulting in a prediction time of only 0.24 s.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111047"},"PeriodicalIF":1.9,"publicationDate":"2024-11-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142720717","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}

{"title":"Evaluation of uranium-233 neutron capture cross section in keV region","authors":"Naohiko Otuka ,&nbsp;Kenichi Tada ,&nbsp;Oscar Cabellos ,&nbsp;Osamu Iwamoto","doi":"10.1016/j.anucene.2024.110977","DOIUrl":"10.1016/j.anucene.2024.110977","url":null,"abstract":"<div><div>The uranium-233 neutron capture cross section between 3 keV and 1 MeV was evaluated with the alpha value recently measured at the Los Alamos National Laboratory LANCE facility and compiled in the EXFOR library. The obtained capture cross section is systematically lower than those in the latest versions of the major general purpose nuclear data libraries, and the reduction from the JENDL-5 library is close to 50% around 20 keV. The newly evaluated cross section was benchmarked against 166 criticality experiments chosen from the ICSBEP handbook by performing Monte Carlo neutron transport calculation with the JENDL-5 library, and slight reduction of the cumulative chi-square value was achieved by adoption of the newly evaluated capture cross section.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 110977"},"PeriodicalIF":1.9,"publicationDate":"2024-11-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142720716","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}

{"title":"Boiling critical characteristics in narrow rectangular channel under local heat flux concentration conditions","authors":"Kepiao Li ,&nbsp;Kui Zhang ,&nbsp;Tianyi Qi ,&nbsp;Ronghua Chen ,&nbsp;Wenxi Tian ,&nbsp;Suizheng Qiu","doi":"10.1016/j.anucene.2024.111077","DOIUrl":"10.1016/j.anucene.2024.111077","url":null,"abstract":"<div><div>For the establishment of predicting method of local concentrated critical heat flux (CHF) in narrow rectangular channels, and also technical support for the preparation of long-life and high-performance dispersive plate fuel, this work simulates Dry-out type CHF under local concentration of heat flux condition by using Eulerian multiphase model and Enhanced near-wall treatment. The distribution of near-wall temperature and void fraction under different heat flux was calculated by Computational Fluid Dynamics (CFD) method. Dry-out CHF was predicted based on highly and quickly wall temperature rise. The influence of system pressure, mass flow rate and inlet subcooling degree were discussed on Dry-out CHF. Compared with uniform heating, the predicted average heat flux on CHF is lower under local heat flux concentration, and location of CHF is related to the position of heat flux concentration area. This work’s research method in Ansys Fluent could be applied to forecast Dry-out CHF in narrow rectangular channel under local concentration of heat flux condition.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111077"},"PeriodicalIF":1.9,"publicationDate":"2024-11-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142720715","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}

{"title":"Numerical simulation of impurity particles migration and deposition within the 5 × 5 fuel assembly under nucleate boiling conditions","authors":"Hongkang Tian ,&nbsp;Tenglong Cong ,&nbsp;Maolong Liu ,&nbsp;Mengke Cai ,&nbsp;Zijian Huang ,&nbsp;Hanyang Gu","doi":"10.1016/j.anucene.2024.111076","DOIUrl":"10.1016/j.anucene.2024.111076","url":null,"abstract":"<div><div>Deposition of corrosion products on the cladding surface can cause many problems in PWRs, such as CRUD-induced power shift, localized corrosion, and increased radioactive risk. To gain a deeper understanding of fouling behavior within the fuel assembly, the migration and deposition of impurity particles were investigated with CFD method, where the Eulerian two-fluid model, species transport model, and fouling deposition model are considered. The relationship between fouling deposition rate and thermal–hydraulic parameters was investigated, and the distribution of fouling thickness along both the axial and circumferential directions was analyzed. The results show that the maximum fouling thickness reaches 44 µm in a refueling cycle. Compared with the inner rods, the corner rods have a higher fouling thickness in single-phase region and a lower fouling thickness in nucleate boiling region. The results obtained in this study could provide a reference for the investigation of deposition behavior within the reactor core.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111076"},"PeriodicalIF":1.9,"publicationDate":"2024-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704090","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}

{"title":"SCALE inventory and reactivity analysis as part of the Hermes 2021 PSAR review","authors":"Friederike Bostelmann ,&nbsp;Benjamin R. Betzler ,&nbsp;Donny Hartanto ,&nbsp;William A. Wieselquist","doi":"10.1016/j.anucene.2024.111063","DOIUrl":"10.1016/j.anucene.2024.111063","url":null,"abstract":"<div><div>The readiness of SCALE for comprehensive studies of pebble-bed reactors has been demonstrated through detailed analysis of a fluoride salt–cooled, high-temperature pebble-bed reactor (PB-FHR). The methods developed for pebble-bed reactor modeling in SCALE, particularly for inventory generation, have proven effective in gaining insights into the reactor physics of this advanced reactor. Excellent agreement with another code package has been observed, further highlighting SCALE’s strong performance.</div><div>The SCALE results supported the US Nuclear Regulatory Commission’s construction permit application review of the <em>Hermes</em> low-power PB-FHR demonstration reactor. A SCALE model of the Hermes reactor was developed at Oak Ridge National Laboratory using information from the Preliminary Safety Analysis Report (PSAR) and supplemented with publicly available data. SCALE reactivity coefficient simulations reproduced PSAR results within 1<span><math><mi>σ</mi></math></span> statistical uncertainties. Sensitivity studies emphasized the importance of graphite specifications for accurate k_eff predictions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111063"},"PeriodicalIF":1.9,"publicationDate":"2024-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704187","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}

{"title":"Investigation on thermal response of high temperature heat pipe under thermal mismatch conditions","authors":"Zhang Jiarui ,&nbsp;Guo Kailun ,&nbsp;Liu Lining ,&nbsp;Xue Meng ,&nbsp;Wang Chenglong ,&nbsp;Tian Wenxi ,&nbsp;Qiu Suizheng ,&nbsp;Su Guanghui ,&nbsp;Chen Chong","doi":"10.1016/j.anucene.2024.111075","DOIUrl":"10.1016/j.anucene.2024.111075","url":null,"abstract":"<div><div>As an efficient and reliable passive heat transfer equipment, high temperature heat pipes (HTHPs) plays an important role in HTHP solid state reactor system. HTHPs often encounter thermal mismatch in reactor applications, which will lead to capillary action, entrainment, and other characteristics, resulting in HTHP failure due to heat transfer limits. In this paper, the thermal response test of liquid metal HTHP under thermal mismatch conditions is studied, and the simulation test platform of HTHP mismatch is designed and built to measure the important physical parameters. The test results show that the heat transfer performance of the HTHP is the best under the condition of horizontal inclination of 60°, and the condenser power is 5.66 kW. This condition is selected as the reference condition for the transient condition test. When the HTHP experienced a transient power increase, the middle of the evaporator section, as the part with the largest heat flux of heating power, appeared a severe overheating phenomenon. The wall temperature of the evaporator section soared from 750 ℃ to 1050 ℃ in a short time, and the heating rate reached 18.9 ℃/min, which affected the smooth operation of the HTHP. When the HTHP encounters an inclination angle transient increase condition, the entrainment phenomenon is more severe, the temperature fluctuation of the condenser section is intensified, and the heat transfer of the HTHP is in an unstable state. When the HTHP encounters an inclination angle transient decrease condition, the temperature rise of the evaporator section is less than 10 ℃, the temperature at the end of the condenser section is increased, and the start-up performance of the HTHP is improved. In summary, this paper obtains the thermal response rule of HTHPs under thermal mismatch conditions through experimental research, which provides support for the safe application of HTHPs for special equipment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111075"},"PeriodicalIF":1.9,"publicationDate":"2024-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704034","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}

{"title":"Research and application of the neutronics and thermal–hydraulic coupling based on the MORE framework","authors":"Bo Wang ,&nbsp;Zeyi Xie ,&nbsp;Dayu Huang ,&nbsp;Wenbo Zhao ,&nbsp;Hongbo Zhang ,&nbsp;Zhang Chen ,&nbsp;Wei Zeng ,&nbsp;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}

{"title":"Tailoring glass characteristics: Unveiling the impact of PbO and ZnO in Titanium-Barium borate glasses for advanced radiation protection","authors":"Jaber Alyami ,&nbsp;Yas Al-Hadeethi ,&nbsp;Othman A. Fallatah ,&nbsp;Shrikant Biradar ,&nbsp;M.I. Sayyed ,&nbsp;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)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 (E_g) decreased. Furthermore, Urbach energy (E_U) 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).</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}

{"title":"Investigation of the RCS-containment integral effect test on intermediate and small break loss-of-coolant accident (LOCA) transients","authors":"Byoung-Uhn Bae,&nbsp;Jae Bong Lee,&nbsp;Yu-Sun Park,&nbsp;Seok Cho,&nbsp;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}

{"title":"Development of radionuclide inventory and fission product release calculation model and its application to HTR-PM","authors":"Sohail Ahmad Raza,&nbsp;Liangzhi Cao,&nbsp;Yongping Wang,&nbsp;Yuxuan Wu,&nbsp;Haoyong Li,&nbsp;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 × 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.</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}