Annals of Nuclear Energy最新文献

{"title":"Built-in physics models and proton-induced nuclear data validation using MCNP, PHITS, and FLUKA – Impact on the shielding design for proton accelerator facilities","authors":"Y. Çelik ,&nbsp;A. Stankovskiy ,&nbsp;H. Iwamoto ,&nbsp;Y. Iwamoto ,&nbsp;G. Van den Eynde","doi":"10.1016/j.anucene.2024.111048","DOIUrl":"10.1016/j.anucene.2024.111048","url":null,"abstract":"<div><div>The MCNP, PHITS, and FLUKA are general-purpose Monte Carlo (MC) radiation transport codes that are widely used for many real-world shielding problems at accelerator facilities around the world. Nuclear interactions are described in these codes by either built-in physics models, or tables with evaluated cross sections and secondary energy-angular distributions, or a combination of both. Over the decades, many code validation efforts have been made, owing to the availability of shielding benchmarks to test the physics models and nuclear data and to verify the accuracy of simulation codes.</div><div>For high beam energy and high beam current accelerator applications, neutron emission through the vacuum pipe along the reverse direction of incident proton beam is an important factor for a shielding design in order to correctly assess the dose rates for workers and the structural materials of the accelerator and handle with the waste activated by the backscattered neutron fluxes. In this work, neutron-production cross sections and thick-target yield predictions from MC codes relying on physics models and nuclear data libraries are benchmarked against the experimental data, in order to assess their accuracy in predicting neutron emission and furthermore to assess the corresponding impact on shielding design.</div><div>The results of this study demonstrate that the nuclear data libraries and physics models, which are not expected to give good results at lower energies (&lt; 150 MeV) but are used anyhow when there is no nuclear data available or above the energy range where the data tables end in the so-called “mix-and-match” strategy, need further improvements. Among the investigated proton induced nuclear data libraries, JENDL-4.0/HE produces the most satisfactory agreement to experimental data for all target materials, but may still benefit from refinement. Concerning the physics models of the codes, FLUKA V4-4.0 has the best performance in terms of reproducibility of the experimental values. It is also shown that all discrepancies between the calculations and the experiments for the energy range &lt; 10 MeV (which is dominant on the dose rates through the shield thickness), are up to factor of two. This might be considered as an acceptable figure as it is equivalent to a normal safety margin (x2) considered in shielding calculations of accelerator facilities around the world.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111048"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658349","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":"3D radiation field reconstruction for multiple unknown radioactive sources based on limited measurements","authors":"Xulin Hu ,&nbsp;Junling Wang ,&nbsp;Jianwen Huo ,&nbsp;Huaifang Zhou ,&nbsp;Li Hu","doi":"10.1016/j.anucene.2024.111053","DOIUrl":"10.1016/j.anucene.2024.111053","url":null,"abstract":"<div><div>In recent years, nuclear energy has played an important role in terms of energy structure optimization and energy security. In order to reduce the radiation exposure of occupational technicians and obtain radiation intensity distribution in the environment, it is essential to reconstruct the three-dimensional (3D) radiation field. However, in some scenes, especially those with multiple radioactive sources, how to accurately reconstruct the 3D radiation field using limited measurements remains a major challenge. This paper explores a novel 3D radiation field reconstruction method based on back-propagation neural network and genetic algorithm to accurately reconstruct the 3D radiation field of multiple radioactive sources with limited measurements. First, the volume of interest is represented as an octree map. Then, the radiation dose distribution of radioactive sources in the octree map is obtained by Monte Carlo (MC) simulation method, and multiple sets of radiation data are collected at a low sampling rate by the random sampling method as the radiation dataset. Further, the radiation dataset is fed into the designed network architecture optimized by genetic algorithm to fit the missing dose rates in the octree map. The feasibility of the proposed method is demonstrated through three representative cases. The experimental results show that in open indoor scenes, the average relative error of the proposed method is less than 2.73% using only 1.625% of measurement data, which is reduced by 29.27% compared with the traditional Gaussian process regression (GPR) method; in indoor scenes with obstacle shielding, the average relative error of the proposed method is less than 3.01%, which is reduced by 30.65% compared to the GPR method. The experimental results reveal the important practicality of our proposed method for 3D radiation field reconstruction tasks with multiple radioactive sources.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111053"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658313","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":"Innovative control mechanism for research and test reactors using mandrel-shaped control rods","authors":"Zhaopeng Zhong ,&nbsp;Mark D. DeHart ,&nbsp;Matthew P. Johnson ,&nbsp;Joseph W. Nielsen","doi":"10.1016/j.anucene.2024.111040","DOIUrl":"10.1016/j.anucene.2024.111040","url":null,"abstract":"<div><div>Research and test reactors have historically played a pivotal role in supporting the initial development of nuclear reactors. They continue to provide essential data for enhancing fuel designs and material knowledge. However, with many such reactors aging and the growing demand for data to bolster advanced reactor development, it is more necessary to research potential design attributes of the next generation of research and test reactors. For test reactors dedicated to fuel and material testing, the design of control mechanisms significantly influences the stabilization of neutron flux levels in irradiation positions while sustaining criticality. This study presents an innovative control mechanism for potential research and test reactor designs. It employs small absorber rods that move in opposite axial directions to maintain axial symmetry of power and neutron flux during burnup cycles. These rods maximize reactivity worth while also offering flexibility to flatten the radial power distribution. An axial translation of the control mechanisms’ absorbers, as compared to the rotational movement of absorbers in control cylinders, also provides a benefit to available excess reactivity and cycle length. This work utilizes a simplified core model of the Advanced Test Reactor to assess the performance of this control mechanism. Compared to the current control system based on rotating control cylinders, the new control mechanism has the potential to enhance, or at least maintain, neutronic performance parameters in this reactor design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111040"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658348","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":"A comparison study between the present air cleaning unit of main control room of AP1000 and its replacement design","authors":"Xiaochen Li ,&nbsp;Jian Li ,&nbsp;Yuan Zhang,&nbsp;Yongguo Li,&nbsp;Kunjun Wang,&nbsp;Jianlu Pei,&nbsp;Xin Li,&nbsp;Xu Shi,&nbsp;Dongan Zhi,&nbsp;Xin Chen,&nbsp;Mang Wang,&nbsp;Jingguo Liu","doi":"10.1016/j.anucene.2024.111060","DOIUrl":"10.1016/j.anucene.2024.111060","url":null,"abstract":"<div><div>The inleakage of the radioactivity caused by ingress/egress could reduce the habitability of the main control room for the operators under the accident conditions. This study used the concentration-decay method to measure the air inleakage caused by ingress/egress, and based on this point, theoretical calculations of the radioactivity in the vestibule and the main control room were conducted to clarify the air cleaning effect. The results showed that: The measured air inleakage induced by each ingress/egress ranged from 0.273 m³ to 1.818 m³ at different opening speed and degree of the door. Compared with the present internal recycle air cleaning unit of main control room of AP1000, we recommended a design which can remove the radioactivity incurred from ingress/egress more effectively and rapidly. Therefore, to reduce the radiation risk of operators in main control room at designed basic accidents, the implementation of replacement is proposed as soon as possible.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111060"},"PeriodicalIF":1.9,"publicationDate":"2024-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658379","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":"Comparison research on numerical modeling methods for the passive heat removal system of fast reactors based on PLANDTL benchmark experiment","authors":"Haijie Song ,&nbsp;Yuhao Zhang ,&nbsp;Haiqi Zhao ,&nbsp;Danting Sui ,&nbsp;Daogang Lu","doi":"10.1016/j.anucene.2024.111044","DOIUrl":"10.1016/j.anucene.2024.111044","url":null,"abstract":"<div><div>The Direct Reactor Auxiliary Cooling System (DRACS) is an innovative Passive Decay Heat Removal System (PDHRS) designed for pool-type Sodium-cooled Fast Reactors (SFR). It comprises an independent heat exchanger submerged in a sodium pool, connected to an air cooling system via a sodium loop. To validate its effectiveness, several studies have been conducted. Initially, system analysis codes were employed; however, they struggled to capture the detailed 3-D thermal–hydraulic phenomena within the sodium pool. The advent of computational fluid dynamics (CFD) has enabled a more comprehensive study of thermal–hydraulic behaviors in sodium pools and reactor cores, which include multiple fuel subassemblies and bundles. Full CFD simulations require substantial computational resources. To address these challenges, alternative methods have been proposed, such as using porous media to represent fuel bundles and employing a partial system + partial CFD approach instead of full CFD. Despite the development of these modeling methods, comprehensive comparisons assessing their applicability and uncertainty remain lacking. This study conducts four types of numerical simulations based on the aforementioned method pairs—“Bundles/Porous Media” and “Full-CFD/System + CFD”—using the PLANDTL benchmark experiment to evaluate their effectiveness. The “System + CFD” coupled approach demonstrated superior accuracy in predicting DRACS operation and its variation boundaries, with an average error of less than 4.2 %. Both models successfully captured the overall thermal–hydraulic characteristics. The rod bundles model provided more detailed understanding of natural circulation flow paths within the core and yielded more accurate temperature distribution, with average error below 4.0 %. Additionally, the simulations accurately captured core outlet backflow and inter-wrapper flow paths. The analysis revealed comprehensive temperature stratification in the upper plenum, resulting in detailed 3-D temperature distributions. These findings offer valuable insights for optimizing calculation and modeling methods and elucidate critical thermal–hydraulic characteristics essential for the innovative design of DRACS in pool-type SFRs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111044"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658373","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":"Finding limiting rods for dry storage analyses","authors":"Piotr Konarski,&nbsp;Alexey Cherezov,&nbsp;Cedric Cozzo,&nbsp;Grigori Khvostov,&nbsp;Hakim Ferroukhi","doi":"10.1016/j.anucene.2024.111043","DOIUrl":"10.1016/j.anucene.2024.111043","url":null,"abstract":"<div><div>Fuel rod status during reactor operation must be numerically modeled before dry storage analyses. Simulating all the rods ever irradiated in Switzerland with state-of-the-art best estimate tools would be too time consuming. The goal of this study is to develop a methodology for identifying limiting BWR rods, where the criterion of limiting is the rod internal pressure. Limiting rods are those whose parameters may challenge rod integrity during dry storage, subsequent transport, or handling. The presence of radial hydrides is of particular importance. The precipitation of radially oriented hydrides is associated with strong tensile hoop stress of the cladding. This can be a consequence of high rod internal pressure which has been chosen the limiting parameter. In the first step, the 3D core simulators provide irradiation data. Then, the core simulator results along with rod geometrical parameters are submitted to the principal component analysis. The multidimensional datasets are reduced to 2D and processed by a clustering algorithm or not, depending on the assembly design. The limiting rods are identified for each batch or cluster and simulated with the fuel performance code Falcon and the hydrogen behavior code HYPE.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111043"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658375","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":"Recursive data reconciliation with nonlinear characteristic constraints for typical heat exchangers in nuclear power plant","authors":"Tianyang Xing,&nbsp;Mudi Jiang,&nbsp;Xiaoliang Zhu,&nbsp;Bin Han,&nbsp;Jianqun Xu,&nbsp;Xinfei Yang,&nbsp;Mengmeng Ji","doi":"10.1016/j.anucene.2024.111054","DOIUrl":"10.1016/j.anucene.2024.111054","url":null,"abstract":"<div><div>Data reconciliation has been extensively studied in the Nuclear Power Industry because of its benefits including reducing the uncertainty of measurement data and economic superiority. Previous reconciliation methods usually neglect necessary characteristic constraints, causing certain deviation under stable or dynamic situations. By making full use of redundant information from equipment thermodynamic state equations and control transformation functions, a recursive data reconciliation method is proposed to narrow estimation deviation in two aspects. First, different reconciled methods including implicit method, explicit method, coupled method and synthesized method were established based on bilinear orthogonal transformation. Second, recursive process was designed for reconciliation between virtual device and real device. Two typical heat exchanger systems in nuclear plant were selected as case studies. Results show that the proposed reconciliation method decreases the system error in both stable and dynamic situations. Moreover, when implementing the new proposed data reconciliation method to a preheating system with two heat exchangers, it can converge to the specified residual error within 10 iterations. Recursive reconciliation method which was proposed in this paper provides systematic guidance for nuclear power plant operating and maintenance involving data reconciliation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111054"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658378","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":"Unresolved resonance parameter evaluation and uncertainty quantification of n+181Ta reactions","authors":"Jesse M. Brown ,&nbsp;Devin P. Barry ,&nbsp;Amanda Lewis ,&nbsp;Timothy H. Trumbull ,&nbsp;Marco T. Pigni ,&nbsp;Travis Greene ,&nbsp;Robert C. Block ,&nbsp;Alec Golas ,&nbsp;Yaron Danon","doi":"10.1016/j.anucene.2024.111013","DOIUrl":"10.1016/j.anucene.2024.111013","url":null,"abstract":"<div><div>Nuclear technology applications, including reactor modeling, accelerator design, and isotope production, strongly depend on evaluated nuclear data libraries and their uncertainty information for the assessment of predictive accuracy of calculated quantities. Major nuclear data libraries such as JENDL-5, JEFF-3.3, and ENDF/B-VIII.0 lack uncertainty information for <span><math><mrow><mi>n</mi><mo>+</mo></mrow></math></span>¹⁸¹Ta reactions. In addition to the lack of evaluated uncertainty information even in major nuclear data library releases, the most current US ENDF/B-VIII.0 evaluation of the unresolved resonance region (URR) does not extend to high enough energies to appropriately account for resonance self-shielding effects. This work addresses these shortcomings through a new evaluation of the URR, performed with the SAMMY evaluation tool, which extends the evaluation of the URR to encompass neutron energies of 2.5 keV to 100 keV. This study reports evaluated covariances and includes newly measured data in the evaluation analysis that were unavailable to previous evaluators. The new evaluation was designed to be closely coupled to the resolved resonance region evaluation to improve consistency across multiple evaluation regions. The updated cross sections in the URR have reduced capture and total cross sections, which improve agreement with differential measurements compared to ENDF/B-VIII.0, but they deviate slightly further from integral benchmarks.<span><span>¹</span></span></div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111013"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658371","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":"Subcooled flow boiling in a horizontal circular pipe under high heat flux and high mass flux conditions","authors":"Deepak K. Solanki ,&nbsp;Kausik Nandi ,&nbsp;Joe Mohan ,&nbsp;Arunkumar Sridharan ,&nbsp;S.V. Prabhu","doi":"10.1016/j.anucene.2024.111030","DOIUrl":"10.1016/j.anucene.2024.111030","url":null,"abstract":"<div><div>Subcooled flow boiling of water is widely observed in high heat flux and high mass flux (HHHM) cooling applications such as heat exchangers, refrigeration equipment, boiler tubes and nuclear reactor core fuel channels in pressurized heavy water reactors (PHWR). In this study, the focus is on investigating the local heat transfer coefficient (HTC) and pressure drop in a horizontal tube experiencing subcooled boiling of water under low pressure and HHHM conditions. The study encompasses different geometrical parameters such as tube diameter (5.5 <em>mm</em>, 7.5 <em>mm</em>, 9.5 <em>mm</em> and 12 <em>mm</em>) and length (550 <em>mm</em> for each of the tubes). The operating parameters that are varied include mass flux (248–2000 <em>kg/m</em>^<em>2</em>^<em>.</em><em>s</em>) and heat flux (0–1837 <em>kW/m</em>^<em>2</em>). Infrared thermography is used to measure the local wall temperature. A non-dimensional correlation for the diabatic pressure drop ratio (ratio of diabatic pressure drop to adiabatic pressure drop) as a function of Jakob number (<span><math><mrow><mi>Ja</mi></mrow></math></span>), Boiling number (<span><math><mrow><mi>Bo</mi></mrow></math></span>) and diameter ratio is developed. Subcooled boiling pressure drop ratio for 5.5 <em>mm</em>, 7.5 <em>mm</em> and 9.4 <em>mm</em> diameter tubes is 2.23 which is independent of diameter. A correlation for the two phase local HTC during subcooled flow boiling conditions as a function of <span><math><mrow><mi>Ja</mi></mrow></math></span>, <span><math><mrow><mi>Bo</mi></mrow></math></span> and Prandtl number (<span><math><mrow><mi>Pr</mi></mrow></math></span>) is also developed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111030"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658377","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 fast and cost-effective partial defect detector for spent fuel transfer verification to enhance nuclear safeguards","authors":"Yeongjun Kim ,&nbsp;Haneol Lee ,&nbsp;Man-Sung Yim","doi":"10.1016/j.anucene.2024.111045","DOIUrl":"10.1016/j.anucene.2024.111045","url":null,"abstract":"<div><div>The current nuclear safeguards approach to spent nuclear fuel inspection at nuclear power stations is based on item counting and limited partial defect analysis. With the expected surge in spent fuel storage, limited spent fuel storage pool capacity, and the increasing need for transferring fuel to long-term storage facilities, there is a growing demand for more efficient and cost-effective nuclear safeguards approaches for nuclear materials management in civilian nuclear power facilities. This study proposes a scintillator-based indirect gamma detector for spent fuel inventory screening inspection, specifically designed for use in interim storage pools prior to fuel transfer to difficult-to-access storage. This paper presents the design of the proposed detector, its application for spent fuel screening inventory inspection, and analysis using MCNP for partial defect detection. Results of analysis indicated that verifying a ∼ 13.6 % level of randomly distributed fuel defect for the Westinghouse 17x17 fuel assembly is possible using this approach. The performance evaluation also indicates that inspection of spent fuel assemblies of various vendor types against 1 SQ diversion may be possible.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111045"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658372","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}