Journal of Nuclear Engineering and Radiation Science最新文献_第4页

Thermomechanical Performance Assessment of U-Mo Monolithic Fuel Plates with Zircaloy Cladding 锆合金包覆U-Mo单片燃料板的热力学性能评价

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-08-11 DOI: 10.1115/1.4063163

H. Ozaltun, J. Cole, B. Rabin

{"title":"Thermomechanical Performance Assessment of U-Mo Monolithic Fuel Plates with Zircaloy Cladding","authors":"H. Ozaltun, J. Cole, B. Rabin","doi":"10.1115/1.4063163","DOIUrl":"https://doi.org/10.1115/1.4063163","url":null,"abstract":"\u0000 Performance of two distinct fuel systems, U-7Mo fuel in zircaloy cladding and U-10Mo fuel in aluminum cladding was studied. First, a mini plate with zircaloy cladding from a previous experiment was evaluated via finite element analysis. By using the same plate geometry and irradiation conditions, another plate consisting of U-10Mo fuel and Al6061-O cladding was simulated. The results were then comparatively evaluated to explore the feasibility of employing zircaloy as an alternative cladding. Simulations indicated the zircaloy cladding plate would operate roughly 50°C hotter as compared with the Al alloy cladding plate. Larger deformations in the thickness direction for the plate with zircaloy cladding were noted. It was observed that the post-fabrication stresses in the fuel would be relieved quickly in the reactor, regardless of cladding type. Although the fuel stresses would still develop at reactor shutdown, the fuel would be stress-free during the irradiation for both cladding types. At shutdown, the plate with zircaloy cladding would have higher stresses due to higher operating temperatures. Similarly, the stresses after shutdown are higher in the foil core for the plates with zircaloy cladding. The Al cladding plate would have higher plastic strains as compared with the zircaloy cladding plate. The zircaloy cladding plate is significantly stiffer, causing higher stresses in the fuel and at the interface. Overall, employing zircaloy as an alternate cladding is not expected to produce a more favorable thermomechanical performance as compared to the performance of an Al alloy cladding plate.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"62 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90633481","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Survey Analysis of Potential Nuclear Safety Research of Thailand for International Research Collaborative Reinforcement in the 2020s 20世纪20年代泰国核安全研究潜力国际合作调研分析

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-08-11 DOI: 10.1115/1.4063162

W. Vechgama, K. Silva

{"title":"Survey Analysis of Potential Nuclear Safety Research of Thailand for International Research Collaborative Reinforcement in the 2020s","authors":"W. Vechgama, K. Silva","doi":"10.1115/1.4063162","DOIUrl":"https://doi.org/10.1115/1.4063162","url":null,"abstract":"\u0000 To achieve the long-term challenge of nuclear energy public acceptance in Thailand, nuclear safety research needed to be properly determined in both domestic and international directions, especially in the 2020s which was a period passing the Fukushima disaster over 10 years. Thailand Institute of Nuclear Technology (TINT) has studied nuclear safety research after the Fukushima accident to answer technical and social issues of nuclear power. An update of nuclear safety research from domestic experts and international surveys was needed in order to identify potential collaborative research to serve the goal of public acceptance reinforcement. The objective of this study was to survey, assess and rank the importance and knowledge level of nuclear safety research in Thailand among domestic experts in various fields. The survey was extended to collect the opinion of international participants of the ASEAN Network on Nuclear Power Safety Research (ASEAN NPSR) to analyze the similarity of the nuclear research interest for reinforcing the future collaborative project. As a result, the importance and knowledge level showed diverse important research topics with the priority of research scopes on human factor, novel reactor technologies, and risk assessment. According to the ASEAN NPSR survey, the nuclear safety research of severe accident, risk assessment, and novel reactor technologies were listed as potential collaborative projects. Also, the domestic and ASEAN NPSR survey results helped support the new collaborative research extension session in the annual ASEAN NPSR meeting to together discuss the potential nuclear safety research between members for the 2020s.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"9 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86769508","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Development Of Analysis Tools For Heat Pipes Used In Small Modular Reactors: Sodium Property Correlations 小型模块化反应器用热管分析工具的发展:钠性质相关性

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Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-08-05 DOI: 10.1115/1.4063112

T. Beuthe, Aleksandar Vasić, C. Azih, Pablo Diazgomezmaqueo

引用次数: 0

Depletion Calculations For An Integral Small Molten Salt Reactor With Serpent 带蛇的整体式小型熔盐堆耗损计算

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-08-05 DOI: 10.1115/1.4063111

Xiaolin Wang, T. S. Nguyen, D. Wojtaszek

{"title":"Depletion Calculations For An Integral Small Molten Salt Reactor With Serpent","authors":"Xiaolin Wang, T. S. Nguyen, D. Wojtaszek","doi":"10.1115/1.4063111","DOIUrl":"https://doi.org/10.1115/1.4063111","url":null,"abstract":"\u0000 The molten salt reactor (MSR) concept is among the Generation IV designs considered feasible for providing clean, safe, sustainable, and economical energy supplies to the world's population. The depletion of fuel for a small modular fluoride molten salt reactor (sm-FMSR) with a closed fuel cycle based on the integral molten salt reactor (IMSR) concept has been investigated using Serpent. Three fueling schemes to control Serpent depletion cycles have been simulated and compared: step fueling (SF), continuous fueling with all fission products (FPs) accumulating in the reactor system (CFA), and continuous fueling with insoluble FPs separated from fuel (CFS). Sm-FMSRs with SF and with CFA require similar quantities of \"top-up\" fuel, consume similar fuel (fissile) amounts, and result in similar fuel isotopic concentrations if keff is kept within a similar range. However, with separation of insoluble FPs from the circulating fuel, CFS gains a large reactivity worth due to the removal of FP poisons. This allows for reduction of fuel enrichment in both initial and total top-up fuel, and leads to savings of a considerable fissile quantity in fueling MSR and in spent fuel. The Serpent depletion calculations require manual arithmetic calculations for adjustment of the Serpent built-in settings before the start of every calculation cycle for all three fueling schemes. Implementation of additional Serpent flow features in changing material volumes and flow constants would facilitate the simulation of the fuel depletion process and allow for more realistic simulations of fuel circulation.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"10 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88048095","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities 在欧盟DEMO概念前设计阶段设计DEMO驱动毯系统的进展:概述、挑战和机遇

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-08-03 DOI: 10.3390/jne4030037

F. Hernández, P. Arena, L. Boccaccini, I. Cristescu, A. Del Nevo, Pierre Sardain, G. Spagnuolo, M. Utili, A. Venturini, Guangming Zhou

{"title":"Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities","authors":"F. Hernández, P. Arena, L. Boccaccini, I. Cristescu, A. Del Nevo, Pierre Sardain, G. Spagnuolo, M. Utili, A. Venturini, Guangming Zhou","doi":"10.3390/jne4030037","DOIUrl":"https://doi.org/10.3390/jne4030037","url":null,"abstract":"The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"25 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83332011","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

The Plutonium Temperature Effect Program 钚温度效应计划

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-08-02 DOI: 10.3390/jne4030035

N. Leclaire, Vaibhav Jaiswal

{"title":"The Plutonium Temperature Effect Program","authors":"N. Leclaire, Vaibhav Jaiswal","doi":"10.3390/jne4030035","DOIUrl":"https://doi.org/10.3390/jne4030035","url":null,"abstract":"Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The PU+ experiments were conducted in the “Apparatus B” facility at the CEA VALDUC research center in France. It involved several sub-critical approach-type experiments using plutonium nitrate solutions with concentrations of 14.3, 15, and 20 g/L at temperatures ranging from 20 to 40 °C. Fourteen (five at 20 g/L, four at 15 g/L, and five at 14.3 g/L) phase I experiments (consisting of independent sub-critical approaches) were performed between 2006 and 2007. The impact of the uncertainties on solution acidity and plutonium concentration made it difficult to demonstrate the positive temperature effect, requiring an additional phase II experiment (with a unique plutonium solution) from 22 to 28 °C that was performed in July 2007. This phase II experiment has shown the existence of a positive temperature effect of ~+5.17 pcm/°C (from 22 to 28 °C for a plutonium concentration of 14.3 g/L). It has recently been possible to confirm the results of this program with MORET 5 calculations by generating thermal scattering data S(α,β) at the correct experimental temperatures. This paper finally presents a fully documented experimental program highlighting the Plutonium Temperature Effect theoretically described in the literature. Its high level of precision and its “one-step” approach to criticality allowed it to show a significant positive temperature effect for a rather small variation of temperature (+6 °C). The order of magnitude of the effect was confirmed with Monte Carlo calculations using thermal scattering data for hydrogen in the solution produced by IRSN for the purpose of the comparison.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"9 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86741510","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor 低温至高中子剂量辐照铍在BR2反应堆中的氚解吸行为及微观结构演变

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-08-02 DOI: 10.3390/jne4030036

V. Chakin, R. Rolli, R. Gaisin, W. Van Renterghem

引用次数: 0

Experimental Analysis Of Bubble Behavior And Critical Heat Flux During Pool Boiling On Vertical Circular Tubes 垂直圆管池沸腾气泡行为及临界热流密度的实验分析

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-07-25 DOI: 10.1115/1.4063041

Bikash Pattanayak, Hardik B. Kothadia

{"title":"Experimental Analysis Of Bubble Behavior And Critical Heat Flux During Pool Boiling On Vertical Circular Tubes","authors":"Bikash Pattanayak, Hardik B. Kothadia","doi":"10.1115/1.4063041","DOIUrl":"https://doi.org/10.1115/1.4063041","url":null,"abstract":"\u0000 The heat transfer during pool boiling incorporates a higher rate of heat dissipation capability at low-temperature differences. This technique is widely used in the nuclear industry for thermal management. In this study, the effect of tube diameter and length on critical heat flux (CHF) at atmospheric conditions in saturated water during pool boiling is analyzed. The tubes of SS 304 are kept in the vertical orientation. The diameter of the tubes ranges from 1.2 mm to 9 mm. The tube lengths varying from 50mm to 1000mm. It has been noted that tubes of smaller diameter show a greater magnitude of CHF for the given length. For a given diameter, a longer tube is found to have lower CHF than the ones having lesser length. The variation in the CHF magnitude is negligible for tubes with a diameter of more than 2.5 mm beyond a length of 200 mm. The location of occurrence of CHF is near the bottom end of the vertical tube. The study illustrates the behavior of bubble nucleation for various tube dimensions and heat fluxes. The inception and detachment of bubbles for different tubes are analyzed. The pool boiling regime is categorized and studied basing the behaviour of the incepted and departed bubble. A mathematical relation that empirically accounts for the effect of tube dimensions i.e. length and diameter on pool boiling CHF is proposed. The experimental CHF data obtained during pool boiling are tabulated towards contributing to the CHF databank.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"31 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87405604","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

Study of γ-ray Spectrum Measurement Analysis Algorithm of Radioactive Solid Waste Steel Boxes 放射性废钢箱γ射线谱测量分析算法研究

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-07-25 DOI: 10.1115/1.4063039

Yurong Li, Lixia He, Jing Wen, Jiewen Shao

{"title":"Study of γ-ray Spectrum Measurement Analysis Algorithm of Radioactive Solid Waste Steel Boxes","authors":"Yurong Li, Lixia He, Jing Wen, Jiewen Shao","doi":"10.1115/1.4063039","DOIUrl":"https://doi.org/10.1115/1.4063039","url":null,"abstract":"\u0000 The steel boxes that contain radioactive solid wastes should be accurately identified and measured. On account of the uneven internal filling and radioactive distribution in the steel boxes that storing radioactive solid wastes, the non-destructive measurement analysis technique can be utilized to perform overall measurement of the wastes, which will not damage the steel boxes and produce secondary wastes. The γ-ray spectrum measurement analysis algorithm of radioactive solid waste steel boxes was set up by way of discrete treatment for γ-ray spectrum measurement analysis algorithm combined with the traditional CT principle in this paper, the progressive split voxel measurement method was applied to conduct discrete treatment for the radioactive solid waste steel boxes, with the influence of voxel interactions on detection efficiency was taken into consideration, the calibration model of detection efficiency was set up through Monte Carlo simulation combined with algebraic reconstruction technique, the inversion correction of efficiency for the physically coincident part of the detected voxel volume was achieved. The proposed algorithm was preliminarily applied on the analyze platform of steel boxes that storing FA-IV- type radioactive solid wastes, and the detection and verification on the standard radioactive source of known activities was also conducted, the findings revealed that the relative deviation between the reconstruction results and truth-value of radioactive activities in the steel boxes is less than 30%, which conformed to the expected results and the proposed algorithm can be popularized and applicated in the detection works in related to other.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"27 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82303388","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Linear And Non-Linear Stability Analysis Of Molten Salt Natural Circulation Loop 熔盐自然循环回路的线性与非线性稳定性分析

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-07-25 DOI: 10.1115/1.4063040

A. Srivastava, Saikrishna Nadella, N. K. Maheshwari

{"title":"Linear And Non-Linear Stability Analysis Of Molten Salt Natural Circulation Loop","authors":"A. Srivastava, Saikrishna Nadella, N. K. Maheshwari","doi":"10.1115/1.4063040","DOIUrl":"https://doi.org/10.1115/1.4063040","url":null,"abstract":"\u0000 A Molten Salt Natural Circulation Loop (MSNCL) has been setup to study the steady state, transient and stability characteristics of molten nitrate salt, mixture of Sodium Nitrate and Potassium Nitrate in 60:40 ratios by weight. Natural circulation experiments in a temperature range of 290 °C to 600 °C have been performed. A semi-analytical linear model was derived for stability analysis of rectangular natural circulation loops with conventional localized surface heating and cooling. The developed model includes the effect of wall thermal inertia, variable internal heat transfer coefficient, finite secondary side heat transfer coefficient at cooler and heat losses for predicting the stability map. These effects are incorporated considering their significant role in modelling high temperature molten salt based natural circulation systems. The developed model has been first validated with the experimental data of water loop available in literature. The developed model is then validated with the experimental data generated in MSNCL. Validation of in-house developed non-linear model has also been performed against the same experimental data. The comparison of both linear and non-linear stability analysis with the experimental data shows good agreement and articulate the importance of various parameters which have been included in the developed model.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"7 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74387250","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0