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Journal of Nuclear Energy. Part B. Reactor Technology最新文献

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The Use of Cadmium Sulphide Photo-Conductors in Radiation Dosimetry 硫化镉光导体在辐射剂量测定中的应用
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1960-01-01 DOI: 10.1016/S0368-3273(15)30029-8
Iu.S. Deev
{"title":"The Use of Cadmium Sulphide Photo-Conductors in Radiation Dosimetry","authors":"Iu.S. Deev","doi":"10.1016/S0368-3273(15)30029-8","DOIUrl":"10.1016/S0368-3273(15)30029-8","url":null,"abstract":"<div><p>The development of new methods for measuring ionizing radiation is a current problem. The use of crystals as detectors seems particularly promising and the present paper is devoted to the question of using as detectors in dosimetry cadmium sulphide photo-conductors, which possess considerably greater sensitivity to various forms of radiation than air ionization chambers. The high sensitivity of small CdS crystals allows them to be used for solving several problems of dosimetry. A review of Soviet and foreign papers on this topic is given. Methods of obtaining polycrystalline and single-crystal photo-conductors based on CdS are briefly considered, together with their dosimetric, time and temperature characteristics. Some shortcomings of CdS detectors and possible ways of overcoming them are indicated. Circuits of actual CdS dosimeters are given.</p></div>","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 3","pages":"Pages 204-210"},"PeriodicalIF":0.0,"publicationDate":"1960-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30029-8","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77704689","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: 3
Critical heat fluxes for a bundle of tubes cooled in pressurized water 在加压水中冷却的一束管子的临界热流密度
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1960-01-01 DOI: 10.1016/S0368-3273(15)30027-4
B.A. Zenkevich, V.I. Subbotin, M.F. Troyanov
{"title":"Critical heat fluxes for a bundle of tubes cooled in pressurized water","authors":"B.A. Zenkevich,&nbsp;V.I. Subbotin,&nbsp;M.F. Troyanov","doi":"10.1016/S0368-3273(15)30027-4","DOIUrl":"10.1016/S0368-3273(15)30027-4","url":null,"abstract":"","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 3","pages":"Pages 197-199"},"PeriodicalIF":0.0,"publicationDate":"1960-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30027-4","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73283571","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: 2
Automatic Reactor Control Systems 自动反应堆控制系统
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30010-9
P. Kovanits, M. Kulka
{"title":"Automatic Reactor Control Systems","authors":"P. Kovanits, M. Kulka","doi":"10.1016/S0368-3273(15)30010-9","DOIUrl":"https://doi.org/10.1016/S0368-3273(15)30010-9","url":null,"abstract":"","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"39 1","pages":"116-122"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80800737","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
Developments in Horizontal Pulsed Contactors for Liquid-Liquid Extraction Processes 液-液萃取过程中卧式脉冲接触器的研究进展
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30062-6
D.H. Logsdail , J.D. Thornton
{"title":"Developments in Horizontal Pulsed Contactors for Liquid-Liquid Extraction Processes","authors":"D.H. Logsdail ,&nbsp;J.D. Thornton","doi":"10.1016/S0368-3273(15)30062-6","DOIUrl":"10.1016/S0368-3273(15)30062-6","url":null,"abstract":"","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 1","pages":"Pages 15-24"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30062-6","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73925339","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: 12
An Empirical Correlation of the Experimental Data on Homogeneous, Highly Enriched Uranium-Hydrogen Critical Assemblies—I. Changes of Geometry 均质高富集铀氢临界组件实验数据的经验关联——ⅰ。几何变化
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30061-4
B.G. Owen, R.A. Gibson
{"title":"An Empirical Correlation of the Experimental Data on Homogeneous, Highly Enriched Uranium-Hydrogen Critical Assemblies—I. Changes of Geometry","authors":"B.G. Owen,&nbsp;R.A. Gibson","doi":"10.1016/S0368-3273(15)30061-4","DOIUrl":"10.1016/S0368-3273(15)30061-4","url":null,"abstract":"<div><p>An empirical approach is proposed to the determination of a constant (<em>B</em><sub>o</sub><sup>2</sup>) of the materials of the critical assembly which corresponds to the material buckling.</p><p>By using two empirical equations and the one-group geometrical buckling relations any change of geometry, reflected by hydrogenous materials or unreflected, may be accomplished for simple shapes.</p></div>","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 1","pages":"Pages 10-14"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30061-4","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84031653","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
Purification of Uranium Fuels by Slagging and Filtratiop 铀燃料的结渣过滤净化
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30002-X
R.F. Taylor, D.L. Banfield
{"title":"Purification of Uranium Fuels by Slagging and Filtratiop","authors":"R.F. Taylor,&nbsp;D.L. Banfield","doi":"10.1016/S0368-3273(15)30002-X","DOIUrl":"https://doi.org/10.1016/S0368-3273(15)30002-X","url":null,"abstract":"<div><p>Using uranium containing 0·1 per cent cerium as a simple model of irradiated fuel, the degree of ‘poison’ removal by selfslagging methods has been found to be comparable with that achieved at very low alloying concentrations. By filtration through oxide beds, removals of 95 per cent Ce have been obtained, the speed of processing being limited only by the time required to reach operating temperature.</p><p>Some information is presented on the behaviour of inclusions and the mechanism of cerium removal from the alloys.</p></div>","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 2","pages":"Pages 77-83, IN2-IN4"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30002-X","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91719529","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
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30008-0
A.D. Hussey
{"title":"","authors":"A.D. Hussey","doi":"10.1016/S0368-3273(15)30008-0","DOIUrl":"https://doi.org/10.1016/S0368-3273(15)30008-0","url":null,"abstract":"","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 2","pages":"Page 106"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30008-0","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91719532","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
Heat Transfer in Pipes to a Sodium-Potassium Alloy and to Mercury 管道中的热传递到钠钾合金和汞
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30011-0
P.L. Kirillov, V.I. Subbotin, M.Ya. Suvorov, M.F. Troyanov
{"title":"Heat Transfer in Pipes to a Sodium-Potassium Alloy and to Mercury","authors":"P.L. Kirillov,&nbsp;V.I. Subbotin,&nbsp;M.Ya. Suvorov,&nbsp;M.F. Troyanov","doi":"10.1016/S0368-3273(15)30011-0","DOIUrl":"10.1016/S0368-3273(15)30011-0","url":null,"abstract":"<div><p>Experimental data on heat transfer to liquid metals are often conflicting and do not accord with theoretical results. The reasons for these discrepancies have not been clear. The present work shows that the main reason for the disagreement is the thermal resistance between the heated surface and the liquid metal. This conclusion is reached from calculations of the heat transfer coefficient by two methods: from measurements of the temperature distribution in the moving liquid and from measurements of the wall temperature. The former agrees with the distribution calculated from <span>Lyon</span>'s equation. It has been found that the presence of oxygen in a sodium-potassium alloy reduces the heat transfer coefficient. The results of experiments with no contact thermal resistance agree with those of <span>Isakoff</span> and <span>Drew</span> and of <span>Brown</span><em>et al</em>.</p><p>A description is also given of the experimental apparatus, the test section, and a thermocouple whereby the temperature distribution in a stream of liquid metal can be measured.</p></div>","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 2","pages":"Pages 123-129"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30011-0","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89569632","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: 4
Similarity criteria for critical heat loading in forced liquid flow 强迫液体流动中临界热负荷的相似准则
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30014-6
B.A. Zenkevich
{"title":"Similarity criteria for critical heat loading in forced liquid flow","authors":"B.A. Zenkevich","doi":"10.1016/S0368-3273(15)30014-6","DOIUrl":"https://doi.org/10.1016/S0368-3273(15)30014-6","url":null,"abstract":"","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 2","pages":"Pages 137-140"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30014-6","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90007113","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
The Generalization of Experimental Data on Critical Heat Fluxes in Forced Convection of Sub-Cooled Water 过冷水强制对流临界热通量实验数据的推广
Journal of Nuclear Energy. Part B. Reactor Technology Pub Date : 1959-01-01 DOI: 10.1016/S0368-3273(15)30012-2
B.A. Zenkevich
{"title":"The Generalization of Experimental Data on Critical Heat Fluxes in Forced Convection of Sub-Cooled Water","authors":"B.A. Zenkevich","doi":"10.1016/S0368-3273(15)30012-2","DOIUrl":"10.1016/S0368-3273(15)30012-2","url":null,"abstract":"<div><p>An equation in dimensionless groups has been proposed by the author on the basis of experimental data on critical heat fluxes in forced convection of sub-cooled water at pressures between 100 and 210 atm. A simplification of this equation at low pressures is predicted from theory.</p><p>It is shown that the dimensionless equation previously given is valid over a wider range of pressures, namely from 35 to 210 atm.</p><p>The reduction of published experimental data concerning qcr at low pressures confirms the theoretical conclusion that the functional relations between one group to be determined and two determining groups become degenerate at water pressures near atmospheric and therefore that the dimensionless equation simplifies in this case. From this equation a formula is derived which can be used for calculations at pressures between 1 and 15 atm.</p></div>","PeriodicalId":100814,"journal":{"name":"Journal of Nuclear Energy. Part B. Reactor Technology","volume":"1 2","pages":"Pages 130-133"},"PeriodicalIF":0.0,"publicationDate":"1959-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/S0368-3273(15)30012-2","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75385818","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: 10
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