Transactions of the American Nuclear Society最新文献_第6页

Revised evaluations of fission-product cross sections 修订的裂变产物截面评估

Transactions of the American Nuclear Society Pub Date : 1998-08-01 DOI: 10.2172/663274

R. Q. Wright

引用次数: 1

Extrapolations to critical for systems with large inherent sources 外推对于具有大量固有源的系统至关重要

Transactions of the American Nuclear Society Pub Date : 1997-12-01 DOI: 10.2172/633994

J. Mihalczo, Wyatt

{"title":"Extrapolations to critical for systems with large inherent sources","authors":"J. Mihalczo, Wyatt","doi":"10.2172/633994","DOIUrl":"https://doi.org/10.2172/633994","url":null,"abstract":"An approach to delayed critical experiment was performed in 1981 at Pacific Northwest Laboratory with a cylindrical tank of plutonium-uranium nitrate solution. During this experiment, various methods to determine the critical height were used, including (1) extrapolation of the usual plot of inverse count rate vs. height, which estimates the delayed critical height (DCH); (2) the inverse count rate vs. height divided by count rate, which corrects somewhat for the change in inherent source size as the height changes; (3) ratio of spectral densities vs. height, which extrapolates to DCH; (4) extrapolations of prompt neutron decay constant vs. height, which extrapolates to the prompt critical height (PCH); and (5) inverse kinetics rod drop (IKRD) methods, which measure {Delta}k/k{Beta} very accurately for a particular solution height. The problem with some of the extrapolation methods is that the measured data are not linear with height, but, for lack of anything better, linear extrapolations are made. In addition to the measurements to determine the delayed critical height subcriticality measurements by the {sup 252}Cf source driven frequency analysis method were performed for a variety of subcriticality heights. This paper describes how all these methods were applied to obtain the critical height of a cylindrical tank of plutonium nitrate solution and how the subcritical neutron multiplication factor was obtained.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"28 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1997-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88790703","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

Numerical stability of the mixture drift flux equations 混合漂移通量方程的数值稳定性

Transactions of the American Nuclear Society Pub Date : 1997-12-01 DOI: 10.13182/NSE99-A2051

J. Doster, J. Kauffman

引用次数: 1

Metal Poisons for Criticality in Waste Streams 废物流中临界金属毒物

Transactions of the American Nuclear Society Pub Date : 1996-06-26 DOI: 10.2172/636037

T. G. Williamson, A. Goslen

引用次数: 0

Site Restoration: Estimation of Attributable Costs From Plutonium-Dispersal Accidents 场地恢复:钚扩散事故可归因成本的估算

Transactions of the American Nuclear Society Pub Date : 1996-05-01 DOI: 10.2172/249283

D. Chanin, W. B. Murfin

{"title":"Site Restoration: Estimation of Attributable Costs From Plutonium-Dispersal Accidents","authors":"D. Chanin, W. B. Murfin","doi":"10.2172/249283","DOIUrl":"https://doi.org/10.2172/249283","url":null,"abstract":"A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"43 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1996-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89391169","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: 13

An exponential discontinuous scheme for X-Y-Z geometry transport problems X-Y-Z几何输运问题的指数不连续格式

Transactions of the American Nuclear Society Pub Date : 1996-04-01 DOI: 10.2172/224949

T. Wareing, R. Alcouffe

引用次数: 9

Rapid measurement of the gaseous {sup 41}Ar released from a nuclear reactor 对核反应堆释放的气态Ar的快速测量

Transactions of the American Nuclear Society Pub Date : 1996-03-01 DOI: 10.13182/NT96-A35214

C. Chung, Chen-Yi Chen, Cheng-Jong Lee

引用次数: 2

Fusion reactor passive safety and ignitor risk-based regulation 核聚变反应堆被动安全性和点火器风险监管

Transactions of the American Nuclear Society Pub Date : 1995-12-31 DOI: 10.13182/FST96-A11963162

M. Zucchetti

引用次数: 5

Virtual radiation fields for ALARA determination 测定ALARA的虚拟辐射场

Transactions of the American Nuclear Society Pub Date : 1995-12-31 DOI: 10.2172/672123

T. Knight

{"title":"Virtual radiation fields for ALARA determination","authors":"T. Knight","doi":"10.2172/672123","DOIUrl":"https://doi.org/10.2172/672123","url":null,"abstract":"As computing power has increased, so too has the ability to model and simulate complex systems and processes. In addition, virtual reality technology has made it possible to visualize and understand many complex scientific and engineering problems. For this reason, a virtual dosimetry program called Virtual Radiation Fields (VRF) is developed to model radiation dose rate and cumulative dose to a receptor operating in a virtual radiation environment. With the design and testing of many facilities and products taking place in the virtual world, this program facilitates the concurrent consideration of radiological concerns during the design process. Three-dimensional (3D) graphical presentation of the radiation environment is made possible through the use of IGRIP, a graphical modeling program developed by Deneb Robotics, Inc. The VRF simulation program was designed to model and display a virtual dosimeter. As a demonstration of the program`s capability, the Hanford tank, C-106, was modeled to predict radiation doses to robotic equipment used to remove radioactive waste from the tank. To validate VRF dose predictions, comparison was made with reported values for tank C-106, which showed agreement to within 0.5%. Graphical information is presented regarding the 3D dose rate variation inside the tank. Cumulative dose predictions were made for the cleanup operations of tank C-106. A four-dimensional dose rate map generated by VRF was used to model the dose rate not only in 3D space but also as a function of the amount of waste remaining in the tank. This allowed VRF to predict dose rate at any stage in the waste removal process for an accurate simulation of the radiological conditions throughout the tank cleanup procedure.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"26 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78838459","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

Neutronic analysis of three-element core configurations for the Advanced Neutron Source Reactor 先进中子源堆三元件堆芯结构的中子分析

Transactions of the American Nuclear Society Pub Date : 1995-12-31 DOI: 10.2172/205866

J. Gehin

{"title":"Neutronic analysis of three-element core configurations for the Advanced Neutron Source Reactor","authors":"J. Gehin","doi":"10.2172/205866","DOIUrl":"https://doi.org/10.2172/205866","url":null,"abstract":"The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW{sub f} nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U{sub 3}Si{sub 2}-Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of {sup 235}U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119{degrees}C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400{degrees}C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"32 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89077384","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