D. Sire, C. Bendixsen, E. Armstrong, R. Henry, G. B. Frandsen
The U.S. Department of Energy (DOE) is the owner of, or has on consignment, a large quantity of nuclear fuel and targets from a variety of nuclear reactors. The majority of this inventory is associated with the nuclear materials production (NMP) complex. This material must be disposed in a safe an controlled manner. The special Fuels Canning and Characterization Facility (SFCCF) is required to characterize, verify the storage can contents, and, if necessary, recan the special fuels to help ensure safe, interim storage (i.e. fission product containment and criticality control) until a special fuels disposal facility is operating, regardless of which disposal technique is adopted. The SFCCF would be located at the Idaho Chemical Processing Plant (ICPP) and would include the capability to condition the special fuels in preparation for final disposal.
{"title":"ICPP special fuels canning and characterization facility","authors":"D. Sire, C. Bendixsen, E. Armstrong, R. Henry, G. B. Frandsen","doi":"10.2172/10157267","DOIUrl":"https://doi.org/10.2172/10157267","url":null,"abstract":"The U.S. Department of Energy (DOE) is the owner of, or has on consignment, a large quantity of nuclear fuel and targets from a variety of nuclear reactors. The majority of this inventory is associated with the nuclear materials production (NMP) complex. This material must be disposed in a safe an controlled manner. The special Fuels Canning and Characterization Facility (SFCCF) is required to characterize, verify the storage can contents, and, if necessary, recan the special fuels to help ensure safe, interim storage (i.e. fission product containment and criticality control) until a special fuels disposal facility is operating, regardless of which disposal technique is adopted. The SFCCF would be located at the Idaho Chemical Processing Plant (ICPP) and would include the capability to condition the special fuels in preparation for final disposal.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"7 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1992-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81605346","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}
Not every student is meant to be a scientist. Students come into a classroom with a variety of experiences, interests, and abilities. Therefore, the goal of any science program is not the production of chemists, physicists, or biologists but the development of scientifically literature individuals: students who can question, hypothesize, test, record, and conclude. The classroom environment cannot always provide the range of real-life experiences necessary for students to internalize the scientific method. The Illinois Junior Academy of Science (IJAS), through its sponsorship of local, regional, and state science fair competitions, seeks to assist the schools by providing just such practical hands-on experiences. The IJAS-sponsored expositions allow students the opportunity to translate classroom knowledge into a form that is tangible and concrete, an event that goes beyond the walls of a classroom into the reality of the world in which they will live and work.
{"title":"Science beyond the classroom","authors":"J. Petric, J. Bonkalski","doi":"10.2505/9781933531373","DOIUrl":"https://doi.org/10.2505/9781933531373","url":null,"abstract":"Not every student is meant to be a scientist. Students come into a classroom with a variety of experiences, interests, and abilities. Therefore, the goal of any science program is not the production of chemists, physicists, or biologists but the development of scientifically literature individuals: students who can question, hypothesize, test, record, and conclude. The classroom environment cannot always provide the range of real-life experiences necessary for students to internalize the scientific method. The Illinois Junior Academy of Science (IJAS), through its sponsorship of local, regional, and state science fair competitions, seeks to assist the schools by providing just such practical hands-on experiences. The IJAS-sponsored expositions allow students the opportunity to translate classroom knowledge into a form that is tangible and concrete, an event that goes beyond the walls of a classroom into the reality of the world in which they will live and work.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"132 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1992-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74224402","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}
Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing.
{"title":"Cadmium safety rod thermal tests","authors":"J. K. Thomas, N. C. Iyer, H. Peacock","doi":"10.2172/5458924","DOIUrl":"https://doi.org/10.2172/5458924","url":null,"abstract":"Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"78 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1992-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84019430","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}
Pub Date : 1992-01-01DOI: 10.1201/9781482273168-49
K. Morehouse
A reliable method is needed to determine if foods have been irradiated and are in compliance with respect to the allowable absorbed radiation dose. Several approaches for the identification of irradiated foods are under development worldwide. These include measurement of o-tyrosine, radiolytically generated hydrocarbons from lipids and chemiluminescence or thermoluminescence, and the use of electron spin resonance (ESR) to measure free radicals trapped in calcified tissues. This paper describes the efforts being undertaken at the FDA to develop analytical procedures to monitor and identify foods that have been treated with ionizing radiation. In particular, it focuses on the use of an ESR approach to measure radiation-induced free radicals trapped in calcified tissues and the use of a capillary gas chromatography (GC)-based procedure to determine radiolytically generated hydrocarbons formed by the radiolysis of lipids in various foods.
{"title":"Identification of irradiated foods","authors":"K. Morehouse","doi":"10.1201/9781482273168-49","DOIUrl":"https://doi.org/10.1201/9781482273168-49","url":null,"abstract":"A reliable method is needed to determine if foods have been irradiated and are in compliance with respect to the allowable absorbed radiation dose. Several approaches for the identification of irradiated foods are under development worldwide. These include measurement of o-tyrosine, radiolytically generated hydrocarbons from lipids and chemiluminescence or thermoluminescence, and the use of electron spin resonance (ESR) to measure free radicals trapped in calcified tissues. This paper describes the efforts being undertaken at the FDA to develop analytical procedures to monitor and identify foods that have been treated with ionizing radiation. In particular, it focuses on the use of an ESR approach to measure radiation-induced free radicals trapped in calcified tissues and the use of a capillary gas chromatography (GC)-based procedure to determine radiolytically generated hydrocarbons formed by the radiolysis of lipids in various foods.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"9 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1992-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88623438","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}
J. Hirschfelder, P. Chambré, W.W.-L. Lee, T. Pigford, M. Sadeghi
Partitioning the actinides in spent fuel and transmuting them in actinide-burning liquid-metal reactors (ALMRs) is a potential method of reducing public risks from the geologic disposal of nuclear waste. In this paper, the authors present a comparison of radionuclide releases from burial at Yucca Mountain of spent fuel and of ALMR wastes. Two waste disposal schemes are considered. In each, the heat generation of the wastes at emplacement is 9.88 {times} 10{sup 7} W, the maximum for the repository. In the first scheme, the repository contains 86,700 tonnes of initial heavy metal (IHM) of light water reactor (LWR) spent fuel. In the second scheme, all current LWRs operate for a 40-yr lifetime, producing a total of 84,000 tonnes IHM of spent fuel. This spent fuel is treated using a pyrochemical process in which 98.4% of the uranium and 99.8% of the neptunium, plutonium, americium, and curium are extracted and fabricated into ALMR fuel, with the reprocessing wastes destined for the repository. The ALMR requires this fuel for its startup and first two reloads; thereafter, it is self-sufficient. Spent ALMR fuel is also pyrochemically reprocessed: 99.9% of the transuranics is recovered and recycled into ALMR fuel, and the wastes are placedmore » in the repository. Thus, in the second scheme, the repository contains the wastes from reprocessing all of the LWR spent fuel plus the maximum amount of ALMR reprocessing wastes allowed in the repository based on its heat generation limit.« less
{"title":"Effects of actinide burning on waste disposal at Yucca Mountain","authors":"J. Hirschfelder, P. Chambré, W.W.-L. Lee, T. Pigford, M. Sadeghi","doi":"10.2172/140748","DOIUrl":"https://doi.org/10.2172/140748","url":null,"abstract":"Partitioning the actinides in spent fuel and transmuting them in actinide-burning liquid-metal reactors (ALMRs) is a potential method of reducing public risks from the geologic disposal of nuclear waste. In this paper, the authors present a comparison of radionuclide releases from burial at Yucca Mountain of spent fuel and of ALMR wastes. Two waste disposal schemes are considered. In each, the heat generation of the wastes at emplacement is 9.88 {times} 10{sup 7} W, the maximum for the repository. In the first scheme, the repository contains 86,700 tonnes of initial heavy metal (IHM) of light water reactor (LWR) spent fuel. In the second scheme, all current LWRs operate for a 40-yr lifetime, producing a total of 84,000 tonnes IHM of spent fuel. This spent fuel is treated using a pyrochemical process in which 98.4% of the uranium and 99.8% of the neptunium, plutonium, americium, and curium are extracted and fabricated into ALMR fuel, with the reprocessing wastes destined for the repository. The ALMR requires this fuel for its startup and first two reloads; thereafter, it is self-sufficient. Spent ALMR fuel is also pyrochemically reprocessed: 99.9% of the transuranics is recovered and recycled into ALMR fuel, and the wastes are placedmore » in the repository. Thus, in the second scheme, the repository contains the wastes from reprocessing all of the LWR spent fuel plus the maximum amount of ALMR reprocessing wastes allowed in the repository based on its heat generation limit.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"46 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74192667","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}
Pub Date : 1991-01-01DOI: 10.1016/0148-9062(91)93690-8
T. Schultheis, S. Pickering, S. Orrell
{"title":"In situ testing program at the Waste Isolation Pilot Plant","authors":"T. Schultheis, S. Pickering, S. Orrell","doi":"10.1016/0148-9062(91)93690-8","DOIUrl":"https://doi.org/10.1016/0148-9062(91)93690-8","url":null,"abstract":"","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"18 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1991-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73583200","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}
Pub Date : 1991-01-01DOI: 10.1007/978-94-011-2781-3_100
L. Hodgson, S. Apple, R. A. Culp
{"title":"Niobium as an ex-vessel neutron dosimeter for PWRs","authors":"L. Hodgson, S. Apple, R. A. Culp","doi":"10.1007/978-94-011-2781-3_100","DOIUrl":"https://doi.org/10.1007/978-94-011-2781-3_100","url":null,"abstract":"","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"45 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1991-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79657245","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}
Experimental and analytical studies of thermal hydraulic processes in commercial steam generator tube sheet sludge have been previously reported. That work was performed because the authors believed that tubing corrosion occurs when the sludge deposit becomes too deep for the liquid to penetrate freely, leading to formation of a liquid deficient region with high chemical concentrations on the tube surface. The primary objective of this work is to determine analytically the extent of liquid penetration into porous sludge. The secondary objectives are determinations of liquid saturation and chemical concentration profiles along the sludge covered tube length. The method described in this paper differs from those used in previous works in that it allows specification of porosity and permeability as a function of distance into the sludge and, also, in some of the auxiliary equations used.
{"title":"Thermal hydraulics of steam generator sludge","authors":"A. Ulke, I. Goldberg","doi":"10.2172/10192807","DOIUrl":"https://doi.org/10.2172/10192807","url":null,"abstract":"Experimental and analytical studies of thermal hydraulic processes in commercial steam generator tube sheet sludge have been previously reported. That work was performed because the authors believed that tubing corrosion occurs when the sludge deposit becomes too deep for the liquid to penetrate freely, leading to formation of a liquid deficient region with high chemical concentrations on the tube surface. The primary objective of this work is to determine analytically the extent of liquid penetration into porous sludge. The secondary objectives are determinations of liquid saturation and chemical concentration profiles along the sludge covered tube length. The method described in this paper differs from those used in previous works in that it allows specification of porosity and permeability as a function of distance into the sludge and, also, in some of the auxiliary equations used.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"22 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1990-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86364763","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}
Measurements of the Portsmouth Gaseous Diffusion Plant's nuclear criticality accident radiation alarm signal response time, sound wave frequency, and sound volume levels were made to demonstrate compliance with ANSI/ANS-8.3-1986. A steady-state alarm signal is produced within one-half second of obtaining a two-out-of-three detector trip. The fundamental alarm sound wave frequency is 440 hertz. The sound volume levels are greater than 10 decibels above background and ranged from 100 to 125 A-weighted decibels. The requirements of the standard were met; however the recommended maximum sound volume level of 115 dBA was exceeded. Emergency procedures require immediate evacuation upon initiation of a facility's radiation alarm. Comparison with standards for allowable time of exposure at different noise levels indicate that the elevated noise level at this location does not represent an occupational injury hazard. 8 refs., 5 figs.
{"title":"Measurement of the Portsmouth Gaseous Diffusion Plant criticality accident alarm","authors":"R. Tayloe, B. McGinnis","doi":"10.2172/6160975","DOIUrl":"https://doi.org/10.2172/6160975","url":null,"abstract":"Measurements of the Portsmouth Gaseous Diffusion Plant's nuclear criticality accident radiation alarm signal response time, sound wave frequency, and sound volume levels were made to demonstrate compliance with ANSI/ANS-8.3-1986. A steady-state alarm signal is produced within one-half second of obtaining a two-out-of-three detector trip. The fundamental alarm sound wave frequency is 440 hertz. The sound volume levels are greater than 10 decibels above background and ranged from 100 to 125 A-weighted decibels. The requirements of the standard were met; however the recommended maximum sound volume level of 115 dBA was exceeded. Emergency procedures require immediate evacuation upon initiation of a facility's radiation alarm. Comparison with standards for allowable time of exposure at different noise levels indicate that the elevated noise level at this location does not represent an occupational injury hazard. 8 refs., 5 figs.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"36 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1990-08-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78210437","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}
Containment performance plays a dominant role in the assessment of risk associated with severe accidents. The primary concerns for containment performance are how well the containment can withstand the pressure and temperature loads associated with severe core damage accidents and whether or not the containment is bypassed. The objective of this paper is to summarize an issue characterization study performed by Brookhaven National Laboratory for an ice-condenser containment during severe accidents. This study was the first to focus in a coordinated way on the challenges and improvements relevant to an ice-condenser containment and provides a comprehensive statement of the relevant issues that can be used in the US Nuclear Regulatory Commission (NRC) staff's evaluation processes and by utilities during their individual plant examinations. The types of pressurized water reactor ice-condenser containment challenges that have been identified in past probabilistic risk assessments, including the second draft of the NUREG-1150 risk analysis for Sequoyah, are summarized and arranged according to the containment function objectives they threaten. A list of proposed potential improvements for an ice-condenser containment is given.
{"title":"Assessment of ice-condenser containment performance issues","authors":"H. Nourbakhsh, J. Lehner, J. Lane","doi":"10.2172/6851680","DOIUrl":"https://doi.org/10.2172/6851680","url":null,"abstract":"Containment performance plays a dominant role in the assessment of risk associated with severe accidents. The primary concerns for containment performance are how well the containment can withstand the pressure and temperature loads associated with severe core damage accidents and whether or not the containment is bypassed. The objective of this paper is to summarize an issue characterization study performed by Brookhaven National Laboratory for an ice-condenser containment during severe accidents. This study was the first to focus in a coordinated way on the challenges and improvements relevant to an ice-condenser containment and provides a comprehensive statement of the relevant issues that can be used in the US Nuclear Regulatory Commission (NRC) staff's evaluation processes and by utilities during their individual plant examinations. The types of pressurized water reactor ice-condenser containment challenges that have been identified in past probabilistic risk assessments, including the second draft of the NUREG-1150 risk analysis for Sequoyah, are summarized and arranged according to the containment function objectives they threaten. A list of proposed potential improvements for an ice-condenser containment is given.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"49 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1990-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73992560","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}
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