The handling and burial of specified quantities of special nuclear material (SNM) at low-level-waste (LLW) facilities require a license from the Nuclear Regulatory Commission (NRC). With assistance from Oak Ridge National Laboratory (ORNL) staff, the NRC Office of Nuclear Material Safety and Safeguards, Low-Level-Waste and Decommissioning Projects Branch, has developed technical specifications for the nuclear criticality safety of {sup 235}U and {sup 239}Pu in LLW facilities. The objective of the development of these technical specifications was to establish a set of review criteria that are rigorously defensible that can be applied uniformly to all license applications, and that conservatively ensures that buried SNM will not pose a criticality safety concern.
{"title":"Criticality safety criteria for license review of low-level waste facilities","authors":"C. Hopper, R. H. Odegaarden, C. Parks, P. B. Fox","doi":"10.2172/33119","DOIUrl":"https://doi.org/10.2172/33119","url":null,"abstract":"The handling and burial of specified quantities of special nuclear material (SNM) at low-level-waste (LLW) facilities require a license from the Nuclear Regulatory Commission (NRC). With assistance from Oak Ridge National Laboratory (ORNL) staff, the NRC Office of Nuclear Material Safety and Safeguards, Low-Level-Waste and Decommissioning Projects Branch, has developed technical specifications for the nuclear criticality safety of {sup 235}U and {sup 239}Pu in LLW facilities. The objective of the development of these technical specifications was to establish a set of review criteria that are rigorously defensible that can be applied uniformly to all license applications, and that conservatively ensures that buried SNM will not pose a criticality safety concern.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"245 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74112912","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}
M. Kitamura, K. Funabashi, M. Kikuchi, H. Yusa, Y. Fukushima, S. Horiuchi
The removal efficiency of methyl iodide for silver-impregnated alumina from gaseous waste has been experimentally evaluated as a function of atmospheric relative humidity. A new adsorbent has been developed for the iodine filter installed in the off-gas treatment system of a radioactive waste tank vent. To improve its removal efficiency under a highly humid atmosphere, the optimum average pore size of alumina was determined to be {approximately}60 nm, and the most effective chemical form of the impregnated silver was identified as silver nitrate. Holding capability of the impregnated silver was also improved by developing a double-pore-structure alumina.
{"title":"Silver-impregnated alumina for removal of radioactive methyl iodide","authors":"M. Kitamura, K. Funabashi, M. Kikuchi, H. Yusa, Y. Fukushima, S. Horiuchi","doi":"10.13182/NT95-A35085","DOIUrl":"https://doi.org/10.13182/NT95-A35085","url":null,"abstract":"The removal efficiency of methyl iodide for silver-impregnated alumina from gaseous waste has been experimentally evaluated as a function of atmospheric relative humidity. A new adsorbent has been developed for the iodine filter installed in the off-gas treatment system of a radioactive waste tank vent. To improve its removal efficiency under a highly humid atmosphere, the optimum average pore size of alumina was determined to be {approximately}60 nm, and the most effective chemical form of the impregnated silver was identified as silver nitrate. Holding capability of the impregnated silver was also improved by developing a double-pore-structure alumina.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"21 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81270056","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}
Gamma well-logging measurements were conducted in an inactive, radioactive waste burial ground of the Savannah River Site to appraise whether any evidence existed for downward movement of radioactivity toward the water table. Similar measurements on the same wells were conducted earlier, providing a baseline from which to measure any changes in their radioactive plumes. In particular, the recent measurements sought to detect significant changes in depth location and radiation magnitude of the plumes, as well as the existence of any new plumes. By comparing measurements on a number of these wells, which were distributed on a grid pattern, it was anticipated that the general status of this section of the burial ground could be established.
{"title":"Gamma well-logging in the Old Burial Ground of the Savannah River Site","authors":"W. G. Winn, K. J. Hofstetter, K. Macmurdo","doi":"10.2172/100396","DOIUrl":"https://doi.org/10.2172/100396","url":null,"abstract":"Gamma well-logging measurements were conducted in an inactive, radioactive waste burial ground of the Savannah River Site to appraise whether any evidence existed for downward movement of radioactivity toward the water table. Similar measurements on the same wells were conducted earlier, providing a baseline from which to measure any changes in their radioactive plumes. In particular, the recent measurements sought to detect significant changes in depth location and radiation magnitude of the plumes, as well as the existence of any new plumes. By comparing measurements on a number of these wells, which were distributed on a grid pattern, it was anticipated that the general status of this section of the burial ground could be established.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"122 6 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88491602","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 : 1994-12-31DOI: 10.1016/0306-4549(94)00077-R
Mark W. Lloyd, M. A. Feltus
{"title":"Statistical core design methodology using the VIPRE thermal-hydraulics code","authors":"Mark W. Lloyd, M. A. Feltus","doi":"10.1016/0306-4549(94)00077-R","DOIUrl":"https://doi.org/10.1016/0306-4549(94)00077-R","url":null,"abstract":"","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"72 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73921945","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}
The shortcoming of the conventional piont-kinetics formulation is that it does not treat changes in the neutron flux distribution that may result from core composition changes and feedback. Two light water reactor core models were used in a numerical study to evaluate the variational method for light water reactors.
{"title":"Variational estimates of point-kinetics parameters","authors":"J. Favorite, W. Stacey","doi":"10.13182/NSE95-A24140","DOIUrl":"https://doi.org/10.13182/NSE95-A24140","url":null,"abstract":"The shortcoming of the conventional piont-kinetics formulation is that it does not treat changes in the neutron flux distribution that may result from core composition changes and feedback. Two light water reactor core models were used in a numerical study to evaluate the variational method for light water reactors.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"3 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81404086","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 : 1994-12-31DOI: 10.1016/0140-6701(95)80430-7
S. R. Fischer, T. Sype
{"title":"Lessons learned from commercial experience with nuclear plant deactivation to safe storage","authors":"S. R. Fischer, T. Sype","doi":"10.1016/0140-6701(95)80430-7","DOIUrl":"https://doi.org/10.1016/0140-6701(95)80430-7","url":null,"abstract":"","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"20 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89667859","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 : 1994-12-31DOI: 10.1007/978-94-011-0996-3_8
M. Bunn
{"title":"Management and disposition of excess weapons plutonium","authors":"M. Bunn","doi":"10.1007/978-94-011-0996-3_8","DOIUrl":"https://doi.org/10.1007/978-94-011-0996-3_8","url":null,"abstract":"","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"13 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77718902","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}
A. Hawari, R. Venkataraman, R. Fleming, E. Charles, J. Grundl, E. McGarry
A new, high-intensity reference neutron field for reactor dosimetry is in the early stages of operation at the Ford Nuclear Reactor (FNR) at the University of Michigan. Designed and constructed by the National Institute of Standards and Technology (NIST), the facility hosts calibration and validation experiments in support of materials neutron dosimetry for the nuclear power industry and for the metallurgical community engaged in estimating radiation damage in steel. This benchmark is a natural extension of a long-term NIST program to develop standard and reference neutron fields for measurement assurance applications and for testing new detectors and techniques. Field characterization and user operation of the facility is a joint effort by NIST and the Phoenix Memorial Laboratory of the University of Michigan. The materials dosimetry reference facility (NDRF) complements the [sup 235]U cavity fission source at NIST by providing a tenfold increase in fast-neutron fluence, a much larger irradiation volume with modest flux gradients and a neutron spectrum rich in intermediate-energy neutrons. Two spectrum options are available to investigate detector response characteristics and to validate the interpretation of dosimetry measurements.
{"title":"The materials dosimetry reference facility","authors":"A. Hawari, R. Venkataraman, R. Fleming, E. Charles, J. Grundl, E. McGarry","doi":"10.1520/STP15134S","DOIUrl":"https://doi.org/10.1520/STP15134S","url":null,"abstract":"A new, high-intensity reference neutron field for reactor dosimetry is in the early stages of operation at the Ford Nuclear Reactor (FNR) at the University of Michigan. Designed and constructed by the National Institute of Standards and Technology (NIST), the facility hosts calibration and validation experiments in support of materials neutron dosimetry for the nuclear power industry and for the metallurgical community engaged in estimating radiation damage in steel. This benchmark is a natural extension of a long-term NIST program to develop standard and reference neutron fields for measurement assurance applications and for testing new detectors and techniques. Field characterization and user operation of the facility is a joint effort by NIST and the Phoenix Memorial Laboratory of the University of Michigan. The materials dosimetry reference facility (NDRF) complements the [sup 235]U cavity fission source at NIST by providing a tenfold increase in fast-neutron fluence, a much larger irradiation volume with modest flux gradients and a neutron spectrum rich in intermediate-energy neutrons. Two spectrum options are available to investigate detector response characteristics and to validate the interpretation of dosimetry measurements.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"14 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89030728","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}
A report recently issued by the National Academy of Sciences describes the need to dispose of 50 metric tons of U.S. weapons-grade plutonium and a similar amount from Russia and makes recommendations for means of disposal. One principal recommendation is to use the plutonium as once-through fuel in existing commercial U.S. light water reactors (LWRs). The report states that a coordinated program of research and development should be undertaken immediately to clarify and resolve the identified technical uncertainties. This paper presents a solution to one needed program: reactor testing of mixed-oxide (MOX) fuels. Currently, weapons-grade plutonium MOX and other types of advanced plutonium-based fuels are being considered as a disposition fuel form. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in-reactor thermal, mechanical, and fission gas release behavior of a prototype fuel will most likely be required in a limited number of test reactor irradiations.
美国国家科学院(National Academy of Sciences)最近发布的一份报告指出，需要处理美国的50吨武器级钚和俄罗斯的类似数量的钚，并就处理方法提出了建议。一项主要建议是在美国现有的商业轻水反应堆(LWRs)中使用钚作为一次性燃料。报告指出，应立即开展一项协调的研究和开发计划，以澄清和解决已确定的技术不确定性。本文提出了一个解决方案，需要一个程序:混合氧化物(MOX)燃料的反应堆测试。目前正在考虑将武器级钚MOX和其他类型的先进钚基燃料作为处置燃料形式。拟议的武器级MOX燃料是不寻常的，甚至相对于国外正在进行的反应堆级MOX动力反应堆燃料的经验。在有限数量的试验反应堆辐照中，很可能需要对原型燃料在反应堆内的热、机械和裂变气体释放行为进行一些演示。
{"title":"Mixed-oxide fuels testing in the advanced test reactor","authors":"J. Sterbentz, J. Ryskamp, S. Mason, G. Chang","doi":"10.2172/125351","DOIUrl":"https://doi.org/10.2172/125351","url":null,"abstract":"A report recently issued by the National Academy of Sciences describes the need to dispose of 50 metric tons of U.S. weapons-grade plutonium and a similar amount from Russia and makes recommendations for means of disposal. One principal recommendation is to use the plutonium as once-through fuel in existing commercial U.S. light water reactors (LWRs). The report states that a coordinated program of research and development should be undertaken immediately to clarify and resolve the identified technical uncertainties. This paper presents a solution to one needed program: reactor testing of mixed-oxide (MOX) fuels. Currently, weapons-grade plutonium MOX and other types of advanced plutonium-based fuels are being considered as a disposition fuel form. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in-reactor thermal, mechanical, and fission gas release behavior of a prototype fuel will most likely be required in a limited number of test reactor irradiations.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"28 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89343896","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}
At the request of the Department of Energy and Westinghouse Hanford Company, the Bureau of Mines has investigated the flammability of mixtures of hydrogen, ammonia, nitrous oxide, and air. The tests were performed in a spherical chamber under quiescent and turbulent conditions. This paper describes combustion calculations using the GASFLOW code and compares the calculated pressure ratios with experiments mentioned above. GASFLOW is a finite-volume computer code that solves the transient, three-dimensional, compressible fluid, Navier-Stokes equations with multiple species coupled with finite-rate chemical kinetics. The computational results show good agreement with the experimental data and confirm GASFLOW to be a valuable tool for evaluating the above combustion process.
{"title":"GASFLOW comparisons with bureau of mines experiments","authors":"C. Mueller, J. Travis","doi":"10.2172/10181611","DOIUrl":"https://doi.org/10.2172/10181611","url":null,"abstract":"At the request of the Department of Energy and Westinghouse Hanford Company, the Bureau of Mines has investigated the flammability of mixtures of hydrogen, ammonia, nitrous oxide, and air. The tests were performed in a spherical chamber under quiescent and turbulent conditions. This paper describes combustion calculations using the GASFLOW code and compares the calculated pressure ratios with experiments mentioned above. GASFLOW is a finite-volume computer code that solves the transient, three-dimensional, compressible fluid, Navier-Stokes equations with multiple species coupled with finite-rate chemical kinetics. The computational results show good agreement with the experimental data and confirm GASFLOW to be a valuable tool for evaluating the above combustion process.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"52 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86542986","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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