This report focuses on the prediction of materials performance for the carbon steel corrosion-allowance container as a function of thermal loading for the potential repository at Yucca Mountain. Low, intermediate and high thermal loads were evaluated as to their performance given assumptions regarding the temperature and humidity changes with time and the resultant depth of corrosion penetration. The reference case involved a kinetic relation for corrosion that was utilized in a sensitivity analysis to examine the impacts of time exponent, pitting, and mirobiologically-influenced corrosion. As a result of this study, the high thermal load appears to offer the best corrosion performance. However, other factors must be considered in making the final thermal loading decision.
{"title":"Impact of thermal loading on waste package material performance","authors":"D. Stahl, J. K. Mccoy, R. D. Mccright","doi":"10.1557/PROC-353-671","DOIUrl":"https://doi.org/10.1557/PROC-353-671","url":null,"abstract":"This report focuses on the prediction of materials performance for the carbon steel corrosion-allowance container as a function of thermal loading for the potential repository at Yucca Mountain. Low, intermediate and high thermal loads were evaluated as to their performance given assumptions regarding the temperature and humidity changes with time and the resultant depth of corrosion penetration. The reference case involved a kinetic relation for corrosion that was utilized in a sensitivity analysis to examine the impacts of time exponent, pitting, and mirobiologically-influenced corrosion. As a result of this study, the high thermal load appears to offer the best corrosion performance. However, other factors must be considered in making the final thermal loading decision.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"56 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87044945","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}
Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1Bmore » and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests.« less
{"title":"Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials","authors":"J. L. Binder, L. McUmber, B. W. Spencer","doi":"10.2172/140597","DOIUrl":"https://doi.org/10.2172/140597","url":null,"abstract":"Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1Bmore » and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"47 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78870917","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}
Sandia National Laboratories, at the request of the Savannah River Technology Center (SRTC), studied the interactions between single drops of molten aluminum-lithium alloys and water. Most experiments were performed with ``B`` alloy (3.1 w/o Li, balance A1). Objectives were to develop experimental procedures for preparing and delivering the melt drops and diagnostics for characterizing the interactions, measure hydrogen generated by the reaction between melt and water, examine debris recovered after the interaction, determine changes in the aqueous phase produced by the melt-water chemical reactions, and determine whether steam explosions occur spontaneously under the conditions studied. Although many H{sub 2} bubbles were generated after the drops entered the water, spontaneous steam explosions never occurred when globules of the ``B`` alloy at temperatures between 700 and 1000C fell freely through water at room temperature, or upon or during subsequent contact with submerged aluminum or stainless steel surfaces. Total amounts of H{sub 2} (STP) increased from about 2 to 9 cm{sup 3}/per gram of melt as initial melt temperature increased over this range of temperatures.
{"title":"Interactions between drops of molten Al-Li alloys and liquid water","authors":"M. Hyder, L. S. Nelson, P. M. Duda, D. Hyndman","doi":"10.2172/10107057","DOIUrl":"https://doi.org/10.2172/10107057","url":null,"abstract":"Sandia National Laboratories, at the request of the Savannah River Technology Center (SRTC), studied the interactions between single drops of molten aluminum-lithium alloys and water. Most experiments were performed with ``B`` alloy (3.1 w/o Li, balance A1). Objectives were to develop experimental procedures for preparing and delivering the melt drops and diagnostics for characterizing the interactions, measure hydrogen generated by the reaction between melt and water, examine debris recovered after the interaction, determine changes in the aqueous phase produced by the melt-water chemical reactions, and determine whether steam explosions occur spontaneously under the conditions studied. Although many H{sub 2} bubbles were generated after the drops entered the water, spontaneous steam explosions never occurred when globules of the ``B`` alloy at temperatures between 700 and 1000C fell freely through water at room temperature, or upon or during subsequent contact with submerged aluminum or stainless steel surfaces. Total amounts of H{sub 2} (STP) increased from about 2 to 9 cm{sup 3}/per gram of melt as initial melt temperature increased over this range of temperatures.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"86 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77995730","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}
To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies. 10 refs., 8 figs., 32 tabs.
{"title":"Cost estimate guidelines for advanced nuclear power technologies","authors":"J. Delene, C. Hudson","doi":"10.2172/10176857","DOIUrl":"https://doi.org/10.2172/10176857","url":null,"abstract":"To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies. 10 refs., 8 figs., 32 tabs.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"15 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83457363","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 in-tank precipitation (ITP) process will be utilized as one of the steps to prepare high-level radioactive liquid wastes for vitrification in the Defense Waste Processing Facility at the Savannah River site. Hydrogen (H2) and benzene (C6H6) will be generated in the ITP waste tanks (tanks 48 and 49) as a result of radiolysis and decomposition. The structural response of these tanks to a hypothetical deflagration accident is being analyzed as part of the ITP waste tank safety analysis, and this work was performed to define the range of potential gas pressure loadings. The ITP waste tanks are equipped with a nitrogen purge gas system that removes combustible gases from the tank's vapor space and displaces oxygen. The deflagration accident scenario assumes that the purge gas system has failed and that a combustible gas mixture accumulates because of the buildup of combustible gases and inleakage of air.
{"title":"Evaluation of the gas pressure resulting from an ITP waste tank deflagration","authors":"J. K. Thomas, S. Hensel","doi":"10.2172/10188865","DOIUrl":"https://doi.org/10.2172/10188865","url":null,"abstract":"The in-tank precipitation (ITP) process will be utilized as one of the steps to prepare high-level radioactive liquid wastes for vitrification in the Defense Waste Processing Facility at the Savannah River site. Hydrogen (H2) and benzene (C6H6) will be generated in the ITP waste tanks (tanks 48 and 49) as a result of radiolysis and decomposition. The structural response of these tanks to a hypothetical deflagration accident is being analyzed as part of the ITP waste tank safety analysis, and this work was performed to define the range of potential gas pressure loadings. The ITP waste tanks are equipped with a nitrogen purge gas system that removes combustible gases from the tank's vapor space and displaces oxygen. The deflagration accident scenario assumes that the purge gas system has failed and that a combustible gas mixture accumulates because of the buildup of combustible gases and inleakage of air.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"14 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76746441","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}
Nuclear electric Propulsion (NEP) systems are expected to play a significant role in the exploration and exploitation of space. Unlike nuclear thermal propulsion (NTP) systems in which the hot reactor coolant is directly discharged from nozzles to provide the required thrust, NEP systems include electric power generation and conditioning units that in turn are used to drive electric thrusters. These thrusters accelerate sub atomic particles to produce thrust. The major advantage of NEP systems is the ability to provide very high specific impulses ([approximately]5000 s) that minimize the requirement for propellants. In addition, the power systems used in NEP could pro vide the dual purpose of also providing power for the missions at the destination. This synergism can be exploited in shared development costs. The NEP systems produce significantly lower thrust that NTP systems and are generally more massive. Both systems have their appropriate roles in a balanced space program. The technology development needs of NEP systems differ in many important ways from the development needs for NTP systems because of the significant differences in the operating conditions of the systems. The NEP systems require long-life reactor power systems operating at power levels that are considerably lower than those formore » NTP systems. In contrast, the operational lifetime of an NEP system (years) is orders of magnitude longer than the operational lifetime of NTP systems (thousands of second). Thus, the critical issue of NEP is survivability and reliable operability for very long times at temperatures that are considerably more modest than the temperatures required for effective NTP operations but generally much higher than those experienced in terrestrial reactors.« less
{"title":"Critical technologies for reactors used in nuclear electric propulsion","authors":"S. Bhattacharyya","doi":"10.2172/10171767","DOIUrl":"https://doi.org/10.2172/10171767","url":null,"abstract":"Nuclear electric Propulsion (NEP) systems are expected to play a significant role in the exploration and exploitation of space. Unlike nuclear thermal propulsion (NTP) systems in which the hot reactor coolant is directly discharged from nozzles to provide the required thrust, NEP systems include electric power generation and conditioning units that in turn are used to drive electric thrusters. These thrusters accelerate sub atomic particles to produce thrust. The major advantage of NEP systems is the ability to provide very high specific impulses ([approximately]5000 s) that minimize the requirement for propellants. In addition, the power systems used in NEP could pro vide the dual purpose of also providing power for the missions at the destination. This synergism can be exploited in shared development costs. The NEP systems produce significantly lower thrust that NTP systems and are generally more massive. Both systems have their appropriate roles in a balanced space program. The technology development needs of NEP systems differ in many important ways from the development needs for NTP systems because of the significant differences in the operating conditions of the systems. The NEP systems require long-life reactor power systems operating at power levels that are considerably lower than those formore » NTP systems. In contrast, the operational lifetime of an NEP system (years) is orders of magnitude longer than the operational lifetime of NTP systems (thousands of second). Thus, the critical issue of NEP is survivability and reliable operability for very long times at temperatures that are considerably more modest than the temperatures required for effective NTP operations but generally much higher than those experienced in terrestrial reactors.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"44 1","pages":"17908"},"PeriodicalIF":0.0,"publicationDate":"1993-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75322916","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 accelerator transmutation of waste (ATW) concept is aimed at destroying key long-lived radionuclides (both actinides and fission products) in nuclear wastes, thereby reducing the long-term risks associated with the storage of such wastes. This technology could evolve into an approach to the production of fission power, utilizing abundant natural fuels and producing minimal long-lived nuclear waste. An ATW system would consist of the following components: 1. proton accelerator; 2. heavy-metal target; 3. moderating blanket; 4. thermal-to-electric power conversion plant; and 5. chemical separation facility. The linear accelerator provides a medium-energy, high-current proton beam that is directed at a heavy-metal target. The target converts the proton beam through spallation reactions into an intense neutron flux that is thermalized in the blanket region surrounding the target. The radioactive material to be transmuted is circulated through the blanket, where it undergoes neutron-induced reactions. Long-lived fission products undergo (n, [gamma]) reactions followed by beta decay, producing short-lived or stable products. The actinides are fissioned, producing additional neutrons and an assortment of fission products to reduce parasitic absorption in the blanket and to prevent further activation of these materials to long-lived radionuclides.
{"title":"The Accelerator Transmutation of Waste (ATW) concept overview","authors":"H. Dewey","doi":"10.2172/10190589","DOIUrl":"https://doi.org/10.2172/10190589","url":null,"abstract":"The accelerator transmutation of waste (ATW) concept is aimed at destroying key long-lived radionuclides (both actinides and fission products) in nuclear wastes, thereby reducing the long-term risks associated with the storage of such wastes. This technology could evolve into an approach to the production of fission power, utilizing abundant natural fuels and producing minimal long-lived nuclear waste. An ATW system would consist of the following components: 1. proton accelerator; 2. heavy-metal target; 3. moderating blanket; 4. thermal-to-electric power conversion plant; and 5. chemical separation facility. The linear accelerator provides a medium-energy, high-current proton beam that is directed at a heavy-metal target. The target converts the proton beam through spallation reactions into an intense neutron flux that is thermalized in the blanket region surrounding the target. The radioactive material to be transmuted is circulated through the blanket, where it undergoes neutron-induced reactions. Long-lived fission products undergo (n, [gamma]) reactions followed by beta decay, producing short-lived or stable products. The actinides are fissioned, producing additional neutrons and an assortment of fission products to reduce parasitic absorption in the blanket and to prevent further activation of these materials to long-lived radionuclides.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"72 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84737825","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 Nuclear Regulatory Commission requires utilities to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the iodine concentration in the reactor coolant system (RCS) due to iodine spikes, primary flashing and bypass fractions, and iodine partitioning in the secondary coolant system (SCS). Experimental and analytical investigations have recently been completed wherein these assumptions were tested to determine whether and to what degree they were conservative (that is, whether they result in a calculated iodine source term/dose that is at least as large or larger than that expected during an actual event). The current study has the objective to assess the overall effects of the results of these investigations on the calculated iodine dose to the environment during an SGTR. To assist in this study, a computer program, DOSE, was written. This program uses a simple, non-mechanistic model to calculate the iodine source term to the environment during an SGTR as a function of water mass inventories and flow rates and iodine concentrations in the RCS and SCS. The principal conclusion of this study is that the iodine concentrationmore » in the RCS is the dominant parameter, due to the dominance of primary flashing on the iodine source term.« less
{"title":"Assessment of dose during an SGTR","authors":"J. Adams","doi":"10.2172/10145930","DOIUrl":"https://doi.org/10.2172/10145930","url":null,"abstract":"The Nuclear Regulatory Commission requires utilities to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the iodine concentration in the reactor coolant system (RCS) due to iodine spikes, primary flashing and bypass fractions, and iodine partitioning in the secondary coolant system (SCS). Experimental and analytical investigations have recently been completed wherein these assumptions were tested to determine whether and to what degree they were conservative (that is, whether they result in a calculated iodine source term/dose that is at least as large or larger than that expected during an actual event). The current study has the objective to assess the overall effects of the results of these investigations on the calculated iodine dose to the environment during an SGTR. To assist in this study, a computer program, DOSE, was written. This program uses a simple, non-mechanistic model to calculate the iodine source term to the environment during an SGTR as a function of water mass inventories and flow rates and iodine concentrations in the RCS and SCS. The principal conclusion of this study is that the iodine concentrationmore » in the RCS is the dominant parameter, due to the dominance of primary flashing on the iodine source term.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"96 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84606243","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}
From the beginning at Los Alamos National Laboratory (LANL), solutions to the transport equation were very important. Many long-forgotten approximate solution techniques, including one by Feynman, were developed to help design nuclear weapons. Most of these methods were based on the methods of mathematical physics familiar to the project physicists and predated the use of computers, but continued research and pressing need produced two new and powerful computer-based systems: Monte Carlo and the S[sub N] method. The healthy and long-term competition between the two LANL groups responsible for these quite different approaches was both stimulating and synergistic.
从洛斯阿拉莫斯国家实验室(Los Alamos National Laboratory, LANL)开始，输运方程的解就非常重要。许多长期被遗忘的近似解技术，包括费曼的一种，都是为了帮助设计核武器而开发的。这些方法大多是基于项目物理学家熟悉的数学物理方法，并且早于计算机的使用，但持续的研究和迫切的需求产生了两个新的强大的基于计算机的系统:蒙特卡罗和S[sub N]方法。负责这些截然不同的方法的两个LANL集团之间健康和长期的竞争既刺激又协同。
{"title":"The early days of the S{sub n} method","authors":"K. D. Lathrop","doi":"10.2172/10149264","DOIUrl":"https://doi.org/10.2172/10149264","url":null,"abstract":"From the beginning at Los Alamos National Laboratory (LANL), solutions to the transport equation were very important. Many long-forgotten approximate solution techniques, including one by Feynman, were developed to help design nuclear weapons. Most of these methods were based on the methods of mathematical physics familiar to the project physicists and predated the use of computers, but continued research and pressing need produced two new and powerful computer-based systems: Monte Carlo and the S[sub N] method. The healthy and long-term competition between the two LANL groups responsible for these quite different approaches was both stimulating and synergistic.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"59 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1992-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83932788","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}
R. Gehrke, M. Putnam, S. G. Goodwin, R. L. Kynaston
A technique has been developed to rapidly measure the presence of plutonium in soils, filters, smears, and glass waste forms by measuring the uranium L-shell x-ray emissions associated with the decay of plutonium. In addition, the technique can simultaneously acquire spectra of samples and automatically analyze them for the amount of americium, and gamma-ray emitting activation and fission products present. The samples are counted with a large area, thin-window, n-type Ge spectrometer which is equally efficient for the detection of low energy x-rays (>10 key), as well as high-energy gamma rays (>1 MeV). A 8192-channel analyzer is used to acquire the entire photon spectrum at one time. A dual-energy, time-tagged pulser, that is injected into the test input of the preamplifier to monitor the energy scale, detector resolution, and pulse pile-up will be installed in FY-92. The L x-ray portion of each spectrum is analyzed by a linear least-squares spectral fitting technique originally developed for the analysis of spectra from NaI(Tl) detectors. The gamma-ray portion of each spectrum is analyzed by a standard Ge gamma-ray analysis package. Detection limits (also referred to as lower limits of detection) for plutonium in contaminated soils that have been achieved by this technique aremore » reported.« less
{"title":"Rapid assay of plutonium in soils by passive L x-ray counting","authors":"R. Gehrke, M. Putnam, S. G. Goodwin, R. L. Kynaston","doi":"10.2172/10163234","DOIUrl":"https://doi.org/10.2172/10163234","url":null,"abstract":"A technique has been developed to rapidly measure the presence of plutonium in soils, filters, smears, and glass waste forms by measuring the uranium L-shell x-ray emissions associated with the decay of plutonium. In addition, the technique can simultaneously acquire spectra of samples and automatically analyze them for the amount of americium, and gamma-ray emitting activation and fission products present. The samples are counted with a large area, thin-window, n-type Ge spectrometer which is equally efficient for the detection of low energy x-rays (>10 key), as well as high-energy gamma rays (>1 MeV). A 8192-channel analyzer is used to acquire the entire photon spectrum at one time. A dual-energy, time-tagged pulser, that is injected into the test input of the preamplifier to monitor the energy scale, detector resolution, and pulse pile-up will be installed in FY-92. The L x-ray portion of each spectrum is analyzed by a linear least-squares spectral fitting technique originally developed for the analysis of spectra from NaI(Tl) detectors. The gamma-ray portion of each spectrum is analyzed by a standard Ge gamma-ray analysis package. Detection limits (also referred to as lower limits of detection) for plutonium in contaminated soils that have been achieved by this technique aremore » reported.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"8 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1992-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73476841","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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