Martyn Swinhoe - Academia.edu (original) (raw)

Papers by Martyn Swinhoe

Powsner/Essentials of Nuclear Medicine Physics and Instrumentation, 2013

A scintillation detector is a device which responds to ionizing radiation by emitting light. Such... more A scintillation detector is a device which responds to ionizing radiation by emitting light. Such devices are used in many areas of scientific research and development as well as in a number of industrial applications. This article provides an overview of the physics of scintillation detectors, their production, and their uses in a number of areas of basic research and industry.

Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component i... more Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF6 cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF 6 cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

Nuclear Science and Engineering, Feb 1, 2011

Abstract Neutron coincidence measurements of plutonium samples with uncertainties <0.5% could ... more Abstract Neutron coincidence measurements of plutonium samples with uncertainties <0.5% could reduce the amount of costly destructive analysis required for nuclear material accountancy in plutonium handling plants. The ratio of (α, n) emission to spontaneous fission neutron emission, α, of plutonium samples is important to the interpretation of neutron coincidence measurements. When the “known alpha” analysis method is used, an error on the α value propagates to approximately the same percentage error on the measured plutonium mass. Molality data of Charrin and the SOURCES code have been used to update the calculation of α for both pure plutonium nitrate solutions and plutonium/uranyl nitrate solutions of different concentrations and acidity. This paper gives equations for the density of the solution as a function of heavy metal concentration and for the α weight factors that can be used in the analysis of neutron coincidence measurements.

Nuclear instruments and methods in physics research, Aug 1, 1983

A method to determine the isotopic composition of lithium compounds has been developed, with two ... more A method to determine the isotopic composition of lithium compounds has been developed, with two variations. In one, the activity of 7Be, produced when the sample is bombarded with 4.8 MeV deuterons [via the 6Li(d, n) reaction], is compared with the activity produced in a sample of the same compound with a known 6Li/VLi ratio. In the second, a cross-comparison is made between samples of different composition for both deuteron-(6Li activation) and proton-(7Li activation) induced activities of 7Be, giving 6Li/VLi ratio measurements independent of calibration. The method works well for compounds not containing hydrogen; a lower limit on the sensitivity to 6Li content was found because of residual hydrogen in the compounds investigated.

Neutron measurements are frequently used to characterize spent nuclear fuel. Curium is the primar... more Neutron measurements are frequently used to characterize spent nuclear fuel. Curium is the primary neutron source from most spent nuclear fuel materials. Recent developments in nuclear safeguards measurements of spent nuclear fuel have increased the reliance upon curium assay for materials accounting on the back end of the fuel cycle. The curium assay is used to determine the fuel composition

The second paper on the non-destructive assay techniques investigated under the Next Generation S... more The second paper on the non-destructive assay techniques investigated under the Next Generation Safeguards Initiative considers innovative instrument designs capable of producing isotope-specific responses for quantifying the fissile material content of spent nuclear fuel. The present research goal is to assess whether expected signatures can be confidently obtained in realistic spent fuel safeguards applications, contribute to the establishment of Pu inventories and determine fissile material diversions at fuel storage, handling and reprocessing facilities. In certain cases, theoretical concepts are supported by the experimental data obtained in the simplified setups. The overall effort will be concluded by nominating the most promising techniques and their combinations for the full-scale demonstration involving actual nuclear fuel assemblies. This paper provides an overview of the following 6 photon-and neutron-based techniques:

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Jul 18, 2006

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), 2009

The objective of safeguarding nuclear material is to det r diversions of significant quantities o... more The objective of safeguarding nuclear material is to det r diversions of significant quantities of nuclear materials by timely monitoring and detection. There are a variety of motivations fo r quantifying plutonium in spent fuel (SF), by means of nondestructive assay (NDA), in order to meet this goal. These motivations include the fo llowing: strengthening the capabilities of the Internati onal Atomic Energy Agencies ability to safeguard nuclear facili ties, shipper/receiver differ nee, input accountability at reprocessing faci lities and burnup credi t at repositories. Many 1 DA techniques exist for measuring signatures from SF; however, no single NDA technique can, in isolation, quanti fy elemental pl utonium in SF. A study has been undertaken to determine the best integrated combination of 13 DA techniques for characterizing Pu mass in spent fuel. This paper fo cuses on the de velopment of a passive gamma measurement system in support the spent fu el assay system. Gamma ray detection for fresh nuclear fuel focuses on gamma ray emissions that directly co inci de with the actinides of interest to the assay. For example, the 186-keV gamma ray is generall y used for 235U assay and the 384-keV complex is generally used for assayi ng pl utoni um. In sp~n t nuclear fuel , these signatures cannot be detected as the Compton continUlll1 created from the fission products dominates the si~nal in this energy range. F or SF, the measured gamma signatures fro m key fi ssion products (' 4CS, m Cs, 154Eu) are used to ascerta in burnup, cooling time, and fis sile content information. In this paper the Monte Carlo modeling setup for a passiv gamma spent fuel assay system will be described. The setup of the system incl ud s a germani um detector and an io n chamber and will be used to gain passive gamma information that will be integrated into a syst m for detemlin ing Pu in SF. The passive gamma signal will be determined from a library of ~ 100 assemblies that have be n created to examine the capability of all 13 NDA techniques. Presented in this paper is a description of the passive gamma mo nitoring instrument, explanation of the work completed thus far invol ving the source set up methodology and the design opti mizati on process, deta ils of key fission product ratios of interest, li mitations and key str ngths of the measurement technique, and considerations fo r integrating this technique with other NDA techniq ues in order to develop a complete spent fu,J assay strategy.

De Gruyter eBooks, Dec 31, 1997

The second paper on the non-destructive assay techniques investigated under the Next Generation S... more The second paper on the non-destructive assay techniques investigated under the Next Generation Safeguards Initiative considers innovative instrument designs capable of producing isotope-specific responses for quantifying the fissile material content of spent nuclear fuel. The present research goal is to assess whether expected signatures can be confidently obtained in realistic spent fuel safeguards applications, contribute to the establishment of Pu inventories and determine fissile material diversions at fuel storage, handling and reprocessing facilities. In certain cases, theoretical concepts are supported by the experimental data obtained in the simplified setups. The overall effort will be concluded by nominating the most promising techniques and their combinations for the full-scale demonstration involving actual nuclear fuel assemblies. This paper provides an overview of the following 6 photonand neutron-based techniques: Delayed Gamma, X-Ray Fluorescence, Nuclear Resonance Fluorescence, Passive Gamma, Lead Slowing-Down Spectrometry, and Neutron Resonance Transmission Analysis.

There are a variety of algorithmic approaches available to process list-mode pulse streams to pro... more There are a variety of algorithmic approaches available to process list-mode pulse streams to produce multiplicity histograms for subsequent analysis. In the development of the INCC v6.0 code to include the processing of this data format, we have noted inconsistencies in the “processed time” between the various approaches. The processed time, tp, is the time interval over which the recorded pulses are analyzed to construct multiplicity histograms. This is the time interval that is used to convert measured counts into count rates. The observed inconsistencies in tp impact the reported count rate information and the determination of the error-values associated with the derived singles, doubles, and triples counting rates. This issue is particularly important in low count-rate environments. In this report we will present a prescription for the processing of list-mode counting data that produces values that are both correct and consistent with traditional shift-register technologies. It is our objective to define conventions for list mode data processing to ensure that the results are physically valid and numerically aligned with the results from shift-register electronics.

A series of experiments using the Idaho National Laboratory (INL) photonuclear inspection system ... more A series of experiments using the Idaho National Laboratory (INL) photonuclear inspection system and a Los Alamos National Laboratory (LANL) supplied, list-mode data acquisition method have shown enhanced performance utilizing pulsed photofission-induced, neutron coincidence counting between pulses of an up-to-10-MeV electron accelerator for nuclear material detection and identification. The enhanced inspection methodology has applicability to homeland security, treaty-related support, and weapon dismantlement applications. For the latter, this technology can directly support Department of Energy/NA241 programmatic mission objectives relative to future Rocky Ridgetype testing campaigns for active inspection systems.

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), 2009

There are a variety of motivations for quantifying plutonium (Pu) in spent fuel assemblies by mea... more There are a variety of motivations for quantifying plutonium (Pu) in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capability of the International Atomic Energy Agency (LAEA) to safeguard nuclear facilities, quantifying shipperlreceiver difference, determining the input accountability value at pyrochemical processing facilities, providing quantitative input to burnup credit and final safeguards measurements at a long-term repository.

The nuclear material contained in the process equipment of a uranium enrichment plant (referred t... more The nuclear material contained in the process equipment of a uranium enrichment plant (referred to as holdup) is an important component of the overall nuclear material inventory for the plant. Accurate quantification and verification of holdup is needed to improve international safeguards and nuclear material accountancy. This is also needed for criticality safety and waste disposition. Passive neutron and gamma-ray nondestructive assay (NDA) methods are used to measure the holdup in process equipment. A key advantage of neutron measurements is that neutrons are highly penetrating and can be measured through thick walled equipment. The dominant source of neutrons in the holdup is from the reaction resulting from alpha decay when uranium is enriched. There is a considerable spread between different historic determinations of the yield from uranium which limits the accuracy of modeling and the calibration of NDA instruments. Furthermore, the compound form and presence of water also significantly affects the neutron emission rate from the holdup. This paper describes a series of experimental measurements performed at Los Alamos National Laboratory (LANL) to determine the absolute neutron emission yield from 10 different working reference materials (WRMs) fabricated at the Portsmouth Gaseous Diffusion Plant (PGDP). The Mini Epithermal Neutron Multiplicity Counter (Mini ENMC) and a NIST certified neutron source were used for these measurements. The high efficiency and short die-away time of the Mini ENMC provides the high measurement precision needed to certify the neutron emission yield. The experiment was designed to achieve sub 1% accuracy in the net counting rate on each item and to provide assurance that important factors such as instrument stability, item placement and background were well understood. The traceable neutron yields measured from the WRMs were used to determine a more accurate neutron yield for the material. The results were compared to historical neutron emission rates. The data obtained from these measurements directly supports nuclear safeguards for NDA of uranium holdup and provides a more accurate calibration for new and existing NDA detector systems.

Nuclear Instruments and Methods in Physics Research, Mar 1, 2010

Abstract Correlations are used to determine the 242 Pu content of material using high resolution ... more Abstract Correlations are used to determine the 242 Pu content of material using high resolution gamma measurements on the other plutonium isotopes because 242 Pu itself has no practically detectable gamma emission lines. This paper presents an improved correlation that is particularly useful because, unlike some previous correlations, no prior knowledge of the reactor type or initial enrichment is required. This correlation has been shown to perform well over a range of plutonium from commercial BWR and PWR reactors. The agreement of the calculated 242 Pu values with IDMS values is within 1% for 239 Pu content of less than 70% and within 4% for 239 Pu content of less than 80%. This simple form of the correlation is somewhat surprising given the complex behavior of 239/240 and 242/240 ratios.

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), May 5, 2011

During the past several years, the demand for 3He gas has far exceeded the gas supply. This short... more During the past several years, the demand for 3He gas has far exceeded the gas supply. This shortage of 3He gas is projected to continue into the foreseeable future. There is a need for alternative neutron detectors that do not require 3He gas. For more than four decades, neutron detection has played a fundamental role in the safeguarding and control of nuclear materials at production facilities, fabrication plants and storage sites worldwide. Neutron measurements for safeguards applications have requirements that are unique to the quantitative assay of special nuclear materials. These neutron systems measure the neutron multiplicity distributions from each spontaneous fission and/or induced fission event. The neutron time correlation counting requires that two or more neutrons from a single fission event be detected. The doubles and triples neutron counting rate depends on the detector efficiency to the 'L'd and :jd power, respectively, so low efficiency systems will not work for the coincidence measurements, and any detector instabilities are greatly amplified. In the current test program, we will measure the alternative detector properties including efficiency, die-away time, multiplicity precision, gamma sensitivity, dead-time, and we will also consider the detector properties that would allow commercial production to safeguards scale assay systems. This last step needs to be accomplished before the proposed technologies can reduce the demand on 3He gas in the safeguards world. This paper will present the methodology that includes MCNPX simulations for comparing divergent detector types such as 10B lined proportional counters with 3He gas based systems where the performance metrics focus on safeguards applications.

Nuclear instruments and methods in physics research, Feb 1, 1982

Abstract The average neutron detection efficiency of a 4.5 cm diameter, 3.0 cm thick NE 213 liqui... more Abstract The average neutron detection efficiency of a 4.5 cm diameter, 3.0 cm thick NE 213 liquid scintillator has been measured for neutron energies between 50 MeV and 450 MeV. Measurements were performed at three detector thresholds of 0.6, 4.2 and 17.5 MeV ee . The experimental results are in good agreement with the predictions of a standard Monte Carlo program for all three threshold values.

Nuclear Instruments and Methods, Sep 1, 1980

The associated particle technique, with a gas target, has been used to measure the absolute centr... more The associated particle technique, with a gas target, has been used to measure the absolute central neutron detection efficiency of two scintillators, (NE213 and NE102A) with an uncertainty of less than-+ 2%, over the energy range 1.5-25 MeV. A commercial n/3, discrimination system was used with NE213. Efficiencies for various discrimination levels were determined simultaneously by two parameter computer storage. The average efficiency of each detector was measured by scanning the neutron cone across the front face. The measurements have been compared with two Monte Carlo efficiency programs (Stanton's and 05S), without artifically fitting any parameters. When the discrimination level (in terms of proton energy) is determined from the measured light output relationship, very good agreement (to about 3%) is obtained between the measurements and the predictions. The agreement of a simple analytical expression is also found to be good over the energy range where n-p scattering dominates.

Powsner/Essentials of Nuclear Medicine Physics and Instrumentation, 2013

A scintillation detector is a device which responds to ionizing radiation by emitting light. Such... more A scintillation detector is a device which responds to ionizing radiation by emitting light. Such devices are used in many areas of scientific research and development as well as in a number of industrial applications. This article provides an overview of the physics of scintillation detectors, their production, and their uses in a number of areas of basic research and industry.

Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component i... more Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF6 cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF 6 cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

Nuclear Science and Engineering, Feb 1, 2011

Abstract Neutron coincidence measurements of plutonium samples with uncertainties <0.5% could ... more Abstract Neutron coincidence measurements of plutonium samples with uncertainties <0.5% could reduce the amount of costly destructive analysis required for nuclear material accountancy in plutonium handling plants. The ratio of (α, n) emission to spontaneous fission neutron emission, α, of plutonium samples is important to the interpretation of neutron coincidence measurements. When the “known alpha” analysis method is used, an error on the α value propagates to approximately the same percentage error on the measured plutonium mass. Molality data of Charrin and the SOURCES code have been used to update the calculation of α for both pure plutonium nitrate solutions and plutonium/uranyl nitrate solutions of different concentrations and acidity. This paper gives equations for the density of the solution as a function of heavy metal concentration and for the α weight factors that can be used in the analysis of neutron coincidence measurements.

Nuclear instruments and methods in physics research, Aug 1, 1983

A method to determine the isotopic composition of lithium compounds has been developed, with two ... more A method to determine the isotopic composition of lithium compounds has been developed, with two variations. In one, the activity of 7Be, produced when the sample is bombarded with 4.8 MeV deuterons [via the 6Li(d, n) reaction], is compared with the activity produced in a sample of the same compound with a known 6Li/VLi ratio. In the second, a cross-comparison is made between samples of different composition for both deuteron-(6Li activation) and proton-(7Li activation) induced activities of 7Be, giving 6Li/VLi ratio measurements independent of calibration. The method works well for compounds not containing hydrogen; a lower limit on the sensitivity to 6Li content was found because of residual hydrogen in the compounds investigated.

Neutron measurements are frequently used to characterize spent nuclear fuel. Curium is the primar... more Neutron measurements are frequently used to characterize spent nuclear fuel. Curium is the primary neutron source from most spent nuclear fuel materials. Recent developments in nuclear safeguards measurements of spent nuclear fuel have increased the reliance upon curium assay for materials accounting on the back end of the fuel cycle. The curium assay is used to determine the fuel composition

The second paper on the non-destructive assay techniques investigated under the Next Generation S... more The second paper on the non-destructive assay techniques investigated under the Next Generation Safeguards Initiative considers innovative instrument designs capable of producing isotope-specific responses for quantifying the fissile material content of spent nuclear fuel. The present research goal is to assess whether expected signatures can be confidently obtained in realistic spent fuel safeguards applications, contribute to the establishment of Pu inventories and determine fissile material diversions at fuel storage, handling and reprocessing facilities. In certain cases, theoretical concepts are supported by the experimental data obtained in the simplified setups. The overall effort will be concluded by nominating the most promising techniques and their combinations for the full-scale demonstration involving actual nuclear fuel assemblies. This paper provides an overview of the following 6 photon-and neutron-based techniques:

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Jul 18, 2006

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), 2009

The objective of safeguarding nuclear material is to det r diversions of significant quantities o... more The objective of safeguarding nuclear material is to det r diversions of significant quantities of nuclear materials by timely monitoring and detection. There are a variety of motivations fo r quantifying plutonium in spent fuel (SF), by means of nondestructive assay (NDA), in order to meet this goal. These motivations include the fo llowing: strengthening the capabilities of the Internati onal Atomic Energy Agencies ability to safeguard nuclear facili ties, shipper/receiver differ nee, input accountability at reprocessing faci lities and burnup credi t at repositories. Many 1 DA techniques exist for measuring signatures from SF; however, no single NDA technique can, in isolation, quanti fy elemental pl utonium in SF. A study has been undertaken to determine the best integrated combination of 13 DA techniques for characterizing Pu mass in spent fuel. This paper fo cuses on the de velopment of a passive gamma measurement system in support the spent fu el assay system. Gamma ray detection for fresh nuclear fuel focuses on gamma ray emissions that directly co inci de with the actinides of interest to the assay. For example, the 186-keV gamma ray is generall y used for 235U assay and the 384-keV complex is generally used for assayi ng pl utoni um. In sp~n t nuclear fuel , these signatures cannot be detected as the Compton continUlll1 created from the fission products dominates the si~nal in this energy range. F or SF, the measured gamma signatures fro m key fi ssion products (' 4CS, m Cs, 154Eu) are used to ascerta in burnup, cooling time, and fis sile content information. In this paper the Monte Carlo modeling setup for a passiv gamma spent fuel assay system will be described. The setup of the system incl ud s a germani um detector and an io n chamber and will be used to gain passive gamma information that will be integrated into a syst m for detemlin ing Pu in SF. The passive gamma signal will be determined from a library of ~ 100 assemblies that have be n created to examine the capability of all 13 NDA techniques. Presented in this paper is a description of the passive gamma mo nitoring instrument, explanation of the work completed thus far invol ving the source set up methodology and the design opti mizati on process, deta ils of key fission product ratios of interest, li mitations and key str ngths of the measurement technique, and considerations fo r integrating this technique with other NDA techniq ues in order to develop a complete spent fu,J assay strategy.

De Gruyter eBooks, Dec 31, 1997

The second paper on the non-destructive assay techniques investigated under the Next Generation S... more The second paper on the non-destructive assay techniques investigated under the Next Generation Safeguards Initiative considers innovative instrument designs capable of producing isotope-specific responses for quantifying the fissile material content of spent nuclear fuel. The present research goal is to assess whether expected signatures can be confidently obtained in realistic spent fuel safeguards applications, contribute to the establishment of Pu inventories and determine fissile material diversions at fuel storage, handling and reprocessing facilities. In certain cases, theoretical concepts are supported by the experimental data obtained in the simplified setups. The overall effort will be concluded by nominating the most promising techniques and their combinations for the full-scale demonstration involving actual nuclear fuel assemblies. This paper provides an overview of the following 6 photonand neutron-based techniques: Delayed Gamma, X-Ray Fluorescence, Nuclear Resonance Fluorescence, Passive Gamma, Lead Slowing-Down Spectrometry, and Neutron Resonance Transmission Analysis.

There are a variety of algorithmic approaches available to process list-mode pulse streams to pro... more There are a variety of algorithmic approaches available to process list-mode pulse streams to produce multiplicity histograms for subsequent analysis. In the development of the INCC v6.0 code to include the processing of this data format, we have noted inconsistencies in the “processed time” between the various approaches. The processed time, tp, is the time interval over which the recorded pulses are analyzed to construct multiplicity histograms. This is the time interval that is used to convert measured counts into count rates. The observed inconsistencies in tp impact the reported count rate information and the determination of the error-values associated with the derived singles, doubles, and triples counting rates. This issue is particularly important in low count-rate environments. In this report we will present a prescription for the processing of list-mode counting data that produces values that are both correct and consistent with traditional shift-register technologies. It is our objective to define conventions for list mode data processing to ensure that the results are physically valid and numerically aligned with the results from shift-register electronics.

A series of experiments using the Idaho National Laboratory (INL) photonuclear inspection system ... more A series of experiments using the Idaho National Laboratory (INL) photonuclear inspection system and a Los Alamos National Laboratory (LANL) supplied, list-mode data acquisition method have shown enhanced performance utilizing pulsed photofission-induced, neutron coincidence counting between pulses of an up-to-10-MeV electron accelerator for nuclear material detection and identification. The enhanced inspection methodology has applicability to homeland security, treaty-related support, and weapon dismantlement applications. For the latter, this technology can directly support Department of Energy/NA241 programmatic mission objectives relative to future Rocky Ridgetype testing campaigns for active inspection systems.

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), 2009

There are a variety of motivations for quantifying plutonium (Pu) in spent fuel assemblies by mea... more There are a variety of motivations for quantifying plutonium (Pu) in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capability of the International Atomic Energy Agency (LAEA) to safeguard nuclear facilities, quantifying shipperlreceiver difference, determining the input accountability value at pyrochemical processing facilities, providing quantitative input to burnup credit and final safeguards measurements at a long-term repository.

The nuclear material contained in the process equipment of a uranium enrichment plant (referred t... more The nuclear material contained in the process equipment of a uranium enrichment plant (referred to as holdup) is an important component of the overall nuclear material inventory for the plant. Accurate quantification and verification of holdup is needed to improve international safeguards and nuclear material accountancy. This is also needed for criticality safety and waste disposition. Passive neutron and gamma-ray nondestructive assay (NDA) methods are used to measure the holdup in process equipment. A key advantage of neutron measurements is that neutrons are highly penetrating and can be measured through thick walled equipment. The dominant source of neutrons in the holdup is from the reaction resulting from alpha decay when uranium is enriched. There is a considerable spread between different historic determinations of the yield from uranium which limits the accuracy of modeling and the calibration of NDA instruments. Furthermore, the compound form and presence of water also significantly affects the neutron emission rate from the holdup. This paper describes a series of experimental measurements performed at Los Alamos National Laboratory (LANL) to determine the absolute neutron emission yield from 10 different working reference materials (WRMs) fabricated at the Portsmouth Gaseous Diffusion Plant (PGDP). The Mini Epithermal Neutron Multiplicity Counter (Mini ENMC) and a NIST certified neutron source were used for these measurements. The high efficiency and short die-away time of the Mini ENMC provides the high measurement precision needed to certify the neutron emission yield. The experiment was designed to achieve sub 1% accuracy in the net counting rate on each item and to provide assurance that important factors such as instrument stability, item placement and background were well understood. The traceable neutron yields measured from the WRMs were used to determine a more accurate neutron yield for the material. The results were compared to historical neutron emission rates. The data obtained from these measurements directly supports nuclear safeguards for NDA of uranium holdup and provides a more accurate calibration for new and existing NDA detector systems.

Nuclear Instruments and Methods in Physics Research, Mar 1, 2010

Abstract Correlations are used to determine the 242 Pu content of material using high resolution ... more Abstract Correlations are used to determine the 242 Pu content of material using high resolution gamma measurements on the other plutonium isotopes because 242 Pu itself has no practically detectable gamma emission lines. This paper presents an improved correlation that is particularly useful because, unlike some previous correlations, no prior knowledge of the reactor type or initial enrichment is required. This correlation has been shown to perform well over a range of plutonium from commercial BWR and PWR reactors. The agreement of the calculated 242 Pu values with IDMS values is within 1% for 239 Pu content of less than 70% and within 4% for 239 Pu content of less than 80%. This simple form of the correlation is somewhat surprising given the complex behavior of 239/240 and 242/240 ratios.

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), May 5, 2011

During the past several years, the demand for 3He gas has far exceeded the gas supply. This short... more During the past several years, the demand for 3He gas has far exceeded the gas supply. This shortage of 3He gas is projected to continue into the foreseeable future. There is a need for alternative neutron detectors that do not require 3He gas. For more than four decades, neutron detection has played a fundamental role in the safeguarding and control of nuclear materials at production facilities, fabrication plants and storage sites worldwide. Neutron measurements for safeguards applications have requirements that are unique to the quantitative assay of special nuclear materials. These neutron systems measure the neutron multiplicity distributions from each spontaneous fission and/or induced fission event. The neutron time correlation counting requires that two or more neutrons from a single fission event be detected. The doubles and triples neutron counting rate depends on the detector efficiency to the 'L'd and :jd power, respectively, so low efficiency systems will not work for the coincidence measurements, and any detector instabilities are greatly amplified. In the current test program, we will measure the alternative detector properties including efficiency, die-away time, multiplicity precision, gamma sensitivity, dead-time, and we will also consider the detector properties that would allow commercial production to safeguards scale assay systems. This last step needs to be accomplished before the proposed technologies can reduce the demand on 3He gas in the safeguards world. This paper will present the methodology that includes MCNPX simulations for comparing divergent detector types such as 10B lined proportional counters with 3He gas based systems where the performance metrics focus on safeguards applications.

Nuclear instruments and methods in physics research, Feb 1, 1982

Abstract The average neutron detection efficiency of a 4.5 cm diameter, 3.0 cm thick NE 213 liqui... more Abstract The average neutron detection efficiency of a 4.5 cm diameter, 3.0 cm thick NE 213 liquid scintillator has been measured for neutron energies between 50 MeV and 450 MeV. Measurements were performed at three detector thresholds of 0.6, 4.2 and 17.5 MeV ee . The experimental results are in good agreement with the predictions of a standard Monte Carlo program for all three threshold values.

Nuclear Instruments and Methods, Sep 1, 1980

The associated particle technique, with a gas target, has been used to measure the absolute centr... more The associated particle technique, with a gas target, has been used to measure the absolute central neutron detection efficiency of two scintillators, (NE213 and NE102A) with an uncertainty of less than-+ 2%, over the energy range 1.5-25 MeV. A commercial n/3, discrimination system was used with NE213. Efficiencies for various discrimination levels were determined simultaneously by two parameter computer storage. The average efficiency of each detector was measured by scanning the neutron cone across the front face. The measurements have been compared with two Monte Carlo efficiency programs (Stanton's and 05S), without artifically fitting any parameters. When the discrimination level (in terms of proton energy) is determined from the measured light output relationship, very good agreement (to about 3%) is obtained between the measurements and the predictions. The agreement of a simple analytical expression is also found to be good over the energy range where n-p scattering dominates.