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All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Nuclear Data Sheets for A = 133

Evaluated nuclear structure and decay data for all nuclei within the A=133 mass chain are presented. The experimental data are evaluated and best values for level and gamma-ray energies, quantum numbers, lifetimes, gamma-ray intensities, and other nuclear properties are recommended. Inconsistencies and discrepancies that exist in the literature are noted. This work supersedes the earlier evaluation by S. Raab (1995Ra12), published in Nuclear Data Sheets75, 491 (1995).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Analysis of beta-decay rates for Ag 108, Ba 133, Eu 152, Eu 154, Kr 85, Ra 226, and Sr 90, measured at the Physikalisch-Technische Bundesanstalt from 1990 to 1996

We present the results of an analysis of measurements of the beta-decay rates of Ag 108, Ba 133, Eu 152, Eu 154, Kr 85, Ra 226, and Sr 90 acquired at the Physikalisch-Technische Bundesanstalt from 1990 through 1995. Although the decay rates vary over a range of 165 to 1 and the measured detector current varies over a range of 19 to 1, the detrended and normalized count rate measurements exhibit a sinusoidal annual variation with amplitude in the small range 0.068%-0.088% (mean 0.081%, standard deviation 0.0072%, a rejection of the zero-amplitude hypothesis) and phase-of-maximum in the small range 0.062-0.083 (January 23 to January 30). In comparing these results with those of other related experiments that yield different results, it may be significant that this experiment, at a standards laboratory, seems to be unique in using a 4π detector. These results are compatible with a solar influence, and do not appear to be compatible with an experimental or environmental influence. It is possible that Ba 133 measurements are also subject to a non-solar (possibly cosmic) influence.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2012

Nuclear model calculations on the production of {sup 125,123}Xe and {sup 133,131,129,128}Ba radioisotopes

In this study, production rates of {sup 125,123}Xe and {sup 133,131,129,128}Ba medical isotopes produced by {sup 127}I(p, 3n){sup 125}Xe, {sup 127}I(p, 5n){sup 123}Xe, {sup 133}Cs(p, n){sup 133mg}Ba, {sup 133}Cs(p, 3n){sup 131mg}Ba, {sup 133}Cs(p, 5n){sup 129}Ba, and {sup 133}Cs(p, 6n){sup 128}Ba reactions have been investigated up to 100 MeV incident proton energy. The preequilibrium calculations involve the hybrid model, the geometry-dependent hybrid model and the cascade exciton model. The calculated results are compared with the experimental data taken from the literature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Measurements of radiological data of some amino acids in the energy range 0.122-1.330MeV

Radiological parameters such as μ{sub m}, σ{sub t}, σ{sub e}, Z{sub eff} and N{sub eff} of amino acids, namely Lysine (C{sub 6}H{sub 15}N{sub 2}O{sub 2}), Histidine (C{sub 6}H{sub 9}N{sub 3}O{sub 2}) and Arginine (C{sub 6}H{sub 15}N{sub 4}O{sub 2}), were measured using NaI (Tl)-based gamma spectrometry. Radioactive sources used in the study are {sup 57}Co, {sup 133}Ba, {sup 137}Cs, {sup 54}Mn, {sup 60}Co and {sup 22}Na. Gamma ray transmission method in a narrow beam good geometry set up was used in the study. The measured data were compared against XCOM-based data. The agreement is within 2%.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Low background HPGe spectrometer in investigations of 2β decay

The low background high sensitive HPGe spectrometer called OBELIX is briefly described. The calibration measurements using {sup 152}Eu, {sup 133}Ba and La{sub 2}O{sub 3} sources in different geometries, the obtained efficiency curves for OBELIX HPGe detector, the results of measurements of radioactivity of the NEMO-3 sources ({sup 100}Mo, {sup 150}Nd) as well as future plans for OBELIX detector (e.g. 0νEC/EC decay of {sup 106}Cd) are presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Measurement of the mass attenuation coefficient from 81 keV to 1333 keV for elemental materials Al, Cu and Pb

The mass attenuation coefficients (μ/ρ) for 3 high purity elemental materials Al, Cu and Pb were measured in the γ-ray energy range from 81 keV up to 1333 keV using {sup 22}Na, {sup 60}Co {sup 133}Ba and {sup 133}Cs as sources of gamma radiation. Well shielded detector (NaI (Tl) semiconductor detector) was used to measure the intensity of the transmitted beam. The measurements were made under condition of good geometry, assuring that any photon absorbed or deflected appreciably does not reach the detector. The measured values are compared with the theoretical ones obtained by Seltzer (1993).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Optimal width of barrier region in X/{gamma}-ray Schottky diode detectors based on CdTe and CdZnTe

The spectral distribution of quantum detection efficiency of X- and {gamma}-ray Schottky diodes based on semi-insulating CdTe or Cd{sub 0.9}Zn{sub 0.1}Te crystals is substantiated and obtained in analytical form. It is shown that the width of the space charge region (SCR) of 6-40 {mu}m at zero bias in CdTe (Cd{sub 0.9}Zn{sub 0.1}Te) Schottky diode is optimal for detecting radiation in the photon energy range above 5-10 keV. Based on the Poisson equation, the relationship between the SCR width and the composition of impurities and the degree of their compensation are investigated. It is shown that the presence of deep levels in the bandgap leads to a considerable increase in space charge density and electric field strength near the crystal surface. However, this effect contributes a small error in the determination of the SCR width using the standard formula for the Schottky diode. It is also shown that the concentration of uncompensated impurities in CdTe and Cd{sub 0.9}Zn{sub 0.1}Te crystals within the 4 Multiplication-Sign 10{sup 11}-10{sup 13} cm{sup -3} range is optimal for the detection efficiency of X- and {gamma}-rays in the photon high-energy range. The record-high values of energy resolution have been obtained in the spectra of {sup 241}Am, {sup 57}Co, {sup 133}Ba and {sup 137}Cs isotopes measured using CdTe crystals with Schottky diodes because the concentration of uncompensated donors in the CdTe crystals (1-2) Multiplication-Sign 10{sup 12} cm{sup -3} falls on an interval of maximum detection efficiency. In the spectrum of {sup 57}Co isotope, the limiting energy resolution has been achieved.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2014

Determination of attenuation coefficients of heavyweight concretes containing colemanite by using {sup 133}Ba radioactive isotope source - Paper 140

As it is well known, colemanite is a boron ore. Boron is one of the most important underground richness of Turkey. Boron is employed as a constituent material of neutron shielding because of its high absorption of thermal neutrons. Barite ore includes BaSO{sub 4}, which is a well photon radiation absorbent. The literature about radiation attenuation in heavy concretes (including boron and barite ores) is fairly abundant. Abdo and Abdo et al. theoretically calculated and determined cross section by using XCOM computer program for photon and neutron radiations shielding. Akkurt et al. measured radiation transmission of heavy concretes including normal and barite aggregates for different γ-ray energies and calculated linear attenuation coefficients. Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite as control absorber in nuclear reactors γ-rays and neutron particles shielding. Bashter and Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite, basalt-magnetite, ilmenite, basalt, steel and magnetite for only photon radiation shielding and calculated linear and mass attenuation coefficients from 10 keV to 1 GeV. Dealmeid et al. studied radiation transmission and angular distribution for 25 MV X-rays by using LINAC. Demir and Keles measured radiation transmission of concrete including boron waste for 59.54 and 80.99 keV γ-rays. Kaplan reported about many heavy concretes studies in his book. Kitis et al. studied heavy concretes including barite aggregates for cyclotron shielding. Mollah et al. determined concretes including ilmenite and magnetite aggregates for neutron shielding by using {sup 252}Cf radioactive source and BF{sub 3} neutron detector. Yarar and Bayuelken studied concretes including colemanite aggregates for neutron shielding. In this paper, we have measured radiation transmission of thirteen different heavy concretes produced by using barite, colemanite and normal aggregates for 276, 303, 356 and 383 keV γ-ray energies of {sup 133}Ba radioactive isotope by using CdTe semiconductor detector and calculated linear and mass attenuation coefficients. Results were compared to X-ray mass attenuation coefficient values of NIST. The attenuation of γ-ray photons is governed primarily by the atomic number and density of the concrete as a shielding material. Photon attenuation depends on pair production, photoelectric effect and Compton scattering. In this study, linear attenuation coefficients (μ) decrease with the increasing the photon energy for these concretes. Heavy elements there is a fairly sharp minimum in the absorption cross section at about this energy region. It is indicated that attenuation coefficient is useful parameter for radiation shielding. As known, the linear attenuation coefficients of the concretes used in this study at these γ-ray energies are not available in the literature by using CdTe semiconductor detectors. The data shows the effects of including boron-rich colemanite or barium-rich barite in addition to normal aggregate in the concrete mixtures. As would be expected, the barite is desirable for photon attenuation, while the colemanite is desirable for neutron attenuation. The concretes including barite and colemanite would be preferred as materials for building construction of radiation oncology departments, radiology departments, nuclear reactors and radiation research laboratories against photon and neutron particle radiations. But other aggregate properties of colemanite such as abrasion resistance and freeze-thaw durability are also found to be important for concrete but they cannot be quantified at this study. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2014

New approach for calibration the efficiency of HpGe detectors

This work evaluates the efficiency calibrating of HpGe detector coupled with Canberra GC3018 with Genie 2000 software and Ortec GEM25-76-XLB-C with Gamma Vision software; available at Neutron activation analysis laboratory in Malaysian Nuclear Agency (NM). The efficiency calibration curve was constructed from measurement of an IAEA, standard gamma–point sources set composed by {sup 214}Am, {sup 57}Co, {sup 133}Ba, {sup 152}Eu, {sup 137}Cs and {sup 60}Co. The efficiency calibrations were performed for three different geometries: 5, 10 and 15 cm distances from the end cap detector. The polynomial parameters functions were simulated through a computer program, MATLAB in order to find an accurate fit to the experimental data points. The efficiency equation was established from the known fitted parameters which allow for the efficiency evaluation at particular energy of interest. The study shows that significant deviations in the efficiency, depending on the source-detector distance and photon energy.

42 ENGINEERING↗
DOE OSTI2016

New approach to calculate the true-coincidence effect of HpGe detector

The corrections for true-coincidence effects in HpGe detector are important, especially at low source-to-detector distances. This work established an approach to calculate the true-coincidence effects experimentally for HpGe detectors of type Canberra GC3018 and Ortec GEM25-76-XLB-C, which are in operation at neutron activation analysis lab in Malaysian Nuclear Agency (NM). The correction for true-coincidence effects was performed close to detector at distances 2 and 5 cm using {sup 57}Co, {sup 60}Co, {sup 133}Ba and {sup 137}Cs as standard point sources. The correction factors were ranged between 0.93-1.10 at 2 cm and 0.97-1.00 at 5 cm for Canberra HpGe detector; whereas for Ortec HpGe detector ranged between 0.92-1.13 and 0.95-100 at 2 and 5 cm respectively. The change in efficiency calibration curve of the detector at 2 and 5 cm after correction was found to be less than 1%. Moreover, the polynomial parameters functions were simulated through a computer program, MATLAB in order to find an accurate fit to the experimental data points.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Decommissioning and PIE of the MEGAPIE spallation target

A key experiment in the Accelerated Driven Systems roadmap, the MEGAwatt PIlot Experiment (MEGAPIE) (1 MW) was initiated in 1999 in order to design and build a liquid lead-bismuth spallation target, then to operate it into the Swiss spallation neutron facility SINQ at Paul Scherrer Institute. The target has been designed, manufactured, and tested during integral tests, before irradiation carried out end of 2006. During irradiation, neutron and thermo hydraulic measurements were performed allowing deep interpretation of the experiment and validation of the models used during design phase. The decommissioning, Post Irradiation Examinations and waste management phases were defined properly. The phases dedicated to cutting, sampling, cleaning, waste management, samples preparation and shipping to various laboratories were performed by PSI teams: all these phases constitute a huge work, which allows now to perform post-irradiation examination (PIE) of structural material, irradiated in relevant conditions. Preliminary results are presented in the paper, they concern chemical characterization. The following radio-nuclides have been identified by γ-spectrometry: {sup 60}Co, {sup 101}Rh, {sup 102}Rh, {sup 108m}Ag, {sup 110m}Ag, {sup 133}Ba, {sup 172}Hf/Lu, {sup 173}Lu, {sup 194}Hg/Au, {sup 195}Au, {sup 207}Bi. For some of these nuclides the activities can be easily evaluated from γ-spectrometry results ({sup 207}Bi, {sup 194}Hg/Au), while other nuclides can only be determined after chemical separations ({sup 108m}Ag, {sup 110m}Ag, {sup 195}Au, {sup 129}I, {sup 36}Cl and α-emitting {sup 208-210}Po). The concentration of {sup 129}I is lower than expected. The chemical analysis already performed on spallation and corrosion products in the lead-bismuth eutectic (LBE) are very relevant for further applications of LBE as a spallation media and more generally as a coolant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Instrument for precision long-term β-decay rate measurements

We describe an experimental setup for making precision measurements of relative β-decay rates of {sup 22}Na, {sup 36}Cl, {sup 54}Mn, {sup 60}Co, {sup 90}Sr, {sup 133}Ba, {sup 137}Cs, {sup 152}Eu, and {sup 154}Eu. The radioactive samples are mounted in two automated sample changers that sequentially position the samples with high spatial precision in front of sets of detectors. The set of detectors for one sample changer consists of four Geiger-Müller (GM) tubes and the other set of detectors consists of two NaI scintillators. The statistical uncertainty in the count rate is few times 0.01% per day for the GM detectors and about 0.01% per hour on the NaI detectors. The sample changers, detectors, and associated electronics are housed in a sealed chamber held at constant absolute pressure, humidity, and temperature to isolate the experiment from environmental variations. The apparatus is designed to accumulate statistics over many years in a regulated environment to test recent claims of small annual variations in the decay rates. We demonstrate that absent this environmental regulation, uncontrolled natural atmospheric pressure variations at our location would imprint an annual signal of 0.1% on the Geiger-Müller count rate. However, neither natural pressure variations nor plausible indoor room temperature variations cause a discernible influence on our NaI scintillator detector count rate.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

A New Light Weight Structural Material for Nuclear Structures

Radiation shielding materials are commonly used in nuclear facilities to attenuate the background ionization radiations to a minimum level for creating a safer workplace, meeting regulatory requirements and maintaining high quality performance. The conventional radiation shielding materials have a number of drawbacks: heavy concrete contains a high amount of elements that are not desirable for an effective shielding such as oxygen, silicon, and calcium; a well known limitation of lead is its low machinability and toxicity, which is causing a major environmental concern. Therefore, an effective and environmentally friendly shielding material with increased attenuation and low mass density is desirable. Close-cell composite metal foams (CMFs) and open-cell Al foam with fillers are light-weight candidate materials that we have studied in this project. Close-cell CMFs possess several suitable properties that are unattainable by conventional radiation shielding materials such as low density and high strength for structural applications, high surface area to volume ratio for excellent thermal isolation with an extraordinary energy absorption capability. Open-cell foam is made up of a network of interconnected solid struts, which allows gas or fluid media to pass through it. This unique structure provided a further motive to investigate its application as radiation shields by infiltrating original empty pores with high hydrogen or boron compounds, which are well known for their excellent neutron shielding capability. The resulting open-cell foam with fillers will not only exhibit light weight and high specific surface area, but also possess excellent radiation shielding capability and good processability. In this study, all the foams were investigated for their radiation shielding efficiency in terms of X-ray, gamma ray and neutron. X-ray transmission measurements were carried out on a high-resolution microcomputed tomography (microCT) system. Gamma-emitting sources: 3.0mCi 60 Co, 1.8mCi 137 Cs, 13.5mCi 241 Am, and 5.0mCi 133 Ba were used for gamma-ray attenuation analysis. The evaluations of neutron transmission measurements were conducted at the Neutron Powder Diffractometer beam facility at North Carolina State University. The experimental results were verified theoretically through XCOM and Monte Carlo Z-particle Transport Code (MCNP). A mechanical investigation was performed by means of quasi-static compressive testing. Thermal characterizations were carried out through effective thermal conductivity and thermal expansion analyses in terms of high temperature guarded-comparative-longitudinal heat flow technique and thermomechanical analyzer (TMA), respectively. The experimental results were compared with analytical results obtained from, respectively, Brailsford and Major’s model and modified Turner’s model for verification. Flame test was performed in accordance with United States Nuclear Regulatory Commission (USNRC) standard. CMF sample and a 304L stainless steel control sample were subjected to a fully engulfing fire with an average flame temperature of 800°C for a period of 30 minutes. Finite Element Analysis was conducted to secure the credibility of the experimental results. This research indicates the potential of utilizing the light-weight close-cell CMFs and open-cell Al foam with fillers as shielding material replacing current heavy structures with additional advantage of high-energy absorption and excellent thermal characteristics.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Radioisotope identification method for poorly resolved gamma-ray spectrum of nuclear security concern

Gamma-ray signal can be used as a fingerprint for radioisotope identification. In the context of radioactive and nuclear materials security at the border control point, the detection task can present a significant challenge due to various constraints such as the limited measurement time, the shielding conditions, and the noise interference. This study proposes a novel method to identify the signal of one or several radioisotopes from a poorly resolved gamma-ray spectrum. In this method, the noise component in the raw spectrum is reduced by the wavelet decomposition approach, and the removal of the continuum background is performed using the baseline determination algorithm. Finally, the identification of radioisotope is completed using the matrix linear regression method. The proposed method has been verified by experiments using the poorly resolved gamma-ray signals from various scenarios including single source, mixing of natural uranium with five of the most common industrial radioactive sources (57Co, 60Co, 133Ba, 137Cs, and 241Am). The preliminary results show that the proposed algorithm is comparable with the commercial method.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

New Advanced Source Identification Algorithm (ASIA-NEW) for radiation monitors with plastic detectors

Radiation Portal Monitors (RPM) with plastic detectors represent the main instruments used for primary border (customs) radiation control. RPM are widely used because they are simple, reliable, relatively inexpensive and have a high sensitivity. However, experience using the RPM in various countries has revealed the systems have some grave shortcomings. There is a dramatic decrease of the probability of detection of radioactive sources under high suppression of the natural gamma background (radiation control of heavy cargoes, containers and, especially, trains). NORM (Naturally Occurring Radioactive Material) existing in objects under control trigger the so-called 'nuisance alarms', requiring a secondary inspection for source verification. At a number of sites, the rate of such alarms is so high it significantly complicates the work of customs and border officers. This paper presents a brief description of new variant of algorithm ASIA-New (New Advanced Source Identification Algorithm), which was developed by the authors and based on some experimental test results. It also demonstrates results of different tests and the capability of a new system to overcome the shortcomings stated above. New electronics and ASIA-New enables RPM to detect radioactive sources under a high background suppression (tested at 15-30%) and to verify the detected NORM (KCl) and the artificial isotopes (Co-57, Ba-133 and other). New variant of ASIA is based on physical principles and does not require a lot of special tests to attain statistical data for its parameters. That is why this system can be easily installed into any RPM with plastic detectors. This algorithm was tested for 1,395 passages of different transports (cars, trucks and trailers) without radioactive sources. It also was tested for 4,015 passages of these transports with radioactive sources of different activity (Co-57, Ba-133, Cs-137, Co-60, Ra-226, Th-232) and these sources masked by NORM (K-40) as well. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Real breakthrough in detection of radioactive sources by portal monitors with plastic detectors and New Advanced Source Identification Algorithm (ASIA-New)

Radiation Portal Monitors (RPM) with plastic detectors represent the main instruments used for primary border (customs) radiation control. RPM are widely used because they are simple, reliable, relatively inexpensive and have a high sensitivity. However, experience using the RPM in various countries has revealed the systems have some grave shortcomings. There is a dramatic decrease of the probability of detection of radioactive sources under high suppression of the natural gamma background (radiation control of heavy cargoes, containers and, especially, trains). NORM (Naturally Occurring Radioactive Material) existing in objects under control trigger the so-called 'nuisance alarms', requiring a secondary inspection for source verification. At a number of sites, the rate of such alarms is so high it significantly complicates the work of customs and border officers. This paper presents a brief description of new variant of algorithm ASIA-New (New Advanced Source Identification Algorithm), which was developed by the Rapiscan company. It also demonstrates results of different tests and the capability of a new system to overcome the shortcomings stated above. New electronics and ASIA-New enables RPM to detect radioactive sources under a high background suppression (tested at 15-30%) and to verify the detected NORM (KCl) and the artificial isotopes (Co- 57, Ba-133 and other). New variant of ASIA is based on physical principles, a phenomenological approach and analysis of some important parameter changes during the vehicle passage through the monitor control area. Thanks to this capability main advantage of new system is that this system can be easily installed into any RPM with plastic detectors. Taking into account that more than 4000 RPM has been installed worldwide their upgrading by ASIA-New may significantly increase probability of detection and verification of radioactive sources even masked by NORM. This algorithm was tested for 1,395 passages of different transports (cars, trucks and trailers) without radioactive sources. It also was tested for 4,015 passages of these transports with radioactive sources of different activity (Co-57, Ba-133, Cs-137, Co-60, Ra-226, Th-232) and these sources masked by NORM (K-40) as well. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Further Development of an ISOCS - based Advanced In Situ Gamma Spectrometry Services Tool for Waste Measurements - 17452

Canberra's In Situ Object Counting System is an established tool used for gamma spectrometry allowing a physical representation of complex geometries and mathematical calculation of the calibration function while avoiding the need for radioisotope standards. The ISOCS Uncertainty Estimator tool can be used to calculate defensible systematic calibration uncertainties. However the uncertainties depend on knowledge of the properties of the item being measured and for poorly characterized materials these can be large, often leading to large costs for waste sentencing. An ISOCS-based advanced in-situ gamma spectrometry services tool has been developed to reduce those uncertainties. This system is based on generating and comparing a range of candidate geometry models that yield figures of merit indicative of improved consistency between modelled data and available diverse measurement data. The approach and results of initial performance testing have previously been published, demonstrating an analysis procedure and enhanced accuracy for measurement of a low density 200 litre waste drum, based on measurements with known sources placed at different positions. It was shown how the accuracy of assay of Am-241 in a light matrix can be significantly improved. Further analysis has now been performed, with reference to measurements of a second drum, containing a standard matrix of higher density, and with a multiple radionuclide line source (Ba-133, Cs-137, Eu-152 and Am-241). These measurements represent a more challenging waste assay scenario, corresponding to potentially much larger assay uncertainties. Our results show how the assay uncertainty can be reduced, by careful application of AIGS, from relative values of the order of 100% to a few 10's of %. Furthermore, we demonstrate, through extended analysis of the test drum data, how an AIGS measurement program can be optimized to reduce the number of required measurement positions substantially. This paper describes these new results with comment on their importance and potential application for real waste assay projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Design and testing of a combined neutron and gamma assay system for measuring fissile material in fuel and waste items - 15340

A combined neutron and gamma ray assay system for the measurement of fissile material in both fuel and waste items has been designed, modelled, developed and tested. The purpose of the system is to measure and characterize a wide range of fissile materials including uranium, plutonium, americium and mixed oxide fuel arising from a variety of processes and facilities across the United Kingdom. Dounreay Site Restoration Limited at Dounreay currently stores these items form the United Kingdom Nuclear Decommissioning Authority's un-irradiated fuel inventory. This inventory comprises historic materials from ex-United Kingdom Atomic Energy Authority nuclear fuel projects and experiments together with fast reactor fuel materials and some commercial fuel items as delivered to Dounreay for recovery. The combined assay system has been designed and constructed to support the operation of a fuel processing facility, which will be operated at Dounreay to treat these historic legacy fuel materials, making them suitable for long term storage. Appropriate characterization of the fissile material is required for safety, criticality control and nuclear materials safeguards purposes prior to final disposition. The system consists of both gamma ray and neutron measurement components. The results of a neutron measurement are combined with the gamma ray isotopic ratio measurement results to generate the assay result for the fissile material being characterized. The gamma ray component of the system employs a safeguards quality high purity Germanium detector with electro-mechanical cooling and the measurements of fissile material containers are made in a shielded chamber with a tin and copper graded lining. Containers of fissile material are rotated during the gamma ray measurement and a variable automatic steel collimator aperture mechanism is employed to adjust the measured count rate to control detector dead time. Gamma ray spectra are analysed using the plutonium and uranium isotopic ratio code PC/FRAM. The gamma ray system can also be configured to perform a direct assay of small quantities of uranium bearing material. For this assay measurement a series of cadmium filters can be positioned in front of the detector to reduce the contribution from 59 keV gamma rays from Am-241. In addition to gamma ray isotopic measurements, both active and passive neutron assay is performed by the system. The neutron-measuring component consists of a thick walled cylindrical polyethylene chamber, which provides neutron moderation for 30 one-inch diameter He-3 detector tubes located in re-entrant channels in the polyethylene moderator. Two removable polyethylene-inserts incorporating lower plug assemblies are provided, one with and one without a cadmium lining. Also two top plug units are provided, again one with cadmium and one without a cadmium liner. The Neutron Subsystem can be operated both as a passive and as an active neutron coincidence counter. In both fast and thermal active neutron mode, the neutron component employs two americium-beryllium (Am-Be) alpha-n neutron sources, one positioned in the top and the other in the bottom plug unit. The neutron sources are removed when the counter is operated in passive mode. Passive neutron detection efficiency and active neutron performance has been modelled using the MCNP Monte Carlo computer simulation code as part of the design process. In the absence of fissile samples, operation of the gamma ray component of the system has been tested initially using calibrated Eu-152 and Ba-133 gamma ray sources and the use of archived spectra from fissile material measurements. Initial testing of the neutron system has employed a calibrated Cf-252 spontaneous fission neutron source. Neutron testing has been confirmed using further MCNP simulations. The results of test measurements with the system are presented and compared with the results of MCNP measurement simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗