Measurement Of Nuclear Materials Research Articles

New data processing method for nuclear material measurement using pulsed neutrons

Active neutron methods using pulsed neutrons are used to detect the presence of nuclear materials. In this study, we investigated a new data processing method and compared its performance to a conventional method. It is shown that the new method offers a significant reduction in measurement time and an increased sensitivity.

Model independent analysis in (γ, n) reactions using deuterium targets

Photonuclear reactions play an important role in nuclear physics, astrophysics and in various applications such as non-destructive measurement of nuclear materials (NDT). The study of (γ, n) reactions using deuterium targets i.e., photodisintegration of deuterons in addition to all the other (γ, n) reactions, is of considerable interest to these fields. In this contribution, we have studied the photodisintegration of deuterons with unpolarized photons. The angular dependence of the differential cross section is studied by expressing it in terms of Legendre polynomials. The analysis of differential cross-section is presented using the model-independent irreducible tensor formalism.

Estimation of the 234U fraction and uncertainty of enriched and depleted uranium

This paper provides a simple model for estimating the 234U fraction and uncertainty for enriched and depleted uranium produced by modern commercial gaseous centrifuge enrichment plants. Accurate estimation of the 234U fraction of uranium has applications to the nuclear engineering fields of radiation protection, nuclear materials measurement, transportation safety, criticality safety, and safeguards. Significantly, the 234U fraction and uncertainties drive radiation safety limits and dose estimates related to uranium and uranium facilities (excluding recycled uranium).

Suggested Modification of the Fundamental Parameter Method: A Case Study to Calculate the Optimum Absolute Intensity of 1001.03 keV Gamma-Ray Transition-

Abstract. The fundamental parameter method (FPM) is an analytical approach for intrinsic calibration of gamma-ray spectrometer using fundamental nuclear and atomic parameters such as gamma-ray branching intensity, half-life time, isotopic ratio and concentration ratio. The main advantage of this approach is the wide range of its applications in gamma-ray efficiency calibration, nuclear safeguards (nuclear materials measurement and isotopic ratios) and others. In this work, the calculation of relative efficiency (RE; photopeak count rate divided by branching ratio) was modified based on the relative intensity concept of 226Ra in equilibrium with 222Rn decay products (214Bi-214Pb). The modified FPM was applied to reevaluate the absolute intensity (I%) of gamma-ray transition of 234mPa at 1001.03 keV. using certified uranium ore samples, 226Ra point source and gamma-ray spectrometers based on hyper pure germanium detector. The newly confirmed I % of 1001.03 keV is 1.0164±0.0636. Key words: fundamental parameter method/ 1001.03 keV/ gamma-ray transition/ 238U

In-Situ object calibration software (ISOCS) technique for 235U mass verification

Measurement of nuclear materials by ISOCS technique an important and significant tool for IAEA activities verification. ISOCS technique able to measuring samples with different geometry. In this work, ISOCS technique was used to calculate the absolute efficiency of the Hyper Pure Germanium (HPGe) detector. The calculated absolute efficiency of the detector was used in combination with experimental work to determine U-235 mass in cylindrical NM samples. A good agreement was obtained when compared estimated and declared values of 235U mass content with difference in the ranged from −0.827% to 1.635%.

Study on Thallium Bromide Radioactive Detector

Thallium bromide (TlBr) is a compound semiconductor material, which can be used for X-ray and gamma-ray detectors and can be used at room temperature. It has excellent physical properties, high atomic number and density, wide bandgap (B = 2.68 eV), and low ionization energy. Compared with other X-ray and gamma-ray detection materials, TlBr devices have high detection efficiency and excellent energy resolution performance. So TlBr is suitable for housing in small tubes or shells, and it can be widely used in nuclear material measurement, safeguards verification, national security, space high-energy physics research, and other fields. Based on the fabrication of TlBr prototype detector, this paper focuses on the device fabrication and signal acquisition technology. Gamma-ray spectrum measurements and performance tests are carried out with AM-241 radioactive source. The results show that the special photoelectric peak of 59.5 keV is clearly visible, and the optimal resolution is 4.15 keV (7%).

Optimizing Nuclear Material Accounting and Measurement Systems

The ability to create nuclear weapons from 235U and 239Pu makes it imperative to closely account for these materials as they progress through a nuclear fuel cycle. Improved measurement systems provide more accurate estimates of material quantities and material unaccounted for (MUF). This paper provides examples of how two safeguards computational toolboxes can optimize and analyze hypothetical nuclear fuel cycle scenarios. The NUclear Measurement System Optimization (NUMSO) toolbox uses operations research techniques to find optimal solutions to safeguards measurement problems based on minimizing the variance of the estimated MUF. The SafeGuards Analysis (SGA) toolbox employs Monte Carlo techniques to analyze a given configuration of measurement methods and material flows to determine the probabilities of Type I (false detection) and Type II (missed detection) errors. Applying these toolboxes to a realistic fuel cycle scenario demonstrates the capability of NUMSO and SGA to address nuclear safeguards problems. Working in tandem, both toolboxes are able to determine how to quickly improve upon an existing safeguards measurement system and to calculate the resulting improvement in the error probabilities of the system. This information shows engineers not only how to develop new measurement systems but also how to improve existing systems in the most efficient manner.

Development of prototype nuclear forensics library for nuclear materials and radioisotopes in Japan Atomic Energy Agency

In 2010, the Japanese Government issued the national statement at Nuclear Security Summit (Washington D.C., USA) to develop technologies related to measurement and detection of nuclear materials for nuclear forensics, and to share them with the international community, in order to contribute to strengthening the nuclear security system. In response to this statement, Integrated Support Center for Nuclear Non-proliferation and Nuclear Security (ISCN) of Japan Atomic Energy Agency (JAEA) has initiated R&Ds on nuclear forensics technical capabilities since Japanese Fiscal Year of 2011. One of the main R&D topics of this project is to develop prototype nuclear forensics library (NFL), which is an organized collection of data and information about nuclear and other radioactive materials.In this paper, the development of prototype NFL for nuclear materials and radioisotopes will be presented. An NFL plays an important role to compare the seized materials with the characteristics of known materials to provide information about a material’s origin and history. Recently, the development of nuclear forensics library has been carried out in some countries and the concept of national NFL to collect all target material in the country is the most popular in current international society. The prototype NFL in JAEA also follows this concept. The design of database in an NFL depends strongly on the scale of nuclear fuel cycle in one country, which refers to the amount and variety of the materials available. The database and its structure of the prototype library were studied and designed based on data of target materials possessed in JAEA. This is because the target materials and fuel cycle facilities in JAEA can cover almost all stages of nuclear fuel cycle. The prototype library developed in JAEA and its experiences could contribute to future national library in our country and also for other countries which intends to develop NFLs as a model of NFL development.

Comparison of experimental results of a Quad-CZT array detector, a NaI(Tl), a LaBr3(Ce), and a HPGe for safeguards applications

A compound semiconductor detector, CdTe (or CdZnTe), has been used in various areas including nuclear safeguards applications. To address its critical drawback, low detection efficiency, which leads to a long measurement time, a Quad-CZT array-based gamma-ray spectrometer in our previous study has been developed by combining four individual CZT detectors. We have re-designed the developed Quad-CZT array system to make it more simple and compact for a hand-held gamma-ray detector. The objective of this paper aims to compare the improved Quad-CZT array system with the traditional gamma-ray spectrometers (NaI(Tl), LaBr3(Ce), HPGe); these detectors currently have been the most commonly used for verification of nuclear materials. Nuclear materials in different physical forms in a nuclear facility of Korea were measured by the Quad-CZT array system and the existing gamma-ray detectors. For measurements of UO2 pellets and powders, and fresh fuel rods, the Quad-CZT array system turned out to be superior to the NaI(Tl) and LaBr3(Ce). For measurements of UF6 cylinders with a thick wall, the Quad-CZT array system and HPGe gave similar accuracy under the same measurement time. From the results of the field tests conducted, we can conclude that the improved Quad-CZT array system would be used as an alternative to HPGes and scintillation detectors for the purpose of increasing effectivenss and efficiency of safeguards applications. This is the first paper employing a multi-element CZT array detector for measurement of nuclear materials—particularly uranium in a UF6 cylinder—in a real nuclear facility. The present work also suggests that the multi-CZT array system described in this study would be one promising method to address a serious weakness of CZT-based radiation detection.

A Novel Technique for Raman Analysis of Highly Radioactive Samples Using Any Standard Micro-Raman Spectrometer

A novel approach for the Raman measurement of nuclear materials is reported in this paper. It consists of the enclosure of the radioactive sample in a tight capsule that isolates the material from the atmosphere. The capsule can optionally be filled with a chosen gas pressurized up to 20 bars. The micro-Raman measurement is performed through an optical-grade quartz window. This technique permits accurate Raman measurements with no need for the spectrometer to be enclosed in an alpha-tight containment. It therefore allows the use of all options of the Raman spectrometer, like multi-wavelength laser excitation, different polarizations, and single or triple spectrometer modes. Some examples of measurements are shown and discussed. First, some spectral features of a highly radioactive americium oxide sample (AmO2) are presented. Then, we report the Raman spectra of neptunium oxide (NpO2) samples, the interpretation of which is greatly improved by employing three different excitation wavelengths, 17O doping, and a triple mode configuration to measure the anti-stokes Raman lines. This last feature also allows the estimation of the sample surface temperature. Finally, data that were measured on a sample from Chernobyl lava, where phases are identified by Raman mapping, are shown.

A Novel Technique for Raman Analysis of Highly Radioactive Samples Using Any Standard Micro-Raman Spectrometer

A novel approach for the Raman measurement of nuclear materials is reported in this paper. It consists of the enclosure of the radioactive sample in a tight capsule that isolates the material from the atmosphere. The capsule can optionally be filled with a chosen gas pressurized up to 20 bars. The micro-Raman measurement is performed through an optical-grade quartz window. This technique permits accurate Raman measurements with no need for the spectrometer to be enclosed in an alpha-tight containment. It therefore allows the use of all options of the Raman spectrometer, like multi-wavelength laser excitation, different polarizations, and single or triple spectrometer modes. Some examples of measurements are shown and discussed. First, some spectral features of a highly radioactive americium oxide sample (AmO2) are presented. Then, we report the Raman spectra of neptunium oxide (NpO2) samples, the interpretation of which is greatly improved by employing three different excitation wavelengths, 17O doping, and a triple mode configuration to measure the anti-stokes Raman lines. This last feature also allows the estimation of the sample surface temperature. Finally, data that were measured on a sample from Chernobyl lava, where phases are identified by Raman mapping, are shown.

Design of an electron-accelerator-driven compact neutron source for non-destructive assay

The threat of nuclear and radiological terrorism remains one of the greatest challenges to international security, and the threat is constantly evolving. In order to prevent nuclear terrorism, it is important to avoid unlawful import of nuclear materials, such as uranium and plutonium. Development of technologies for non-destructive measurement, detection and recognition of nuclear materials is essential for control at national borders. At Tokyo Institute of Technology, a compact neutron source system driven by an electron-accelerator has been designed for non-destructive assay (NDA). This system is composed of a combination of an S-band (2.856GHz) RF-gun, a tungsten target to produce photons by bremsstrahlung, a beryllium target, which is suitable for use in generating neutrons because of the low threshold energy of photonuclear reactions, and a moderator to thermalize the fast neutrons. The advantage of this system can accelerate a short pulse beam with a pulse width less than 1μs which is difficult to produce by neutron generators. The amounts of photons and neutron produced by electron beams were simulated using the Monte Carlo simulation code PHITS 2.82. When the RF-gun is operated with an average electron beam current of 0.1mA, it is expected that the neutron intensities are 1.19×109n/s and 9.94×109n/s for incident electron beam energies of 5MeV and 10MeV, respectively.

Hot-Test Results of the Advanced Spent Fuel Conditioning Process Safeguards Neutron Counter for PWR Spent Fuel Rods

A passive neutron coincidence counter for nuclear material measurement of the advanced spent fuel conditioning process (ACP) has been developed by the Korea Atomic Energy Research Institute (KAERI) since 2003 and was deployed in a hot cell of the ACP Facility (ACPF) in 2005. The most dominant neutron source among the spontaneous fission nuclides contained in spent fuel is 244Cm. To obtain the neutron counting rates of the singles, doubles, and triples coincidences of the neutron counter with an increment of the 244Cm mass, a hot test of the neutron counter was performed in 2007 with several spent fuel rod-cuts in the ACPF hot cell. The source term of the spent fuel rod-cuts was obtained using the ORIGEN-ARP burnup simulation code, and a set of preliminary calibration curves of the neutron counter for 244Cm was generated. The calibration curves were also obtained from the results of an MCNPX code simulation, but there was a wide difference of ˜30% in the slope of the double-rate calibration curve between the measurements and the MCNPX results. Chemical analysis results of the spent fuel samples were obtained in September 2008, and it was found that the difference between the measurements and the MCNPX results is due to an error in the declared burnup since the chemical analysis burnups of the samples differ from the declared ones by ˜10%. The expected burnup of each rod-cut was also obtained from the results of self-multiplication correction for the 244Cm mass of the rod-cuts, and the difference between the expected burnup results and the chemical analysis results is <2%. This study shows high performance of the neutron coincidence counter for 244Cm measurements of spent fuel and also shows that the burnup of spent fuel samples can be obtained through a series of ORIGEN-ARP code simulations if it is possible to acquire the measurement data of neutron counting rates for 244Cm of the samples.

Large microcalorimeter arrays for high-resolution X- and gamma-rayspectroscopy

Abstract Microcalorimeter detectors provide unprecedented energy resolution for the measurement of X-rays and soft gamma-rays. Energy resolution in the 100 keV region can be up to an order of magnitude better than planar high-purity germanium (HPGe) detectors. The technology is well-suited to analysis of materials with complex spectra presenting closely spaced photopeaks. One application area is the measurement and assay of nuclear materials for safeguards and fuel cycle applications. In this paper, we discuss the operation and performance of a 256-pixel array, and present results of a head-to-head comparison of isotopic determination measurements with high-purity germanium using a plutonium standard. We show that the uncertainty of a single measurement is smaller for the microcalorimeter data compared to the HPGe data when photopeak areas are equal. We identify several key areas where analysis codes can be optimized that will likely lead to improvement in the microcalorimeter performance.

Microcalorimeter arrays for ultra-high energy resolution X- and gamma-ray detection

Microcalorimeter detectors provide superior energy resolution for the detection of X-rays and gamma-rays. The technology utilizes a cryogenic transition-edge sensor (TES) coupled to a tin bulk absorber. We are working on fabrication methods for the production of arrays with many sensors. In this paper, we present data collected with an array of microcalorimeters using as many as 26 sensor elements simultaneously. Advances in sensor design have extended the useful dynamic range to photon energies up to ∼200 keV, while providing resolution performance in the 80–90 eV FWHM range, significantly better than planar high-purity germanium. These sensor arrays have applications in the measurement of nuclear materials. We present data collected from 153Gd, a highly-enriched uranium sample, and a plutonium isotopic standard source. We also demonstrate clean separation of the 235U 185.715 keV peak from the ubiquitous 226Ra 186.211 keV background peak interference.

Fissile and fertile nuclear material measurements using a new differential die-away self-interrogation technique

This paper presents a new technique for the measurement of fissile and fertile nuclear materials in spent fuel and plutonium-laden materials such as mixed oxide (MOX) fuel. The technique, called differential die-away self-interrogation, is similar to traditional differential die-away analysis, but it does not require a pulsed neutron generator or pulsed beam accelerator, and it can measure the fertile mass in addition to the fissile mass. The new method uses the spontaneous fission neutrons from 244Cm in spent fuel and 240Pu effective neutrons in MOX as the “pulsed” neutron source, with an average of ∼2.7 neutrons per pulse. The time-correlated neutrons from the spontaneous fission and the subsequent induced fissions are analyzed as a function of time to determine the spontaneous fission rate, the induced fast-neutron fissions, and the induced thermal-neutron fissions. The fissile mass is determined from the induced thermal-neutron fissions that are produced by reflected thermal neutrons that originated from the spontaneous fission reaction. The sensitivity of the fissile mass measurement is enhanced by the use of two measurements, with and without a cadmium liner between the sample and a hydrogenous moderator that surrounds the sample. The fertile mass is determined from the multiplicity analysis of the neutrons detected soon after the initial triggering neutron is detected. The method obtains good sensitivity by the optimal design of two different neutron die-away regions: a short die-away for the neutron detector region and a longer die-away for the sample interrogation region.

Analysis of A Nuclear Material Accounting Uncertainty for ACP Spent Fuel Standards: MCNPX and ORIGEN-S Code Simulations

The target nuclide to be measured by using a neutron coincidence counter is Cm-244 for the PWR spent fuel since the dominant neutron-emitting source is Cm-244 for the spent fuel. Spent fuel standards (SFS) are usually used to calibrate the neutron coincidence counter for nuclear material measurements of spent fuel. To make the best selection and minimize the measurement uncertainty of neutron coincidence counter for the SFS, expected measurement errors for several kinds of burnups were obtained by using the ORIGEN-S code and the expected singles and doubles count rates were also acquired by using MCNPX code. The expected doubles count errors for 1 pellet of spent fuel are less than 4% for a burnup from 27 to 40 GWD/MtU. Considering everything, a higher burnup of the spent fuel seems to be better to obtain more accurate data from a nuclear material measurement by using a neutron coincidence counter than that of a lower burnup.

The Effect of the Neutron Multiplication Factor on a Nuclear Material Measurement in Dry-Processed Spent Fuel Material

This study addresses the neutron multiplication which could occur in the assay sample in dry-processed spent fuel material. The neutron multiplication may cause measurement error in case of accounting by a coincident neutron counter. In this study, the multiplication effects in the dry processed DUPIC pellets were investigated by two methods : direct measurement and simulation by MCNP code. From the direct measurement, it showed that about 0.5% ~ 1% of coincident neutron was due to the multiplicity by induced fission, and that the results of the measurement method also showed good agreement with those of simulation by the MCNP code. The neutron multiplicity correction curve obtained from this study will be used to increase measurement accuracy of the well-typed neutron accounting system in the future.

Development of high efficiency, multi-element CdZnTe detectors for portable measurement applications

We describe the development of detector arrays and electronics for large-volume, hand-held CdZnTe detectors with the same counting efficiency as portable NaI(Tl) detectors presently used for nuclear material measurement applications. The pulse-height resolution of the multi-element detectors is at least three times better than NaI(Tl) over a wide energy range (from 100 keV to several MeV), enabling more accurate measurements of gamma-rays emitted by special nuclear material. Arrays of up to eight coplanar grid detectors can be combined to make detectors ranging in size from 4 to 14 cm3. Because the number of spectroscopy channels is small, low-power, hand-held detectors can be manufactured with conventional printed circuit board technology, thus keeping the cost of multi-element detectors to a minimum. The design and performance of an 8-element detector is presented.

A High-Performance Neutron Time Correlation Counter

A new, high-performance neutron time correlation, or coincidence, counter has been developed for the measurement of nuclear materials. This detector is an improvement over the high-level neutron coincidence counter that was used previously in that it has better counting rate capability, electronic stability, noise resistance, uniform counting zone, and portability. The system uses /sup 3/He gas tubes coupled to the shift register coincidence electronics through a new hybrid charge-sensitive preamplifier/discriminator. Counting rates of 1.3 MHz have been obtained and a counting stability of 0.002% has been measured. The system has been applied to the assay of plutonium content in metal, PuO/sub 2/, and PuF/sub 4/, as well as in uraniumplutonium mixtures.