Fissile Assay Research Articles

Sensitivity simulation on isotopic fissile measurement using neutron resonances

Uranium and plutonium are required to be accounted in spent fuel head-end and major recovery area in pyro-process for safeguards purpose. The possibility of neutron resonance technique, as a non-destructive analysis, was simulated on isotopic fissile analysis for large scale process. Neutron resonance technique has advantage to distinguish uranium from plutonium directly in mixture. Simulation was performed on U235 and Pu239 assay in spent fuel and for scoping examination of assembly type. The resonance energies were determined for U235 and Pu239. The linearity in the neutron transmission was examined for the selected resonance energies. In addition, the limit for detection was examined by changing sample density, thickness and content for actual application. Several factors were proposed for neutron production and the moderated neutron source was simulated for effective and efficient transmission measurement. From the simulation results, neutron resonance technique is promising to analyze U235 and Pu239 for spent fuel assembly. An accurate fissile assay will contribute to an increased safeguards for the pyro-processing system and international credibility on the reuse of fissile materials in the fuel cycle.

Nuclear measurement in Pyro-processed wastes

Nuclear fissile material was analyzed for the waste produced in the pyro-process. Nuclear waste is produced at each stage in the pyro-process, spent fuel head-end, major group recovery area and TRU-RE process. Each yields waste of different material composition and properties. Chemical analysis is basically determined for the measurement of nuclear material content. However, the possibility of non-destructive analysis techniques were examined as an additional and supplementary way for cost effectiveness and time saving in the nuclear material analysis. Examinations were undertaken on RE-salt waste, anode sludge, and hull. A gamma peak of Cm244 and Pu239 was simulated for the RE waste. The peak was well distinguishable in the waste. The linearity in the gamma measurement was approved with the content increase in the plutonium isotope. For fissile assay in anode sludge and hull, a neutron method was suggested. From the simulation results, a probable technique was recommended for plutonium content assay in each waste of the pyro-process. An accurate fissile assay will contribute to an increased safeguards for the pyro-processing system and international credibility for the reuse of fissile materials in the fuel cycle.

Nuclear isotopic fissile content assay

Isotopic fissile content assay was performed in a designed lead slowing down spectrometer (LSDS) system. In the system, the fission measurement was performed for 11 – 20 wt% of Pu239 with 1 wt% of Pu241 and U235. The self-shielding factor was applied to the measured fission signal. The fissile assay was performed on U235, Pu239, and Pu241. The content assay of an unknown fissile sample was also performed on U235, Pu239, and Pu241. The procedure of an unknown sample assay was setup. In several simulations, the content of Pu239 was accurately obtained with ∼1% error in the rod base. For the unknown fissile assay, the result was very close to the actual content of Pu239, ∼1% error. The LSDS provided very powerful result for the fissile assay in the mixture. Therefore, the LSDS is a very promising technique in isotopic fissile content assays, especially for Pu239 in uranium and plutonium. Precise fissile material content should be known for reactor safety in fuel cycles and international safeguards on nuclear fuel cycle development.

Thermal analysis at target for neutron generation in fissile assay

A target was designed to produce intense neutrons of 1012 neutrons/sec in order to analyze the fissile content by interrogation in the Lead Slowing Down Spectrometer (LSDS) system. A high energy and current electron linear accelerator was used in neutron production by bremsstrahlung radiation. However, the interaction by electron and radiation in the target generates large amounts of heat. Several coolants were simulated in the target and the heat removal capacity was performed. The thermal property of coolant in the target was calculated using the CFX code. The heat distribution by the cooling method was simulated and the forced convection method was used to calculate the temperature distribution on the different coolants and flow rate. Helium shows a good cooling property, but it needs high flow rate with high pressure. Air may be a good candidate coolant if the limit of neutron activation is satisfied in the circulating system. When the coolants were used in the target, each metal plate of the target was kept under the melting temperature. For the estimation of the coolant performance, the neutron influence by the coolant was examined as well. The material property was also evaluated for various target cases. The displacement per atom (DPA) analysis of the target was performed for a life time operation of the LSDS system. For the target container material (HT-9), the optimum thickness was obtained from the sensitivity simulation.

Quantitative Fissile Assay In Used Fuel Using LSDS System

A quantitative assay of isotopic fissile materials (U235, Pu239, Pu241) was done at Korea Atomic Energy Research Institute (KAERI), using lead slowing down spectrometer (LSDS). The optimum design of LSDS was performed based on economics, easy maintenance and assay effectiveness. LSDS system consists of spectrometer, neutron source, detection and control. LSDS system induces fissile fission and fast neutrons are collected at fission chamber. The detected signal has a direct relation to the mass of existing fissile isotopes. Many current commercial assay technologies have a limitation in direct application on isotopic fissile assay of spent fuel, except chemical analysis. In the designed system, the fissile assay model was setup and the correction factor for self-shield was obtained. The isotopic fissile content assay was performed by changing the content of Pu239. Based on the fuel rod, the isotopic content was consistent with ~2% uncertainty for Pu239. By applying the covering (neutron absorber), the effective shielding was obtained and the activation was calculated on the target. From the assay evaluation, LSDS technique is very powerful and direct to analyze the isotopic fissile content. LSDS is applicable for nuclear fuel cycle and spent fuel management for safety and economics. Additionally, an accurate fissile content will contribute to the international transparency and credibility on spent fuel.

DEVELOPMENT OF LEAD SLOWING DOWN SPECTROMETER FOR ISOTOPIC FISSILE ASSAY

A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ∼E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.

Design of Lead Spectrometer for Spent Fuel Fissile Assay

Design of Lead Spectrometer for Spent Fuel Fissile Assay

DESIGN OF LSDS FOR ISOTOPIC FISSILE ASSAY IN SPENT FUEL

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

Neutron tomographic fissile assay in spent fuel using the lead slowing down time spectrometer

A tomographical fissile assay technique for light water reactor spent fuels was developed based on the lead slowing down time spectrometer (LSDTS). The Monte Carlo method was used for system performance and for simulation of the sensitivity of detection of fissile materials using neutron emission tomography. A typical spent PWR fuel element was simulated and the spatial and mass sensitivity of the fissile components were investigated. From these studies, the LSDTS system is shown to be a very sensitive device for analyzing the spatial distribution of total fissile materials of a spent fuel assembly in a properly selected assay neutron energy range. This method is also applicable to nuclear waste assay.

The Expanding Role of the Small Van De Graaff in Nuclear Nondestructive Analysis

In support of Nuclear Safeguards, a 3-MeV Van de Graaff accelerator was installed at LASL to aid research in techniques for nondestructive assay of fissionable materials. Monoenergetic neutrons make possible fissile assay techniques less influenced by self-absorption than thermal interrogation and more discriminating of fissile over fertile material than use of partially moderated high-energy neutrons. Delayed fission-neutron counting, a valuable assay technique, is facilitated by pulsed neutron on-to-off ratios in excess of 109, achieved with special beam-handling techniques. Where high neutron backgrounds, e.g. from samples containing 240Pu, preclude delayed-neutron counting, fission prompt-neutron counting with an energy-biasable detector is applied. The increasing facility workload, which includes detector development and calibration, research in trace analysis by proton-induced x-ray fluorescence and research in light-element isotopic assay by means of charged-particle induced reactions, indicates that technological application may more than supplant dwindling nuclear research as a market for accelerators of this class.