[en] The PRIDE (PyRoprocessing Integrated inactive DEmonstration) is an engineering-scale pyroprocessing test-bed facility that utilizes depleted uranium (DU) instead of spent fuel as a process material. As part of the ongoing effort to enhance pyroprocessing safeguardability, UNDA (Unified Non-Destructive Assay), a system integrating three different non-destructive assay techniques, namely, neutron, gamma-ray, and mass measurement, for nuclear material accountancy (NMA) was developed. In the present study, UNDA's NMA capability was evaluated by measurement of the weight, "2"3"8U mass, and U enrichment of oxide-reduction-process feed material (i.e., porous pellets). In the "2"3"8U mass determination, the total neutron counts for porous pellets of six different weights were measured. The U enrichment of the porous pellets, meanwhile, was determined according to the gamma spectrums acquired using UNDA's NaI-based enrichment measurement system. The results demonstrated that the UNDA system, after appropriate corrections, could be used in PRIDE NMA applications with reasonable uncertainty. It is expected that in the near future, the UNDA system will be tested with next-step materials such as the products of the oxide-reduction and electro-refining processes. - Highlights: • PRIDE UNDA has been developed and characterized for nuclear material accountancy. • The performance was evaluated with pyroprocessing feed material: UO_2 porous pellets made of depleted uranium. • Total neutron counting, U enrichment, and mass measurements were performed for porous pellet samples of various weights. • "2"3"8U mass determination by neutron measurement showed the relative difference of 0.79–13.28% compared with actual mass. • The enrichment measured by gamma-ray spectroscopy was significantly underestimated by 42.07%.

[en] The Advanced spent fuel Conditioning Process Facility (ACPF) at KAERI has been refurbished for the test of an electrolytic oxide reduction process using spent fuels. Also, KAERI has manufactured processrelated instruments as well as safeguards-related one. ACP Safeguards Neutron Counter (ASNC) has been developed to be tested for nuclear material accountancy (NMA) of the facility based on a coincidence neutron counting. KAERI has refurbished the ACPF hot-cell facility for the test of oxide reduction process using spent fuels, and manufactured process-related instruments as well as safeguards-related one. In this study, a neutron coinci-dence counter was fabricated for the test of a safeguards instrument in a hot-cell environment; hence, remote operation and maintenance capability were the impor-tant factors considered in design phase. In the near future, the modified ASNC will be installed in the hot-cell and tested with a standard Cf source in terms of various detector parameters and remote control capability.

[en] One of the options for spent-fuel management in Korea is pyroprocessing whose main process flow is the head-end process followed by oxide reduction, electrorefining, and electrowining. In the present study, a well-type passive neutron coincidence counter, namely, the ACP (Advanced spent fuel Conditioning Process) safeguards neutron counter (ASNC), was redesigned for safeguards of a hot-cell facility related to the oxide reduction process. To this end, first, the isotopic composition, gamma/neutron emission yield and energy spectrum of the feed material (i.e., the UO2 porous pellet) were calculated using the OrigenARP code. Then, the proper thickness of the gammaray shield was determined, both by irradiation testing at a standard dosimetry laboratory and by MCNP6 simulations using the parameters obtained from the OrigenARP calculation. Finally, the neutron coincidence counter’s calibration curve for 100- to 1000-g porous pellets, in consideration of the process batch size, was determined through simulations. Based on these simulation results, the neutron counter currently is under construction. In the near future, it will be installed in a hot cell and tested with spent fuel materials.

[en] Due to the high gamma sensitivity of organic scintillators, it is essential to discriminate signals induced by neutron from gamma-ray in fast-neutron detection. With the improvement of digital signal processing techniques, diverse discrimination methods based on pulse-shape variation by radiation type have been developed. The main purpose of this study was to verify the applicability of a deep-learning model, especially convolution neural network (CNN), to pulse-shape discrimination (PSD) in organic scintillation detectors, such as BC-501A (liquid) and EJ-276 (plastic). To that end, waveforms of neutron and gamma-ray were experimentally collected using point sources of ¹³⁷Cs (gamma-ray) and ²⁵²Cf (neutron/gamma-ray) and pre-processed for being compatible with deep-learning. The PSD performance was evaluated for both detectors using the charge comparison method (CCM) which is one of the representative conventional PSD techniques of time-domain. In addition, the CNN-based discriminating algorithms were tested, and its preliminary results were confirmed with confusion matrices which indicate the discrimination accuracy of a deep-learning mode

[en] Self-induced X-ray fluorescence (XRF) is a technique by which plutonium (Pu) content in spent nuclear fuel can be directly quantified. In the present work, this method successfully measured the plutonium/uranium (Pu/U) peak ratio of a pressurized water reactor (PWR)’s spent nuclear fuel at the Korea atomic energy research institute (KAERI)’s post irradiation examination facility (PIEF). In order to reduce the Compton background in the low-energy X-ray region, the Compton suppression system additionally was implemented. By use of this system, the spectrum’s background level was reduced by a factor of approximately 2. This work shows that Compton-suppressed self induced XRF can be effectively applied to Pu accounting in spent nuclear fuel.

[en] One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different ²³⁵U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.

[en] Plutonium (Pu) contents in spent nuclear fuels, recovered uranium (U) or uranium/transuranium (U/TRU) products must be measured in order to secure the safeguardability of a pyroprocessing facility. Self-induced X-Ray fluorescence (XRF) and gamma-ray spectroscopy are useful techniques for determining Pu-to-U ratios and Pu isotope ratios of spent fuel. Photon measurements of spent nuclear fuel by using high-resolution spectrometers such as high-purity germanium (HPGe) detectors show a large continuum background in the low-energy region, which is due in large part to Compton scattering of energetic gamma rays. This paper proposes a Compton suppression system for reducing of the Compton continuum background. In the present study, the system was configured by using an HPGe main detector and a BGO (bismuth germanate: Bi4Ge3O12) guard detector. The system performances for gamma-ray measurement and XRF were evaluated by means of Monte Carlo simulations and measurements of the radiation source. The Monte Carlo N-Particle eXtended (MCNPX) simulations were performed using the same geometry as for the experiments, and considered, for exact results, the production of secondary electrons and photons. As a performance test of the Compton suppression system, the peak-to-Compton ratio, which is a figure of merit to evaluate the gamma-ray detection, was enhanced by a factor of three or more when the Compton suppression system was used.

[en] In order to enhance the safeguardability of a pyroprocessing facility, the Korea Atomic Energy Research Institute (KAERI) has been endeavoring to develop more efficient and effective safeguards technologies for nuclear material accountancy (NMA), process monitoring, and containment and surveillance (C/S). NMA has two components: destructive analysis (DA) and non-destructive assay (NDA). Although DA is more accurate, it is typically time-consuming and cost-intensive. NDA, on the other hand, can provide reasonable accuracy on a real-time or near-real-time basis, which maximizes the utilization efficiency of a facility. In this study, the PRIDE (PyRoprocessing Integrated inactive DEmonstration) UNDA (unified non-destructive assay) was developed for testing NDA techniques at PRIDE, a demonstration facility within KAERI for integrated pyroprocessing using depleted uranium and surrogate materials. Each component of the PRIDE UNDA (i.e., neutron, gamma-ray, and mass measurement systems) was characterized and calibrated using calibration sources and standard weights as well as nuclear material used in the facility (depleted uranium). It is expected that in the near future, the PRIDE UNDA will be installed and tested with various types of process materials. (author)

[en] Highlights: • The ACP safeguards neutron counter (ASNC) for nuclear material accountancy of the ACPF was upgraded. • The remote-handling and maintenance capabilities for hot-cell operation were improved. • Various parameters were measured and determined for detector characterization. A safeguards neutron coincidence counter for nuclear material accountancy of the Advanced spent-fuel Conditioning Process Facility (ACPF), known as the ACP Safeguards Neutron Counter (ASNC), was upgraded to improve its remote-handling and maintenance capabilities. Based on the results of the previous design study, the neutron counter was completely rebuilt, and various detector parameters for neutron coincidence counting (i.e., high-voltage plateau, efficiency profile, dead time, die-away time, gate length, doubles gate fraction, and stability) were experimentally determined. The measurement data showed good agreement with the MCNP simulation results. To the best of the authors’ knowledge, the ASNC is the only safeguards neutron coincidence counter in the world that is installed and operated in a hot-cell. The final goals to be achieved were (1) to evaluate the uncertainty level of the ASNC in nuclear material accountancy of the process materials of the oxide-reduction process for spent fuels and (2) to evaluate the applicability of the neutron coincidence counting technique within a strong radiation field (e.g., in a hot-cell environment).

[en] A SiPM is an array of several thousand micro-Geiger-mode APDs (GAPD), and a GAPD generally consists of a PN diode (C_d), a quenching resistor (R_q) and any parasitic capacitor (C_q) in parallel to the quenching resistor. The single-photon pulse shape is determined by a RC decay time, a product of C_total (C_d+C_q) and R_q. A large fraction of current in the long tailed single-photon pulse shape causes Scintillation detectors with SiPM to exhibit a poor timing resolution compared to PMTs, which have very narrow Gaussian single-photon pulse shape. In this paper, a ∼40fF Metal-Insulator-Metal (MIM) quenching capacitor (C_q-MIM) parallel to a quenching resistor in a Geiger mode APD is proposed to modify the single-photon pulse shape of a SiPM, in order to improve the timing performance of PET detectors. A single-photon pulse shape with C_q-MIM has a very fast pulse with a low slow tail, consequently the increasing initial current.