Engineering topics

Chung, Heejun

Publications and source records attributed to Chung, Heejun.

7 records

DOE OSTI2020

Sampling and Analysis Plan for Investigation of Undeclared Nuclear Activities - 20104

The nuclear-centered energy policy in South Korea is moving towards a nuclear-free era. It also facilitates an action or progress on the denuclearization of the Korean peninsula. The Nuclear Safety and Security Commission (NSSC) launched a new R and D project which attempts to verify nuclear activities via analyzing environmental samples. Environmental samples possibly contain significant radiological information relevant to past undeclared activities. In order to precisely verify them, it is essential to provide 1) the concrete information of target facilities, 2) appropriate methodologies and procedures, and 3) well-suited instrumentation and equipment corresponding to sample signatures. It is also important to provide scientific reliability or validity in any process such as sample collection, handling, transportation and laboratory analysis. Therefore, the information of nuclear facilities (fuel fabrication plant, graphite moderated reactor, reprocessing plant) in North Korea are reviewed in order to determine parameters for appropriate sampling methods representing a facility's characteristics. Secondly, a sampling procedure for radioactive elements is being developed based on international and domestic standards such as ASTM (American Society for Testing and Materials), KS (Korean industrial Standards), and International Atomic Energy Agency (IAEA) manuals. The procedure contains some specific information including sampling process, sampling tools, amount of sample, sample identification, and other related information. Developing the qualified national guidelines is most important since sampling and analyzing results can be used as evidences of nuclear activities. The guideline developed through this work will clearly define various methodologies such as document verification, sampling procedures, collection methods, sample storage criteria, chain of custody, and transportation conditions. The final objective of this work is to establish the infrastructure and regulatory framework for assessment of nuclear activities, and provide a national guideline for sampling and analysis. This will bring confidence in the results and play a critical role in complete denuclearization progress in South Korea. (authors)

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2020

An Experimental Feasibility Study on Applying Neutron Tomography to Encapsulated Spent Nuclear Fuel - 20050

Visual inspection makes easier to ensure the integrity and safety of spent nuclear fuel (SNF) than any classical techniques. Various classical techniques have been applied but there are no reliable methods to qualitatively and quantitatively verify spent fuel in dry storage. Thus, the present authors have developed the prototype safeguards apparatus for dry storage employing the array of He-4 gas scintillation detectors (S670E, Arktis Radiation Detectors Ltd., Switzerland), newly designed to simultaneously measure thermal and fast neutrons without any moderators. The S670E detector has a cylindrical shape with a diameter of 52 mm and active length of 600 mm (total length: 875 mm). The detector is filled by He-4 gas with an approximate pressure of 180 bar for fast neutron detection, and its inner wall is coated by Li-6 for thermal neutron detection. The scintillation lights generated via Li-6 nuclear reaction and elastic scattering are collected by 24 SiPMs linearly paired at the center of the detector. The detector delivers a TTL (Transistor-Transistor Logic) output for pulse readout and UART (Universal Asynchronous Receiver Transmitter) for device control. In order to assess feasibility of the apparatus, an experimental system has been designed, built, and optimized via computational studies. Cf-252 neutron sources and linearly arrayed detectors, working as a single detector, were occupied for this study due to the difficulties in working with SNF. The laboratory scale cask (diameter: 0.67 m, height: 1.5 m), minimized by a factor of 10 compared to the actual thickness of a commercial TN-32 cask, was also manufactured. The detector array was designed to rotate the lab-scale cask and obtain 36 image profiles at every 10 degrees. All profiles were aligned in single frame image called a sinogram, and the cross-sectional image was then fabricated by the inverse radon transform algorithm. These experiments have been repeated with different configurations and numbers of sources. Some gamma-ray sources were also measured with neutron sources in order to distinguish between neutron and gamma-ray pulses. Basically, a He-4 detector is designed to run on Linux OS so it is difficult to directly apply to Windows-based equipment widely used in S. Korea. Therefore, a new data acquisition board working on Windows OS was designed and built. The board mainly consists of FPGA (Field Programmable Gate Array) and SoC (System on Chip) for TTL pulse readout, sorting measured data, and transferring data to a user interface. In conclusion, the tomographic system at lab-scale has shown considerable potential to detect a partial or gross defect of encapsulated assemblies in dry storage. Next steps of this study will be to 1) repeatedly carry out experiments to demonstrate scientific reliability and validity, and 2) numerically integrate signals with weight factors to enhance the image quality since the suggested system based on passive interrogation method requires longer measurement time. Finally, the system will apply to a commercial dry storage phased out soon in S. Korea. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2018

A New Safeguards Technique for Spent Fuel Verification in Dry Storage by Using He-4 Detectors - 18262

It is demanded to find key issues and critical paths for verifying and monitoring a spent nuclear fuel cask or storage with a purpose of non-proliferation of nuclear weapon. However, unfortunately there are no reliable methods to verify the containment in a spent nuclear fuel cask or storage. Therefore, research and development (R and D) has been continuously required to develop new and alternative technologies for searching and verifying undeclared nuclear materials, activities, and facilities. Therefore, the present authors have developed a new safeguards technique of passively scanning a spent fuel cask or storage using pressurized He-4 scintillation neutron detectors. A new approach is to quickly verify content in a spent nuclear fuel cask or presence of all special nuclear materials in unknown packages by comparing experimental and computational results. For that, the present authors have developed a new algorithm and user-interface coupled ORIGEN-ARP and MCNP codes, and the prototype safeguards apparatus employing the array of He-4 fast neutron detectors has been designed, built, and tested. The feasibility test of this new apparatus has been performed with Cf-252 neutron sources and the laboratory scale model cask, minimized by a factor of ten compared to the actual dimension of a KN-21 cask. This technique has several advantages over current methods based on neutron and gamma detection; (1) incident neutron energy information is preserved, (2) disturbance of gamma-ray is excellently rejected (discriminated), and (3) much higher neutron counts and operational cost and space-effectiveness are expected since no moderator is required for fast neutrons. The final goal of this study will (1) conduct a prototype instrument and employ it to monitor a nuclear spent fuel cask or storage at a commercial nuclear power reactor facilities, (2) use spectral-specific measurement data to show proof of concept for measurable data successfully converted to quantifiable signatures, such as diversion/removal of fuel, and maintenance of fuel and cask integrity of fuel, and (3) develop a cask imaging capability. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Computational Safeguards Analysis of PWR Spent Nuclear Fuel in Dry Storage - 17435

The master plan for spent nuclear fuel (SNF) management in Korea recently adopted by Korean atomic energy promotion council. According to the master plan, the detailed practical and R and D strategies for SNF transferring, storage and disposal should be determined in the near future. Besides, it is also necessary to make plan how to secure SNF at each step in terms of nuclear safeguards, safety, security and nonproliferation. Several nondestructive analysis (NDA) techniques have been studied for assuring the safety of SNF in PWR methods. Obviously, gamma-ray is most dominant radiation emitted by SNF so several gamma-ray measurement methods have been developed and applied for the purpose of SNF monitoring. Those methods, however, suffer from a lack of nuclear data and interfering background radiation. To date, several neutron measurement methods have been developed as a solution to the problem of gamma-ray methods. Korea Institute of Nuclear Nonproliferation and Control (KINAC) currently develops a new novel technique of fast neutron detection for spent fuel monitoring at dry spent fuel storage facilities. As the first step, a computational program has been developed in order to evaluate neutron flux on the surface of PWR dry storage cask and study for some potential sabotage events. Typically, neutron flux on the surface of dry storage cask can be evaluated via two steps. The source term, based on burn-up rate of SNF, its cooling time and others related to operation condition at nuclear power plants, can be firstly defined by ORIGEN code. The surface flux can be then evaluated via MCNP code. These two separated steps are complicated and time consuming so KINAC developed a new algorithm and user-interface, coupled ORIGEN and MCNP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Evaluation of Neutron Flux on the Surface of a PWR Dry Storage Cask - 17574

Since the construction of a dry storage facility for spent nuclear fuel is expected in Korea, several methods have been studied for assuring the safety of pressurized-water reactor (PWR) spent fuels in dry storage casks. Existing evaluation methods are mostly based on detection technology of gamma-rays. However, the gamma-ray detection methods have high uncertainties due to background radiation. Therefore, we developed an evaluation method for the spent nuclear fuel monitoring based on neutron detection. In this study, we evaluated neutron flux on the surface of a PWR dry storage cask using computational simulation. Neutron release rate and energy spectra from a Westinghouse (WH) 17x17 type spent fuel with 10 years cooling time were calculated using ORIGEN-ARP code. MCNP code was used to simulate geometry and material properties of the spent fuel assemblies and casks. Neutron flux after 10 years of cooling time of the spent fuels was calculated on the surface of top, side, and bottom of the casks. The metal cask and the concrete cask, called as KORAD-21 casks which newly developed by Korea Radioactive Waste Agency (KORAD), were selected for this study. As a result, for the metal cask, average neutron flux were 4.79x10{sup 2} neutrons/cm{sup 2} sec, 5.29x10{sup 2} neutrons/cm{sup 2} sec, and 1.47x10{sup 3} neutrons/cm{sup 2} sec on the surface of top, side, and bottom of the cask, respectively. For the concrete cask, average neutron flux on the surface of top, side and bottom of the casks were 5.07x10{sup 1} neutrons/cm{sup 2} sec, 1.18x10{sup 1} neutrons/cm{sup 2} sec, and 1.46x10{sup 3} neutrons/cm{sup 2} sec, respectively. The study results can be applied to evaluate characteristics of PWR spent fuels and the neutron flux of dry storage cask. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Improved fission neutron energy discrimination with 4 He detectors through pulse filtering

Our paper presents experimental and computational techniques implemented for 4 He gas scintillation detectorsfor inducedfission neutron detection. Fission neutrons are produced when natural uranium samples are activelyinterrogated by 2.45 MeV deuterium-deuterium fusion reaction neutrons. Fission neutrons of energies greaterthan 2.45 MeV can be distinguished by their different scintillation pulse height spectra since 4 He detectorsretain incident fast neutron energy information. To enable the preferential detection of fast neutrons up to10 MeV and suppress low-energy event counts, the detector photomultiplier gain is lowered and triggerthreshold is increased. Pile-up and other unreliable events due to the interrogating neutronflux and backgroundradiation arefiltered out prior to the evaluation of pulse height spectra. With these problem-specific calibrationsand data processing, the 4 He detector's accuracy at discriminatingfission neutrons up to 10 MeV is improvedand verified with 252 Cf spontaneousfission neutrons. Given the 4 He detector's ability to differentiate fastneutron sources, this proof-of-concept active-interrogation measurement demonstrates the potential of specialnuclear materials detection using a 4 He fast neutron detection system.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Multi-element CZT array for nuclear safeguards applications

To address drawbacks of a CZT detector due to its limited size, a dual-CZT array, consisting of two CZT detectors and a DAS (Data Acquisition System), was constructed and algorithm to handle the spectra from the CZT array was discussed in this study. Spectra of 4.5% enriched UO{sub 2} powder were obtained using the dual-CZT array. The detection efficiency acquired by the CZT array was about twice higher than that of a single CZT spectra, while FWHMs are nearly identical (about 7.0 keV) for each spectrum. The results of this study show that a much better detection efficiency of CZT can be achieved without degrading its energy resolution. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗