• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.709-716
• /
• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• Nuclear Engineering and Technology
• /
• 제54권12호
• /
• pp.4731-4742
• /
• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2003년도 가을 학술논문집
• /
• pp.609-613
• /
• 2003

사용후핵연료 및 방사성물질을 저장하기 위한 저장시스템은 사용후핵연료를 저장하는 동안 안전성 문제를 야기하지 않도록 격납을 설계하고 평가하여야 하며, 격납 평가는 ANSI Nl4.5 또는 ISO 12807에서 규정하고 있는 절차에 따른 허용 누설률을 계산하여 평가할 수 있다. 따라서, ISO 12807에서 규정한 평가방법에 따라 PWR 사용후핵연료 24 다발을 저장하였을 경우와 금속전환체 24다발을 저장하였을 경우에 대한 허용 누설률을 평가하였다. OWR 사용후핵연료 24다발을 저장하였을 경우 허용 누설률은 $1.38{\times}10_{-10}m_3/s$로, 금속전환체 24다발을 저장하였을 경우 $4.46{\times}10_{-10}m_3/s$로 평가되었다. 따라서, 사용후핵연료를 저장하였을 경우보다 금속전환체를 저장하였을 경우 격납 조건이 수월해 짐을 알 수 있었다.

• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.788-793
• /
• 2018

High-density spent fuel (SF) storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others.

• Nuclear Engineering and Technology
• /
• 제17권1호
• /
• pp.1-7
• /
• 1985

원자력 발전소의 조사후 핵연료 저장풀에시 조사후 핵연료 집합체에 대한 감마선 분광분석 실험을 비파괴적인 방법으로 수행하였다. 조사후 핵연료 집합체가 갖는 연소분포를 알기 위해서 1차핵분열 생성물과 2차핵분열 생성물간의 감마선 강도비인 $^{134}$ Cs$^{137}$Cs을 사용했으며 그 결과는 이들 집합체가 노심내에서 연소시에 가졌든 중성자 분포의 기대치와 상응하였다. 이로부터 감마선 강도비 $^{134}$ Cs$^{137}$Cs은 연소도 해석을 위한 좋은 인디케이터임을 확인하였다. 핵물질의 안전관리면에서 중요시되고 있는 조사후 핵연료의 냉각시간을 감마선 강도비 $^{144}$ Ce$^{137}$Cs을 사용하여 구했으며 이를 핵연료 관리기록에 의한 냉각시간과 비교해 본 결과 각각 2%, 10%이내의 차이를 나타내었다. 이로부터 본 실험에서 냉각시간을 하기 위해서 유도한 방정식을 단일 주기로 연소된 핵연료에 대해서 사용할 수 있음을 실증하였다.

• Nuclear Engineering and Technology
• /
• 제50권3호
• /
• pp.504-511
• /
• 2018

In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for "Channelized Large Power Reactor"-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(2)
• /
• pp.674-679
• /
• 1995

The hold-down spring forces of Kori nuclear fuels were measured for seven fuel assemblies having 1 to 4 cycles of irradiation histories in the Kori Unit-1 and -2 reactor. The fuel assemblies examined had burnup from 17 to 38 GWD/MTU and the examination was conducted in KAERI PIEF spent fuel storage pool with the newly developed underwater hold-down suing force measuring device. The measurement was made within the elastic deformation ranges and the trends of hold-down spring force relaxation behavour were examined.

• 제어로봇시스템학회:학술대회논문집
• /
• 제어로봇시스템학회 1987년도 한국자동제어학술회의논문집; 한국과학기술대학, 충남; 16-17 Oct. 1987
• /
• pp.613-618
• /
• 1987

Rod consolidation is a method of increasing spent nuclear fuel storage capacity by disassembling fuel assemblies thus storing the fuel rods in a tighter array. It involves some basic operations which closely resemble to the material handling of a manufacturing process. But all the operations must be controlled remotely in shielded environment from outside due to the highly radioactive nature of the workpiece. In this study the status of the rod consolidation technology in other countries has been surveyed and a feasibility study for the conceptual design of this process have been made.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(2)
• /
• pp.847-852
• /
• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• pp.601-607
• /
• 2022

In recent years, monitoring of spent nuclear fuel inside dry cask storage has become an important area of national security. Muon tomography is a useful method for monitoring spent nuclear fuel because it uses high energy muons that penetrate deep into the target material and provides a 3-D structure of the inner materials. We designed a muon tomography system consisting of four 2-D position sensitive detector and characterized and optimized the system parameters. Each detector, measuring 200 × 200 cm², consists of a plastic scintillator, wavelength shifting (WLS) fibers and, SiPMs. The reconstructed image is obtained by extracting the intersection of the incoming and outgoing muon tracks using a Point-of-Closest-Approach (PoCA) algorithm. The Geant4 simulation was used to evaluate the performance of the muon tomography system and to optimize the design parameters including the pixel size of the muon detector, the field of view (FOV), and the distance between detectors. Based on the optimized design parameters, the spent fuel assemblies were modeled and the line profile was analyzed to conduct a feasibility study. Line profile analysis confirmed that muon tomography system can monitor nuclear spent fuel in dry storage container.