• Nuclear Engineering and Technology
• /
• 제50권7호
• /
• pp.1060-1067
• /
• 2018

A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial $k_{eff}$ level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially $^{241}Am$ is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

• 한국압력기기공학회 논문집
• /
• 제16권2호
• /
• pp.18-24
• /
• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• 한국표면공학회지
• /
• 제56권1호
• /
• pp.84-93
• /
• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

• 방사성폐기물학회지
• /
• 제18권spc호
• /
• pp.37-50
• /
• 2020

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy·hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

• 대한전자공학회논문지
• /
• 제22권5호
• /
• pp.75-82
• /
• 1985

Weighted Least Square(W.L.S.)알고리즘을 통한 계수식별 이론과 출력 증차를 최소로 하기위해 한단계 앞서서 제어입력을 결정하는 제어이론을 결수한 기준 모델 통용제어를 응용한 연료펌프 시스템을 설계하였다. 시스템의 성능에 관계되는 샘플링 이기 7, 비중계수 공및 State Yarlable Filter(5. Y. F.)의 릴터계수의 값은 컴퓨터 시뮬레이션을 통해 결정했으며, 특히 주독온도의 변화로 인해 연료의 점성도가 변하므로서 공정의 동특성을 나타내는 공량계수가 변하는 것에 대한 제어시스템의 적응도를 레팔레이션(regu lation)과 추종(tracking)관점에서 고찰하였다.

• 대한조선학회논문집
• /
• 제52권2호
• /
• pp.110-118
• /
• 2015

Weather routing method is one of the best practices of SEEMP (Ship Energy Efficiency Management Plan) for fuel-efficient operation of ship. KR is carrying out a basic research for development of the weather routing algorithm and making a monitoring system by FOC (Fuel Oil Consumption) analysis compared to the reference, which is the great circle route. The added resistances applied global sea/weather data can be calculated using ship data, and the results can be corrected to ship motions. The global sea/weather data such as significant wave height, ocean current and wind data can be used to calculate the added resistances. The reference route in a usual navigation is the great circle route, which is the shortest distance route. The global sea/weather data can be divided into grids, and the nearest grid data from a ship's position can be used to apply a ocean going vessel's sea conditions. Powell method is used as an optimized routing technique to minimize FOC considered sea/weather conditions, and FOC result can be compared with the great circle route result.

• Nuclear Engineering and Technology
• /
• 제53권8호
• /
• pp.2454-2463
• /
• 2021

The objective of this research is to the use of americium (AmO₂) as a burnable absorber effectively instead of conventional gadolinium (Gd₂O₃) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO₂ as a burnable absorber. From these studies, it is found that AmO₂ can decrease the excess reactivity like Gd₂O₃ without changing the criticality life span and enrichment of ²³⁵U. A homogeneous mixture of the 0.20% AmO₂+ 4.95% enriched UO₂ fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO₂+UO₂ in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

• 한국연소학회지
• /
• 제18권4호
• /
• pp.21-28
• /
• 2013

HCCI engines have mainly focused on achieving low temperature combustion in order to obtain higher efficiency and lower emission. One of practical difficulties in HCCI combustion is to control the start of combustion and subsequent combustion phasing. The choice of primary reference fuels in HCCI strategy is one of various promising solutions to make HCCI combustion ignition-controlled. The behavior of ignition delay to the frequency variation of sinusoidal velocity oscillation is computationally investigated under HCCI conditions of PRF75 using a reduced chemistry in a counterflow configuration. The second-stage ignition is more delayed as the higher frequency is imposed on nozzle velocity fluctuation whereas the first-stage ignition is not much influenced.

• Nuclear Engineering and Technology
• /
• 제53권5호
• /
• pp.1723-1729
• /
• 2021

Energy and directional distribution of neutrons at the Beloyarsk NPP workplaces is a subject of this study. Measurements of H^*(10) rate and neutron energy distribution were taken at 8 workplaces, which can be divided into three categories: work with spent or fresh nuclear fuel, work with radionuclide neutron sources, work at the rooms adjusted to reactors. The Hp(10) measurements were performed only at 6 out of 8 locations, due to the fact that long term placing of an effective neutron moderator in fresh nuclear fuel storage facility is forbidden. As a result of the research energy and direction distribution of the neutron fields at 8 locations of the Beloyarsk NPP workplaces was obtained. To estimate the accuracy of the H^*(10) rate and Hp (10) measurements the reference values of dose equivalents were calculated using energy and directional distribution. To take into account the difference between the reference values and the measured results site-specific correction factors were calculated.

• Nuclear Engineering and Technology
• /
• 제18권1호
• /
• pp.17-26
• /
• 1986

연소를 고려하는 사용후핵연료저장조의 핵임계 안전성 분석에서 검증용 계산 체제와 rack계산 체제 사이의 편차를 신뢰성 있게 결정하는 방법을 시험하였다. 이를 위하여 고리 1호기의 사용후핵연료저장조를 연소를 고려하는 가장 조밀한 rack으로 개념설계하고, 핵연료의 농축도 및 연소도에 따라 증배계수를 계산하였다. 표준값 생산용 Monte Carlo 코드로는 KENO-IV를 그리고 실제 rack 설계용으로는 2차원 충돌화률 코드인 FATAC을 사용하였다. 이 두 계산의 결과를 상호 비교하여 계산 체제 사이의 편차와 이의 경향성 및 신뢰도를 평가하였다. 이 방법을 사용하면 확실한 신뢰도 근거를 마련할 수 있을 뿐만 아니라 반응도 여유면에서 기존의 방법보다 불리하지 않음이 입증되었다.