• Nuclear Engineering and Technology
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• 제48권4호
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• 한국유체기계학회 논문집
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• 제20권1호
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• pp.65-73
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• 2017

Intercoolers for multi-stage turbocharger of the hydrogen reciprocating engine for HALE UAV are installed for reducing the charged air inlet temperature of the engine. The intercooler is air to air, cross flow, plate-fin type and the fin configuration is offset-strip fin which is referenced from the heat exchanger of the ERAST. Most of HALE UAV's cruising altitude is 60,000 ft and the density of air for this altitude is very low compared to sea level. Therefore the required heat transfer area for the HALE UAV is about three-times bigger than the sea level. Consequently, it is essential to design to meet the required efficiency of intercooler in the range of not excessively growing the weight of the heat exchanger. The quasi-one dimensional heat transfer design/analysis for satisfying the requirement of the engine are written in this paper. The numerical analyses for estimating the coolant flow rate of the engine bay and pressure loss in the header and core are also summarized.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.274-284
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• 1992

가압경수로 최적 열수력 분석용 전산코드인 CATHRE의 모델 평가를 위하여 가압경수로의 가상 냉각재 상실사고시 원자로 용기내의 유동현상을 모의한 1/15축소의 CREARE 실험을 모의 계산하였다. 이 실험에서 주요변수들은 비상노심 탱각재 주입량과 아냉정도 그리고 계통압력 및 노심에서 발생되는 증기유량이지만. 본 연구에서는 우선 Downcomer에서 역방향유동의 정성적 분석에 촛점을 맞추었다. 모의 계산 결과와 실험 결과를 비교할 때 정량적인 값 뿐 아니라 변화의 경향에서도 차이가 나타난 것은 주로 적절하지 못한 일부의 수치해석 모델과 상간의 계면마찰 때문으로 판단된다. 따라서 매개변수적 민감도 분석을 통하여 CATHARE 전산코드의‘VOLUME’에 접한 접합점에서 운동량 보존방정식의 상세연구 혹은 다차원 분석을 통해서 이 경우의 물리적 현상을 보다 현실적으로 나타낼 수 있다는 결론을 얻었다.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.700-711
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• 2015

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called "self-wastage phenomena." A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).

• 에너지공학
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• 제28권3호
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• pp.26-33
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• 2019

ATLAS와 LSTF 장치에서 수행된 저온관(CL) 파단 13% 및 17% IBLOCA 대응실험들을 비교하고 특히, 핵심 관심 인자인 노심 첨두피복온도(PCT)에 대하여 비교 검토하고 아울러 주요 열수력 현상에 대하여 토론하였다. 비교.검토에서 두 건의 CL 파단 IBLOCA 대응실험들은 PCT 거동에 있어서 꽤 큰 차이를 보이고 있는 것을 확인하였는데 이는 두 장치의 척도 차이로 인한 왜곡현상을 벗어나는 경향을 보이고 있다는 점에서 두 장치의 원자로냉각재계통에 대한 자세한 설계 비교를 수행하였다. 이에 두 장치 사이에 핵연료조정판(FAP) 설계에 있어서 차이가 있다는 점을 확인하였다. 이에 따라 IBLOCA 사고시 Reflux 응축수의 노심 유입에 중요한 역할을 하는 CCFL 관련 무차원직경 값에서도 두 장치에서 매우 다른 차이를 보이고 있다는 점에서 CL 파단 IBLOCA 대응실험에서의 PCT 거동의 현격한 차이를 설명할 수 있는 원인일 수 있는 인자라는 것을 발견하였다. 향후 관련 설계 차잇점을 근거로 더 자세한 검토와 분석을 통해 관련 현상을 이해할 수 있을 것으로 판단된다.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2008년 영문 학술대회
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• pp.109-114
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• 2008

This paper presents the development status of a subscale precooled turbojet engine "S-engine" for the hypersonic cruiser and space place. S-engine employs the precooled-cycle using liquid hydrogen as fuel and coolant. It has $23cm{\times}23cm$ of rectangular cross section, 2.6 m of the overall length and about 100 kg of the target weight employing composite materials for a variable-geometry rectangular air-intake and nozzle. The design thrust and specific impulse at sea-level-static(SLS) are 1.2 kN and 2,000 sec respectively. After the system design and component tests, a prototype engine made of metal was manufactured and provided for the system firing test using gaseous hydrogen in March 2007. The core engine performance could be verified in this test. The second firing test using liquid hydrogen was conducted in October 2007. The engine, fuel supplying system and control system for the next flight test were used in this test. We verified the engine start-up sequence, compressor-turbine matching and performance of system and components. A flight test of S-engine is to be conducted by the Balloon-based Operation Vehicle(BOV) at Taiki town in Hokkaido in October 2008. The vehicle is about 5 m in length, 0.55 m in diameter and 500 kg in weight. The vehicle is dropped from an altitude of 40 km by a high-altitude observation balloon. After 40 second free-fall, the vehicle pulls up and S-engine operates for 60 seconds up to Mach 2. High altitude tests of the engine components corresponding to the BOV flight condition are also conducted.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1825-1834
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• 2022

Performing high-fidelity computational fluid dynamics (HF-CFD) to predict the flow and heat transfer state of the coolant in the reactor core is expensive, especially in scenarios that require extensive parameter search, such as uncertainty analysis and design optimization. This work investigated the performance of utilizing a multi-fidelity reduced-order model (MF-ROM) in PWR rod bundles simulation. Firstly, basis vectors and basis vector coefficients of high-fidelity and low-fidelity CFD results are extracted separately by the proper orthogonal decomposition (POD) approach. Secondly, a surrogate model is trained to map the relationship between the extracted coefficients from different fidelity results. In the prediction stage, the coefficients of the low-fidelity data under the new operating conditions are extracted by using the obtained POD basis vectors. Then, the trained surrogate model uses the low-fidelity coefficients to regress the high-fidelity coefficients. The predicted high-fidelity data is reconstructed from the product of extracted basis vectors and the regression coefficients. The effectiveness of the MF-ROM is evaluated on a flow and heat transfer problem in PWR fuel rod bundles. Two data-driven algorithms, the Kriging and artificial neural network (ANN), are trained as surrogate models for the MF-ROM to reconstruct the complex flow and heat transfer field downstream of the mixing vanes. The results show good agreements between the data reconstructed with the trained MF-ROM and the high-fidelity CFD simulation result, while the former only requires to taken the computational burden of low-fidelity simulation. The results also show that the performance of the ANN model is slightly better than the Kriging model when using a high number of POD basis vectors for regression. Moreover, the result presented in this paper demonstrates the suitability of the proposed MF-ROM for high-fidelity fixed value initialization to accelerate complex simulation.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.701-709
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• 1995

RELAP5 /MOD3/KAERl의 임계유동모델을 위한 실제적인 배출계수들을 9개의 MARVIKEN 임계유동실험 의 평가계산을 통하여 과냉각과 이상임계유동에 대하여 구하였다. 선택된 실험에는 높은 초기 과냉각도와 큰 노즐 세 장비(L/D)인 것들이 포함되었다. 코드의 평가결과는 RELAP5/MOD3/KAERI은 과냉각임계유동을 크게 예측하고 이 상임계유동은 작게 예측함을 보이고 있다. 이러한 결과들을 이용하여 임계유동모델의 실제적인 배출계수들을 반복법으로 정량화 하였다. 실제적인 배출계 수는 과냉각임계유동이 0.89 그리고 이상임계유동이 1.07로 결정되었으며 관련 표준편차는 각 각 0.0349과 0.1189이다. 본 연구로부터 얻어진 결과는 대형냉각재 상실사고의 실제적인 계통반응 계산과 비상노심냉각계통 성능평가에 적용할 수 있다.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2534-2546
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• 2023

Gas-cooled space reactor, which adopts He-Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He-Xe gas mixture in 2×2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%-50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He-Xe gas cooled space nuclear reactor.