• Nuclear Engineering and Technology
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• 제55권1호
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• pp.330-338
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• 2023

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.

• 한국자동차공학회논문집
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• 제16권1호
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• pp.37-44
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• 2008

The global warming potential (GWP) of $CO_2$ refrigerant is 1/1300 times lower than that of R134a. Furthermore, the size and weight of the automotive heat pump system can decrease because $CO_2$ operates at high pressure with significantly higher discharge temperature and larger temperature change. The presented $CO_2$ heat pump system was designed for both cooling and heating in fuel cell vehicles. In this study, the performance characteristics of the heat pump system were analyzed for heating, and results for performance were provided for operating conditions when using recovered heat from the stack coolant. The performance of the heat pump system with heater core was compared with that of the conventional heating system with heater core and that of the heat pump system without heater core, and thus the heat pump system with heater core showed the best performance among the selected heating systems. On the other hand, the heating performance of two different types of coolant/air heat pump systems with heater core was compared each other at various coolant inlet temperatures. Furthermore, to use exhausted thermal energy through the radiator, experiments were carried out by changing the arrangement of a radiator and an outdoor evaporator, and quantified the heating effectiveness.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.656-661
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• 2010

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER$^2$) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• Nuclear Engineering and Technology
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• 제32권1호
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1999년도 하계학술대회 논문집 A
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• pp.372-375
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• 1999

This paper presents an arc(hot-gas flow field) analysis method in GCB. This method includes the Lorentz's force due to magnetic field, turbulent viscous effect and radiation heat transfer which are indispensable to the analysis of hot-gas flow. To verify the applicability of the Proposed method, steady state hot-Eas flow analysis within a simplified interrupter has been carried out. Inlet boundary pressure values were assumed to be 9.0atm and 12.0atm. For each inlet boundary condition, three cases of hot-gas flow field analyses were performed according to the values of arc currents which were assumed to be D.C 0.6kA. 1.0kA and 2.0kA. The results revealed that the arc radius at nozzle throat has been concentrated by increasing the pressure of nozzle upstream and that the maximum temperature of arc core has been decreased along to nozzle exit and the high temperature lesion come to be wide in nozzle downstream. From these results, it is confirmed that the proposed method will be applicable to predict the large current interruption capability of GCB.

• 한국기계가공학회지
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• 제18권6호
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• pp.55-61
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• 2019

The purpose of this study was to numerically analyze the heat flow characteristics of an automotive radiator. Heat flow analyses were conducted on the cooling water and outdoor air of the radiator, as well as the temperature distribution of the cooling water after heat transfer. The results of the study revealed that neither heat transfer nor radiator volume was affected by the position of the inlet of cooling water. However, temperature distribution was affected by the position of both the inlet and outlet. In case of heat transfer, three models underwent about 158 kW of heat transfer. The difference in cooling water temperature was about $10^{\circ}C$. In case of pressure drop, the core external air side was reduced to about 1,375 Pa, and the internal cooling water side about 14,570 Pa.

• Nuclear Engineering and Technology
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• 제18권3호
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• pp.209-217
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• 1986

보론 희석이 불충분한 주기 말에서 부하추종 운전영역을 확장시키기 위해 냉 각재 입구 온도 감소 방식을 사용하였으며, 이 방식을 모사하고 부하추종 운전시 가압경수로 노심의 핵적 특성을 해석하기 위해 3차원적 전산체제를 확립하였다. 해석은 고리 1호기 1주기말에서 12-3-6-3부하추종 운전에 대해 MINB 및 SPINR 모드로 수행되었으며, 부가적으로 이 두 방식에 대한 출력복귀능력도 RETRAN02 코드를 사용한 계통해석과 더불어 시험하였다. 계산결과 냉각재 입구 온도를 감소시킴으로써 부하 추종운전이 주기 말에서도 가능함이 입증되었으며 14$^{\circ}$F의 입구 온도 감소에 따라, SRC가 제어봉만 사용하는 방식보다 MINB 모드에서는 13%, SPINR모드에서는 14%까지 증가됨을 보여주었다.

• 한국추진공학회지
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• 제21권3호
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• pp.49-55
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• 2017

가스터빈엔진의 성능시험을 위한 엔진 입구덕트를 1D 기법으로 Sizing 하였으며, 압축기 입구유동측정면(AIP, Aerodynamic interface plane)에서 경계층 두께를 최소화하고, 코어부 마하수분포가 균일하도록 설계하였다. 노즈콘 형상은 Haack-series 모델을 적용하고, 덕트 안쪽과 바깥쪽 면적변화율이 동일하도록 입구덕트 채널 바깥반경($r_o$)를 결정하여 설계목적을 구현하고자 하였으며, 이러한 형상이 설계목표에 부합하는지 확인하기 위하여 CFD를 수행하였다. AIP면에서 정압력분포는 최대값과 최소값 차이가 0.16% 이었으며, 마하수분포에서 경계층은 덕트반경 길의 2% 이내로 설계목표를 만족하였다. 이때 균일유동 코어부는 채널높이의 95% 이상이었다. 또한 입구유동의 전온도를 측정하기 위한 키엘 전 온도레이크 위치는 온도 회복계수가 최대화 되도록 마하수가 0.1 이하 지역인 노즈콘 전방 100 mm 이내이어야 함을 확인하였다.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.