• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2670-2689
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• 2022

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS - also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4685-4694
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• 2023

The ability to calculate the material density sensitivity coefficients of power with respect to the material density has broad application prospects for accelerating Monte Carlo-Thermal Hydraulics iterations. The second-order material density sensitivity coefficients for the general Monte Carlo score have been derived based on the differential operator sampling method in this paper, and the calculation of the sensitivity coefficients of cell power scores with respect to the material density has been realized in continuous-energy Monte Carlo code RMC. Based on the power-density sensitivity coefficients, the sensitivity coefficients of power scores to some other physical quantities, such as power-boron concentration coefficients and power-temperature coefficients considering only the thermal expansion, were subsequently calculated. The effectiveness of the proposed method is demonstrated in the power-density coefficients problems of the pressurized water reactor (PWR) moderator and the heat pipe reactor (HPR) reflectors. The calculations were carried out using RMC and the ENDF/B-VII.1 neutron nuclear data. It is shown that the calculated sensitivity coefficients can be used to predict the power scores accurately over a wide range of boron concentration of the PWR moderator and a wide range of temperature of HPR reflectors.

• Journal of Korea Water Resources Association
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• v.55 no.spc1
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• pp.1211-1222
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• 2022

Abnormal condition inevitably occurs during operation of water distribution system (WDS) and requires the isolation of certain areas using isolation valves. In general, the determination of the optimal location of isolation valves considered minimization of hydraulic failures as isolation of certain areas causes a change in hydraulic states (e.g., flow direction, velocity, pressure, etc.). Water quality failure can also be induced by changes in hydraulics, which have not been considered for isolation valve system design. Therefore, this study proposes a new isolation valve system design methodology to prevent unexpected water quality failure events. The new methodology considers flow direction change ratio (FDCR), which accounts for flow direction changes after isolation of the area, as a constraint while reliability is used as the objective function. The optimal design model has been applied to a synthetic grid network and the results are compared with the traditional design approach. Results show that considering FDCR can eliminate flow direction changes while average pressure and coefficient of variation of pressure, velocity, and hydraulic geodesic index (HGI) outperform compared to the traditional design approach. The proposed methodology is expected to be a useful approach to minimizing unexpected consequences by traditional design approaches.

• 한국전산유체공학회:학술대회논문집
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• 2009.04a
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• pp.290-297
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• 2009

A two-phase (gas and liquid) flow analysis solver, named CUPID, has been developed for a realistic simulation of transient two-phase flows in light water nuclear reactor components. In the CUPID solver, a two-fluid three-field model is adopted and the governing equations are solved on unstructured grids for flow analyses in complicated geometries. For the numerical solution scheme, the semi-implicit method of the RELAP5 code, which has been proved to be very stable and accurate for most practical applications of nuclear thermal hydraulics, was used with some modifications for an application to unstructured non-staggered grids. This paper is concerned with the effects of interpolation schemes on the simulation of two-phase flows. In order to stabilize a numerical solution and assure a high numerical accuracy, the second-order upwind scheme is implemented into the CUPID code in the present paper. Some numerical tests have been performed with the implemented scheme and the comparison results between the second-order and first-order upwind schemes are introduced in the present paper. The comparison results among the two interpolation schemes and either the exact solutions or the mesh convergence studies showed the reduced numerical diffusion with the second order scheme.

• Journal of Energy Engineering
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• v.14 no.4 s.44
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• pp.232-240
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• 2005

In this study, the IAEA Benchmark problems far HTR-10 and HTTR RCCS were assessed in order to assess the applicability of MARS code, a thermal-hydraulic safety analysis code developed for water reactors. The calculated results were compared with those or THERMIX, THANPACST2 code, and available experimental data. The calculated results showed generally good agreements with those obtained by the THERMIX code and THANPACST2 code. Deviations were analyzed to be originated from the simplification of complicated geometry and from the modeling capability of heat transfer characteristics in the HTGR components such as water cooler and air tooler. Especially, it was found that the radiation heat transfer in the reactor cavity played an important role in the after heat removal in the RCCS. Thus, it is concluded that MARS code can be successfully applied to the calculation of the RCCS cooling capability of the HTGR in this study.

• Journal of Korea Water Resources Association
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• v.42 no.11
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• pp.989-1004
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• 2009

The objective of this study is to examine validation of Grid Reconstruction Method, which is developed to simulate drying/wetting in complex natural rivers with wetting and drying domain areas. To verify application of the developed model, the model was applied to natural rivers with wetting and drying domain areas such as Han river and Nakdong river. The simulation results have shown good agreements with observed data and the results for the developed model were more accurate and improved stability of numerical computation than those of RMA-2 model. If the analysis of contaminant advection-diffusion and sediment transport are performed with the study results, the results can be effectively applied to river flow analysis and ecological hydraulics.

• Water for future
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• v.29 no.3
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• pp.143-152
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• 1996

The hydraulics of flow within the covered reach of urban streams is very complicated due to the accumulation and interference effect of eddies around the multiple piers supporting the covering slab. An extensive experimental study is done to quantitatively estimate the backwater rise effect caused by various arrays of multiple piers. The factors governing the backwater rise are found out to be the contraction ratio due to the piers, Froude number of the flow, longitudinal pier spacing, and the length of the covered reach. For a single section of lateral pier arrays the effect of contraction ratio and Froude number on the backwater rise is analyzed and a multiple regression equation is derived. The effect of multiple piers, arrayed in both lateral and longitudinal directions, on the backwater rise is analyzed in terms of the contraction ratio, Froude number,longitudinal pier spacing and the total length of the covered reach. A multiple regression equation for the backwater rise estimation is proposed based on the experimental data collected in the present study.

• Proceedings of the Korea Water Resources Association Conference
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• 2017.05a
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• pp.281-281
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• 2017

하천에서 하도불안정(stream instability)으로 인하여 하상의 형태가 변화하고 하상파(sand wave)가 발생한다. 사련(ripple), 사구(dune) 등과 같은 하상파는 흐름저항을 유발하여 홍수시 수위를 증가시킨다. 수리실험 및 수치모형을 이용하여 사련 및 사구의 발달과정 그리고 이를 지나는 난류흐름에 대한 연구가 국외에서는 이루어지고 있지만 국내의 경우 거의 찾아보기 힘들다. 수치모형을 활용한 연구는 주로 횡방향으로 하상파가 일정하다는 가정하에 연직 2차원 수치모형을 적용하였으나 최근 컴퓨터 기술 및 수치기법의 고도화로 3차원 RANS(Reynolds averaged Navier-Stokes) 또는 LES(Large eddy simulation)를 이용한 수치모형이 개발되고 있다. 본 연구에서는 LES에 유사이송 및 하상변동 모형과 결합하여 사구발달에 대한 수치모의를 수행하였다. LES와 유사이송 및 하상변동 모형의 결합은 순간유속성분을 하상변동모형에 직접 적용되기 때문에 난류영향을 고려할 수 있는 것이 장점이다. 특히 사구의 발달에 따라 복잡한 흐름이 발생하며 3차원 와구조가 발생하므로 난류특성의 고려는 필수적이다. 수치모의는 Delft Hydraulics (Bakker et al., 1986)에서 수행한 수리실험 T39를 활용하였다. 수리실험은 길이 100 m, 폭 0.5 m 개수로에서 수행되었으며 평균유속은 0.611 m/s, 수심은 0.436 m이다. 하상파 실험에 사용된 유사입경은 0.78 mm 균일사를 사용하였다. 수치모의 조건은 수리실험과 동일하게 하였으나 계산시간의 효율을 고려하여 흐름방향의 계산영역은 4.0 m로 하고 주기경계조건(periodic boundary condition)을 부여하여 계산을 수행하였다. 수치모의 계산은 사구의 길이 및 파고가 평형상태에 이를 때까지 수행되었다. 수치모의 통해 사구발달에 따른 흐름 및 하상변동 특성을 분석하였다.

• Proceedings of the Korea Water Resources Association Conference
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• 2015.05a
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• pp.245-245
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• 2015

In the recent years, due to long-lasting heavy rainfall events, a large number of landslides have been observed in the mountainous area of the world. Such landslides can also form a dam as it blocks the course of a river, which may burst and cause a catastrophic flood. Numerical analysis of landslide dam formation is rarely available, while laboratory experimental studies often use assumed shape to analyze the landslide dam failure and flood hydraulics in downstream. In this study, both experimental and numerical studies have been carried out to investigate the formation of landslide dam. Two case laboratory experiments were conducted in two flumes simultaneously. The first flume (2.0 m 0.6 m 0.5 m) was set at $22^{\circ}$ and $27^{\circ}$ slope to generate the landslide using rainfall intensity of 70.0 mm/hr. On the other hand, the second flume (1.5 m 0.25 m 0.3 m) was set perpendicularly at the downstream end of the first flume to receive the landslide mass forming landslide dam. The formation of landslide dam was observed at $15^{\circ}$ slope of the second flume. The whole processes including the landslide initiation and movement of the landslide mass into the second channel was captured by three digital cameras. In numerical analysis, a two-dimensional (2D) seepage flow model, a 2D slope stability model (Spencer method) and a 2D landslide dam-geometry evaluation model were coupled as a single unit. This developed model can determine the landslide occurrence time, the failure mass and the geometry of landslide dam deposited in the second channel. The data obtained from numerical simulation results has good agreement with the experimental measurements.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.