• 한국소음진동공학회논문집
• /
• 제23권4호
• /
• pp.308-314
• /
• 2013

The vibration and stress analysis program of comprehensive vibration assessment program(CVAP) is to theoretically verify the structural integrity of reactor vessel internals(RVI) and to provide the basis for selecting the locations monitored in measurement and inspection programs. This paper covers the applicability of the vibration and stress analysis method of APR1400 RVI CVAP. The analysis method was developed to use 3-dimensional detail hydraulic and structural models with ANSYS and CFX. To assess the method, the hydraulic loads and structural reponses of OPR1000 were predicted and compared with the measured data in the OPR1000 RVI CVAP. The results predicted with this method were close to the measured values considerably. Therefore, the analysis method was developed properly.

• Nuclear Engineering and Technology
• /
• 제53권12호
• /
• pp.3892-3901
• /
• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• Nuclear Engineering and Technology
• /
• 제48권6호
• /
• pp.1291-1302
• /
• 2016

This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

• Nuclear Engineering and Technology
• /
• 제41권7호
• /
• pp.921-928
• /
• 2009

As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

• Nuclear Engineering and Technology
• /
• 제51권8호
• /
• pp.1939-1950
• /
• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• Nuclear Engineering and Technology
• /
• 제41권8호
• /
• pp.1005-1014
• /
• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
• /
• 제50권8호
• /
• pp.1217-1233
• /
• 2018

Following a surge of interest in multi-unit risk in the last few years, many recent studies have suggested methods for multi-unit probabilistic safety assessment (MUPSA) and addressed several related aspects. Most of the existing studies though focused on two-unit nuclear power plant (NPP) sites or used rather simplified probabilistic safety assessment (PSA) models to demonstrate the proposed approaches. When considering an NPP site with three or more units, some approaches are inapplicable or yield very conservative results. Since the number of such sites is increasing, there is a strong need to develop and validate practical approaches to the related MUPSA. This article provides several detailed approaches that are applicable to multi-unit Level 1 PSA for sites with up to six or more reactor units. To validate the approaches, a multi-unit Level 1 PSA model is developed and the site core damage frequency is estimated for each of four representative multi-unit initiators, as well as for the case of a simultaneous occurrence of independent single-unit initiators in multiple units. For this purpose, an NPP site with six identical OPR-1000 units is considered, with full-scale Level 1 PSA models for a specific OPR-1000 plant used as the base single-unit models.

• Nuclear Engineering and Technology
• /
• 제51권6호
• /
• pp.1487-1503
• /
• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• Nuclear Engineering and Technology
• /
• 제55권12호
• /
• pp.4607-4616
• /
• 2023

We implement machine learning regression models to predict peak pressures of primary and secondary systems, a major safety concern in Loss Of Condenser Vacuum (LOCV) accident. We selected the Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code to analyze the LOCV accident, and the reference plant is the Korean Optimized Power Reactor 1000MWe (OPR1000). eXtreme Gradient Boosting (XGBoost) is selected as a machine learning tool. The MARS-KS code is used to generate LOCV accident data and the data is applied to train the machine learning model. Hyperparameter optimization is performed using a simulated annealing. The randomly generated combination of initial conditions within the operating range is put into the input of the XGBoost model to predict the peak pressure. These initial conditions that cause peak pressure with MARS-KS generate the results. After such a process, the error between the predicted value and the code output is calculated. Uncertainty about the machine learning model is also calculated to verify the model accuracy. The machine learning model presented in this paper successfully identifies a combination of initial conditions that produce a more conservative peak pressure than the values calculated with existing methodologies.

• 에너지공학
• /
• 제22권2호
• /
• pp.148-155
• /
• 2013

한국표준형원전(OPR-1000)의 증기발생기 전열관은 관재료로 Alloy 600과 Alloy 690이 사용되고 있으며 전열관 확관방법은 폭발확관법이 적용되었다. 원전 운전경험에 따르면 전열관 확관법으로 폭발확관법이 사용된 증기발생기의 전열관들은 원주방향 균열이 잘 발생하였으며 그 위치는 전열관의 확관천이부위 외면이었고 슬러지에 의해 둘러싸인 지역이었다. 그러나 같은 조건의 전열관이라도 수압확관법이 적용된 경우는 원주방향 균열보다는 축방향 균열이 우세한 경향을 보여왔다. 따라서 본 연구에서는 상기와 같은 차이를 규명하기 위하여 전열관과 관판의 형상을 모델링하였으며 확관법에 따라 운전중 조건에서 전열관의 확관천이 부위에 작용하는 응력의 크기를 전산프로그램을 사용하여 계산하였고, 균열의 방향성과 우세성을 평가하였다.