• Nuclear Engineering and Technology
• /
• v.42 no.3
• /
• pp.259-270
• /
• 2010

New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.824-829
• /
• 1998

미국의 개량형 원자력 발전소 개념설계단계에서 중대사고시 사고완화를 위한 전략으로 원자로 압력용기 외부냉각 개념이 제안되었다. 중대사고 진행과정에서 노심용융물이 원자로 압력용기 하부헤드로 재배치 되었을 때 압력용기 외벽을 냉각함으로서 노심용융물을 압력용기 내부에 가두어 두어 격납건물 내로의 유출을 방지하는 방식이다. 이 연구에서는 원자로 압력용기 하부헤드 내의 노심용융물 거동중 자연 순환에 의한 거동을 수치적으로 모의하여 보았다. 연구결과, 정상상태의 온도 및 속도분포는 현상학적으로 적절하게 모의되나 고화와 액화의 경우에는 고유모델의 필요성이 요구되었다.

• Journal of Korea Foundry Society
• /
• v.40 no.5
• /
• pp.129-133
• /
• 2020

The outcomes of solution annealing and stress corrosion cracking in cold-worked 316L austenitic stainless steel have been studied using x-ray diffraction (XRD) and the slow strain rate test (SSRT) technique. The good compatibility with a high-temperature water environment allows 316L austenitic stainless steel to be widely adopted as an internal structural material in light water reactors. However, stress corrosion cracking (SCC) has recently been highlighted in the stainless steels used in commercial pressurized water reactor (PWR) plants. In this paper, SCC and inter granular cracking (IGC) are discussed on the basis of solution annealing in a chloride environment. It was found that the martensitic contents of cold-worked 316L stainless steel decreased as the solution annealing time was increased at a high temperature. Moreover, mode of SCC was closely related to use of a chloride environment. The results here provide evidence of the vital role of a chloride environment during the SCC of cold-worked 316L.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.54-67
• /
• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
• /
• v.34 no.3
• /
• pp.259-267
• /
• 2002

NFPA-803 has been referred as the Fire Protection Standard at the Nuclear Power Plants of Pressurized Water Reactor. This Standard has been used as the fire protection regulation, containing prescriptive requirements with deterministic methodology. Recently, with cumulative efforts by the U.S. Nuclear Regulatory Commission and Utilities in America to establish a new Standard, including a quantitative evaluation methodology, NFPA-805, the Performance-Based Standard for FIRE Protection for Light Water Reactor Electric Generating Plants was issued and approved by the American National Standards Institute as an American National Standard with an effective date of February 9, 2001. This paper presents an analysis result from the computer modeling for the fire simulation In addition, it proposes the idea that this kind of analytic method can be available for the facilities design of fire prevention and protection fields, as well as an evaluation for the fire suppression system with a quantitative analysis for the thermal phenomena in fire compartments in Nuclear Power Plants.

• Journal of the Korean Society of Safety
• /
• v.36 no.2
• /
• pp.111-119
• /
• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Nuclear Engineering and Technology
• /
• v.40 no.4
• /
• pp.255-268
• /
• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

• Proceedings of the KSME Conference
• /
• 2001.11a
• /
• pp.355-360
• /
• 2001

Design requirements for the nuclear fuel assembly grid of the pressurized light water reactor(PLWR) are scrutinized from the mechanical/structural point of view. As a result of the scrunity, mechanical/structural test facilities on the spacer grid of the PLWR Fuel are set up in KAERI to find out their mechanical/structural performance.

• Journal of the Korean Society of Safety
• /
• v.33 no.3
• /
• pp.92-98
• /
• 2018

The flow of cooling water in a passive containment cooling system (PCCS), used to remove heat released in design basis accidents from a concrete containment of light water nuclear power plant, was conducted in order to investigate the thermo-fluid equilibrium among many parallel tubes of PCCS. Numerical simulations of the subcooled boiling flow within a coolant loop of a PCCS, which will be installed in innovative pressurized-water reactor (PWR), were conducted using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. As the first step, the simplified geometry of PCCS with 36 tubes was modeled in order to reduce computational resource. Even and uneven thermal loading conditions were applied at the outer walls of parallel tubes for the simulation of the coolant flow in the PCCS at the initial phase of accident. It was observed that the natural circulation maintained in single-phase for all even and uneven thermal loading cases. For uneven thermal loading cases, coolant velocity in each tube were increased according to the applied heat flux. However, the flows were mixed well in the header and natural circulation of the whole cooling loop was not affected by uneven thermal loading significantly.

• Environmental and Resource Economics Review
• /
• v.27 no.2
• /
• pp.261-286
• /
• 2018

This paper evaluated the economic feasibility of the life extension of Kori unit 1 and Wolsong unit 1 according to the types of the nuclear power plants (NPPs) and the life extension period comparing to the levelized costs of energy (LCOE) of the new NPPs, coal-fired plants (CFPs), and combined cycle gas turbine (CCGTs) which proposed in the $7^{th}$ Basic Plan for Electricity Supply and Demand. The economic feasibility of the life extension of NPPs using LCOE method is affected by the types of NPPs, lifetime extension periods, discount rate, and capacity factor. According to the analysis results, the pressurized light water reactor (PWR) is more economical than the pressurized heavy water reactor (PHWR). Comparing the economical efficiency between the life extension of NPPs and other alternatives, the operation of the PWR for 20 years is more economical than the one of new NPPs and CFPs. However, 20 years of life extension of PHWR is more economical than the CCGTs, but less economical than new NPPs and CFPs. In summary, the 20 years of life extension of the NPPs seems to be more, especially for the PWR, which is more cost effective than other generation alternatives. Therefore, the government policy of the life extension of NPPs need to be a selective approach that simultaneously considers both safety and economics rather than closing all NPPs.