• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.515-519
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• 1997

The purpose of this study is to reduce the fast neutron fluence at the reactor pressure vessel(RPV) and to provide a basis for plant-life extension. In this study, different neutron absorbers were employed in the core outer assemblies of Kori Unit 1 Cycle 14. The modified assemblies were used to calculate fast neutron fluence at the RPV and to evaluate reduction of outer assembly power and total power in core. By comparison with the case of no suppression fixture, the fast neutron fluence of a case with two rows stainless steel around the assembly with natural uranium pins is decreased by 85.8%. It is noted that the modification of outer assembly is more efficient than the previous low leakage loading pattern (LLLP) applied to Kori Unit 1. Also, compared fast neutron fluence in Cycle 1 with Cycle 14, fast neutron fluence at the RPV between Cycle 1 and Cycle 14 is not significantly different. It is found that LLLP applied to the Kori Unit 1 has not contributed to fast neutron fluence reduction at the RPV.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• 대한기계학회논문집A
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• 제21권7호
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• pp.1089-1096
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• 1997

In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.

• Nuclear Engineering and Technology
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• 제40권2호
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• pp.117-126
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• 2008

In this paper, a new reactor core design is proposed on the basis of a mixed core concept consisting of a thermal zone and a fast zone. The geometric structure of the fuel assembly of the thermal zone is similar to that of a conventional thermal supercritical water-cooled reactor(SCWR) core with two fuel pin rows between the moderator channels. In spite of the counter-current flow mode, the co-current flow mode is used to simplify the design of the reactor core and the fuel assembly. The water temperature at the exit of the thermal zone is much lower than the water temperature at the outlet of the pressure vessel. This lower temperature reduces the maximum cladding temperature of the thermal zone. Furthermore, due to the high velocity of the fast zone, a wider lattice can be used in the fuel assembly and the nonuniformity of the local heat transfer can be minimized. This mixed core, which combines the merits of some existing thermal SCWR cores and fast SCWR cores, is proposed for further detailed analysis.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3242-3249
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• 2022

As Kori unit 1 is undergoing the decommissioning process, estimating the disposal amount of waste from the decommissioned nuclear reactor has become one of the challenging issues. Since the waste disposal amount estimation depends on the packing of the waste, it is highly desirable to optimize the waste packing plan. In this study, we developed an efficient scheme for packing waste component segments. The scheme consists of 1) preparing three-dimensional models of segments, 2) orienting each segment in such a way to minimize the bounding box volume, and 3) applying hybrid genetic algorithm to pack the segments in the disposal containers. When the packing solution converges in the algorithm, it comes up with the number of containers used and the placement of segments in each container. The scheme was applied to Kori-1 reactor pressure vessel. The required number of containers calculated by the developed scheme was 24 compared to 42 that was the estimation of the prior packing plan, resulting in disposal volume savings by more than 40%. The developed method is flexible for applications to various packing problems with waste segments from different cutting options and different sizes of containers.

• 대한기계학회논문집A
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• 제26권11호
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• pp.2390-2398
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• 2002

For integrity analysis of nuclear reactor pressure vessel, including the Pressurized thermal shock analysis, the fast and accurate calculation of the stress intensity factor at the crack tip is needed. For this, a simple approximation scheme is developed and the resulting stress intensity factors for axial semi-elliptical cracks in cylindrical vessel under various loading conditions are compared with those of the finite element method and other approximation methods, such as Raju-Newman's equation and ASME Sec. Xl approach. For these, three-dimensional finite-element analyses are performed to obtain the stress intensity factors for various surface cracks with t/R = 0.1. The approximation methods, incorporated in VINTIN (Vessel INTegrity analysis-INner flaws), utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite clement analysis for internal pressure, cooldown, and pressurized thermal shock loading conditions. The approximation solutions are within $\pm$2.5% of the those of FEA using symmetric model of one-forth of a vessel under pressure loading, and 1-3% higher under pressurized thermal shock condition. The analysis results confirm that the VINTIN method provides sufficiently accurate stress intensity factor values for axial semi-elliptical flaws on the surface of the reactor pressure vessel.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2836-2846
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• 2020

During severe nuclear power plant (NPP) accidents, a H₂/CO mixture can be generated in the reactor pressure vessel by core degradation and in the containment as well by molten corium-concrete interaction. In spite of its importance, a state-of-the-art methodology predicting H₂/CO combustion risk relies predominantly on empirical correlations. It is therefore necessary to develop a proper methodology for flammability evaluation of H₂/CO mixtures at ex-vessel phases characterized by three factors: CO concentration, high temperature, and diluents. The developed methodology adopted Le Chatelier's law and a calculated non-adiabatic flame temperature model. The methodology allows the consideration of the individual effect of the heat transfer characteristics of hydrogen and carbon monoxide on low flammability limit prediction. The accuracy of the developed model was verified using experimental data relevant to ex-vessel phase conditions. With the developed model, the prediction accuracy was improved substantially such that the maximum relative prediction error was approximately 25% while the existing methodology showed a 76% error. The developed methodology is expected to be applicable for flammability evaluation in chemical as well as NPP industries.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.61-68
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• 2017

Reactor pressure vessel (RPV) under severe accident conditions accompanied by core melting is exposed to direct high-temperature thermal loads. Understanding the creep behavior of the material is one of the most important factors for evaluating the structural integrity at these conditions. While damage evaluation studies have been conducted on critical structures of nuclear power plants through finite element (FE) analyses considering creep behavior, for accurate creep damage evaluation, constitutive equations considered in the FE analyses may have different results depending on the time hardening and strain hardening models as well as the tertiary creep consideration. The purpose of this study is to evaluate the creep damage under severe accident conditions by using FE method for a representative domestic RPV material, SA508 Gr.3. The effect of material hardening models and constitutive equations which are the main variables were also investigated.

• 한국소음진동공학회논문집
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• 제22권5호
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.