• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.290-295
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• 1996

The purpose of this study is to establish the pressure/temperature curves of Reactor Coolant System for brittle fracture prevention. The pressure/temperature curve is the basis to select RC Pump and limits to operate the plant. Based on the plant operation experience, this curve should be re-generated periodically in order to ensure the structural integrity using data from the test of reactor vessel surveilance materials to compensate for the irradiation effects. This study provides the procedure of pressure/temperature curve generation in term of brittle fracture prevention of reactor vessel. Using the UCN 3&4 data, the sample pressure/temperature curve was generated, and it was compared with those of YGN 3&4 based on the stress and $RT_{NDT}$value.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.771-776
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• 1998

TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Vnit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) far all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library. BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The makimum fast neutron nun calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effgctive full power days is 1.784x10$^{18}$ n/$\textrm{cm}^2$. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60cm below the midplane at zero degree.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.44-50
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• 2014

There were many flaw issues of reactor vessel head penetration in USA fleets. USNRC issued 10CFR50.55a to implement reactor vessel head penetration ultrasonic examination performance demonstration(PD) in US for enhancement of inspection reliability. After September 2009, all US utilities inspected their RVHP with PD qualified system. Korea Hydro and Nuclear Power Company(KHNP) have developed reactor vessel head penetration performance demonstration system for ultrasonic test to apply for pressurized light-water reactor power plants in accordance with 10CFR50.55a since September 2011. RVHP configuration surveying and analysis, code requirement analysis, and performance demonstration specimen design were performed up to this day. Fingerprinting of manufactured specimen, development of test data management program, development of operation procedure, input of flawed data, and development of final report will be performed for the next step. This paper describes the development status of the performance demonstration system for reactor vessel head penetration ultrasonic examination in Korea.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.85-92
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• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.3
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• pp.191-198
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• 2014

In this paper, structural integrity evaluation of reactor pressure vessel bottom head without penetration nozzles in core melting accident has been performed. Considering the analysis results of thermal load, weight of molten core debris and internal pressure, thermal load is the most significant factor in reactor vessel bottom head. The failure probability was evaluated according to the established failure criteria and the evaluation showed that the equivalent plastic strain results are lower than critical strain failure criteria. Thermal-structural coupled analyses show that the existence of elastic zone with a lower stress than yield strength is in the middle of bottom head thickness. As a result of analysis, the elastic zone became narrow and moved to the internal wall as the internal pressure increases, and it is evaluated that the structural integrity of reactor vessel is maintained under core melting accident.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Nuclear Engineering and Technology
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• v.33 no.5
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• pp.498-513
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• 2001

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.