• Nuclear Engineering and Technology
• /
• v.49 no.3
• /
• pp.621-631
• /
• 2017

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

• Proceedings of the KSME Conference
• /
• 2008.11b
• /
• pp.2505-2510
• /
• 2008

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor or a heavy water power reactor. When HANARO is normally operated, the fuel loaded in the irradiation hole has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operation condition of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. This paper describes the cold function test results of the MCWS. It was confirmed through the test results that the system met the design requirements under a cold operation condition.

• Journal of Energy Engineering
• /
• v.20 no.4
• /
• pp.259-266
• /
• 2011

Following the reactor shutdown, the reactor shutdown cooling system must be designed to supply the coolant sufficiently not only to remove the decay heat but to maintain the adequate cooling rate to protect the reactor equipments. In this study, KDESCENT code for the light water reactor and SOPHT, SDCS codes for the heavy water reactor were compared and analyzed to investigate the cooling capability during the shutdown cooling process. The shutdown cooling system design requirements were satisfied during cooling process for both the SDCP and the HTP modes and the design cooling rate of $2.8^{\circ}C/min$ or below was maintained using the SDC heat exchangers. This study shows that the shutdown cooling system in the Wolsong 2, 3, 4 reactors provides sufficient cooling to maintain the nuclear fuel integrity by removing the decay heat of the nuclear fission product.

• Journal of Environmental Science International
• /
• v.21 no.5
• /
• pp.563-572
• /
• 2012

Unlike water applications, the photocatalytic technique utilizing light-emitting-diodes as an alternative light source to conventional lamp has rarely been applied for low-level indoor air purification. Accordingly, this study investigated the applicability of UV-LED to annular-type photocatalytic reactor for removal of indoor-level benzene and toluene at a low concentration range associated with indoor air quality issues. The characteristics of photocatalyst was determined using an X-ray diffraction meter and a scanning electron microscope. The photocatalyst baked at $350^{\circ}C$ exhibited the highest photocatalytic degradation efficiencies(PDEs) for both benzene and toluene, and the photocatalysts baked at three higher temperatures(450, 550, and $650^{\circ}C$) did similar PDEs for these compounds. The average PDEs over a 3-h period were 81% for benzene and close to 100% for toluene regarding the photocatalyst baked at $350^{\circ}C$, whereas they were 61 and 74% for benzene and toluene, respectively, regarding the photocatalyst baked at $650^{\circ}C$. As the light intensity increased from 2.4 to 3.5 MW $cm^{-1}$, the average PDE increased from 36 to 81% and from 44% to close to 100% for benzene and toluene, respectively. In addition, as the flow rate increased from 0.1 to 0.5 L $min^{-1}$, the average PDE decreased from 81% to close to zero and from close to 100% to 7% for benzene and toluene, respectively. It was found that the annular-type photocatalytic reactor inner-inserted with UV-LEDs can effectively be applied for the decomposition of low-level benzene and toluene under the operational conditions used in this study.

• Nuclear Engineering and Technology
• /
• v.46 no.3
• /
• pp.281-290
• /
• 2014

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.9
• /
• pp.1175-1183
• /
• 2010

A spacer grid assembly is one of the most important structural components in a Light Water Reactor(LWR) nuclear fuel assembly. In the case of the Zircaloy spacer grid assembly, the primary design consideration is to ensure that lateral dynamic crush strength of the spacer grid assembly is sufficient to resist design basis loads and thereby prevent seismic accidents, without a significant increase in the hydraulic head loss for the reactor coolant in the reactor core. In this study, factors affecting the lateral dynamic crush strength of a spacer grid assembly were analyzed by performing lateral dynamic crush tests and finite element analyses. Further, an effective and economical method to enhance the lateral dynamic crush strength of the spacer grid assembly is proposed.

• Nuclear Engineering and Technology
• /
• v.52 no.5
• /
• pp.878-888
• /
• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

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

• Journal of the Korea institute for structural maintenance and inspection
• /
• v.5 no.2
• /
• pp.147-154
• /
• 2001

The prestressed concrete containment is one of the most important structures in nuclear power plants, which is required to prevent release of radioactive or hazardous effluents to the environment even in the case of a severe accident. Numerical analyses are carried out by using the ABAQUS finite element program to assess the ultimate pressure capacity of the Y prestressed concrete containment with light water reactor at design criteria condition and aging condition considering varied properties of time-dependant materials respectively. From the results, it is verified that the structural capacity of the Y prestressed concrete containment building under the present, aging condition is still robust. In addition, the parameter studies for the reduction of the ultimate pressure capacity of containment building according to the degradation levels of the main structural materials are carried out. The results show that when the degradations of each materials are considered as individual and combined forms, the influence is large in the order of tendon, rebar and concrete degradation, and tendon-rebar, tendon-concrete and rebar-concrete degradation respectively.

• Korean Journal of Materials Research
• /
• v.29 no.1
• /
• pp.7-10
• /
• 2019

Cr-Al alloys are attracting attention as oxidation resistant coating materials for high temperature metallic materials due to their excellent high temperature stability. However, the mechanical properties and oxidation resistance of Cr-Al alloys can be further enhanced, and such attempts are made in this study. To improve the properties of Cr-Al alloys, Si is added up to 5 wt%. Casting specimens with different amounts of Si content are prepared by a vacuum arc remelting method and isothermally heated under steam conditions at $1,100^{\circ}C$ for 1 hour. The as-cast microstructure of low Si alloys is mainly composed of only a Cr phase, while $Al_8Cr_5$ and $Cr_3Si$ phases are also observed in the 5 % Si alloy. In the high Si alloy, only Cr and $Cr_3Si$ phases remain after the isothermal heating at $1,100^{\circ}C$. It is found that Si additions slightly decrease the oxidation resistance of the Cr-Al alloy. However, the microhardness of the Cr-Al alloy is observed to increase with an increasing Si content.