• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Corrosion Science and Technology
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• v.19 no.1
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• pp.1-7
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• 2020

Chalk River Unidentified Deposits (CRUDs) were collected from the Korean pressurized water reactor (PWR) plant (A, B, and C) where the axial offset anomaly (AOA) occurred. AOA, also known as a CRUD-induced power shift, is one of the key issues in maintaining stable PWR plant operations. CRUDs were sampled from spent nuclear fuel rods and analyzed using an electron probe micro-analyzer (EPMA). This paper describes the characteristics of boron-deposits from the CRUDs sampled from twice-burnt assemblies from the Korean PWR. The primary coolant of a PWR contains boron and lithium. It is known that boron deposition occurs in a thick CRUD layer under substantial sub-cooled nucleate boiling (SNB). The results of this study are summarized as follows. Boron was not found at the locations where the existence was confirmed in simulated CRUDs, in other words, the cladding and CRUD boundaries. Nevertheless, we clearly observed the presence of boron and confirmed that boron existed as a lump in crystalline form. In addition, the study confirmed that CRUD existed in a crystal form with a unique size of about 10 ㎛.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• Korean Journal of Materials Research
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• v.25 no.8
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• pp.376-385
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• 2015

The effects of Nb and Cr addition on the microstructure, corrosion and oxide characteristics of Zr based alloys were investigated. The corrosion tests were performed in a pressurized water reactor simulated-loop system at $360^{\circ}C$. The microstructures were examined using OM and TEM, and the oxide properties were characterized by low-angle X-ray diffraction and TEM. The corrosion test results up to 360 days revealed that the corrosion rates were considerably affected by Cr content but not Nb content. The corrosion resistance of the Zr-xNb-0.1Sn-yCr quaternary alloys was improved by an increasing Nb/Cr ratio. The crystal structure of the precipitates was affected by a variation of the Nb/Cr ratio. The Zr-Nb beta-enriched precipitates were mainly formed in the high Nb/Cr ratio alloy while $Zr(NbCr)_2$ precipitates were frequently observed in the low Nb/Cr ratio alloy. The studies of oxide characteristics revealed that the corrosion resistance was related to the crystal structure of the precipitate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.289-295
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• 2016

Kori Unit 1, the first PWR (Pressurized Water Reactor) plant in Korea, began its commercial operation in 1978 and will permanently shut down on June 18, 2017. After moving the spent fuels to SFP (Spent Fuel Pool) system, Kori Unit 1 will perform a full system chemical decontamination to reduce radiation levels inside the various plant systems. This paper will describe the operation concept of the full system chemical decontamination for Kori Unit 1 based on experiences overseas.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.339-350
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• 2016

Based on the resonance integral (RI) tables produced by the NJOY program, the conventional subgroup method usually ignores both the resonance elastic scattering and the resonance interference effects. In this paper, on one hand, to correct the resonance elastic scattering effect, RI tables are regenerated by using the Monte Carlo code, OpenMC, which employs the Doppler broadening rejection correction method for the resonance elastic scattering. On the other hand, a fast resonance interference factor method is proposed to efficiently handle the resonance interference effect. Encouraging conclusions have been indicated by the numerical results. (1) For a hot full power pressurized water reactor fuel pin-cell, an error of about +200 percent mille could be introduced by neglecting the resonance elastic scattering effect. By contrast, the approach employed in this paper can eliminate the error. (2) The fast resonance interference factor method possesses higher precision and higher efficiency than the conventional Bondarenko iteration method. Correspondingly, if the fast resonance interference factor method proposed in this paper is employed, the $k_{inf}$ can be improved by ~100 percent mille with a speedup of about 4.56.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.794-800
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• 2018

The proliferation resistance (PR) of Th/U and Th/Pu fuels used in CANada Deuterium Uranium for the deep geological repository was assessed by combining the multiattribute utility analysis proposed by Chirayath et al., 2015 with the transport model of radionuclides in the repository and comparing with that of the used natural U fuel case. It was found that there was no significant advantage for Th/U and Th/Pu fuels from the viewpoint of the PR in the repository. It was also found that the PR values for used nuclear fuels in the repository of Th/U, Th/Pu, and natural U was comparable with those for enrichment and reprocessing facilities in the pressurized water reactor (PWR) nuclear fuel cycle. On the other hand, the PR values considering the transport of radionuclides in the repository were found to be slightly smaller than those without their transport after the used nuclear fuels started dissolving after 1,000 years.

• Korean Journal of Metals and Materials
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• v.50 no.8
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• pp.583-589
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• 2012

This work is concerned with a statistical approach to the life prediction on irradiation-assisted stress corrosion cracking (IASCC) of stainless steel (SS) for core internals of a pressurized water reactor (PWR). The previous results of the time-to-failure of IASCC measured on neutron-irradiated stainless steel components were statistically analyzed in terms of stress and irradiation. The accelerating life testing model of IASCC of cold worked Type 316 SS was established based on an inverse power model with two stress-variables, the applied stress and irradiation dose. Considering the variation of the yield strength and applied stress with the irradiation dose in the model, the remaining life of the baffle former bolt was statistically predicted during operation under complex environments of stress and irradiation.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1171-1180
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• 2009

The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The $Electricit\acute{e}$ De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code_Saturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer.