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

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.463-478
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• 2019

The most important non-functional requirements for dependability of any Embedded Real-Time Safety Systems are safety, availability and reliability requirements. System architecture plays the primary role in achieving these requirements. Compliance with these non-functional requirements should be ensured early in the development cycle with appropriate considerations during architectural design. In this paper, we present an application of system architecture modeling for quantitative assessment of system dependability. We use probabilistic model checker (PRISM), for dependability analysis of the DTMC model derived from system architecture model. In general, the model checking techniques do not scale well for analyzing large systems, because of prohibitively large state space. It limits the use of model checking techniques in analyzing the systems of practical interest. We propose abstraction based compositional analysis methodology to circumvent this limitation. The effectiveness of the proposed methodology has been demonstrated using the case study involving the dependability analysis of safety system of a large Pressurized Water Reactor (PWR).

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.420-424
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• 1994

This paper presents an improved water level control scheme for Pressurized Water Reactor(PWR) Steam Generator(S/G) at the low power operation and transient states. To reduce fluctuations of the water level by the swell and shrink phenomena, the scheme adds feedforward terms considering S/G pressure and the feedwater temperature into the conventional proportional-integral feedback controller. The simulation results using the Compact Nuclear Simulator show that smaller level errors and much faster settling time than those of the conventional scheme can be obtained. The proposed algorithm is easily implementable and has a potential for the real applications.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.371-380
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• 1993

A model free controller utilizing a set of linguistic fuzzy logic of the human operator's experience is developed to control the steam generator water level in a pressurized water reactor. Only 9 rules for control action are generated from the inputs of water level error and mass flow error implicitly representing the time variation of the collapsed water level. The bell type membership functions of the premise side and the result side are tuned by the sensitivity study. This compact fuzzy logic controller shows a robust control during transient and no offset error and oscillation during steady state operation. For a multi-ramp power increase from start-up to full power, the proposed controller shows good performance for the entire range.

• Journal of the Korean Society for Nondestructive Testing
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• v.25 no.4
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• pp.294-303
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• 2005

This paper describes the analysis results of a series f Round-Robin test that was performed to intercompare inspection and diagnosis techniques for characterization of pressure tube f a pressurized heavy water reactor under the Coordinated Research Project(CRP) of IAEA's nuclear Power Programme. For this test, six nations, Korea, Canada, India, Argentina, Rumania, and China that currently have pressurized heavy water reactors under operation involved, and the "KOR-1" pressure tube sample prepared by Korea was used. Two kinds of NDE technique, ultrasonic and eddy current test, were applied for these tests. The "KOR-1" pressure tube sample contains total 12 artificial flaws such as crack-like EDM notches, wear that is similar to the real flaws and can be produced on the pressure tubes during plant operation. Test results showed that seven laboratories from six nations detected all twelve flaws in "KOR-1" specimen by using ultrasonic and eddy current test methods, and ultrasonic test method was more accurate than eddy current test method in flaw detectin and sizing. ID flaws in pressure tube sample were more easily detected and accurately sized than OD flaws.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.1950-1958
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• 2024

For the pressurized water reactor two-step calculation, the traditional assembly homogenization and two-group neutron diffusion calculation have been widely used. When it comes to the core pin-by-pin simulation, many models and techniques are different and unsettled. In this paper, the homogenization methods based on the pin discontinuity factors and super homogenization factors are used to get the pin-cell homogenized parameters. The heterogeneous leakage model is applied to modify the infinite flux spectrum of the single assembly with reflective boundary condition and to determine the diffusion coefficients for the SP3 solver which is used in the core simulation. To reduce the environment effect of the single-assembly reflective boundary condition, the online method for the SPH factors updating is applied in this paper, and the functionalization of SPH factors based on the least-squares method will be pre-made alone with the table of the group constants. The fitting function will be used to update the thermal-group SPH factors with a whole-core pin-by-pin homogeneous solution online. The three-dimensional Watts Bar Nuclear Unit 1 (WBN1) problem was utilized to test the performance of pin-by-pin calculation. And numerical results have demonstrated that PWR pin-by-pin core calculation has more accurate results compared with the traditional assembly-homogenization scheme.

• Tribology and Lubricants
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• v.18 no.2
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• pp.105-108
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• 2002

Specially designed grooved journal bearings are installed in the main coolant pump for SMART (System-integrated Modular Advanced ReacTor) to support radial load on the rotating shaft. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and filled with circulating primary coolant which is pure water. The main coolant pump bearings are lubricated with this coolant without any other external lubricant supply. Because lubricating condition is too severe for this bearing to generate proper hydrodynamic film, investigation of lubrication characteristics of the journal bearing is important to satisfy life constraint of whole pump system, and the results will be applied to the analysis of dynamic characteristics of the shaft system. The bearing is made of silicon graphite which has self$.$lubricating effect. A lubrication analysis method is proposed for this vertically grooved journal bearing in the main coolant pump of SMART, and lubricational characteristics of the bearings are examined in this paper.

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