• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.236-245
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• 2016

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.825-837
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• 2017

Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB) under strong short-period ground motions (SPGMs) and long-period ground motions (LPGMs). The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s) of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

• 대한전기학회논문지:시스템및제어부문D
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• 제52권8호
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• pp.452-461
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• 2003

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indicator system and its actual implementation in the existing nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indicator. The hysteresis of reed switches is one of the important factors in a repeat accuracy of control rod position indicator as well. This paper investigates efficiency of the magnetic flux concentrator and the hysteresis using FEM and verified differences in physicals characteristics by comparing the results of FEM and those of the experiment. As a result, it is shown that the characteristics of prototype control rod position indicator have a good agreement with the results of FEM.

• Nuclear Engineering and Technology
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• 제36권2호
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• pp.142-152
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• 2004

The liquid entrainment and vapor pull-through offtake model of RELAP5/MOD3 had been developed for SBLOCA (Small Break Loss of Coolant Accident). The RELAP5/MOD3 model for horizontal volumes accounts for the phase separation phenomena and computes the flux of mass and energy through a branch when stratified conditions occur in the horizontal pipe. In the case of CANDU reactor, this model should be used in the coolant flow of 95 feeders connected to the reactor header component under the horizontal stratification in header. The current RELAP5 model can treat the only 3 directions junctions; vertical upward, downward, and side oriented junctions, and thus improvements for the liquid entrainment and vapor pull-through model were needed for considering the exact angles. The RELAP5 off-take model was modified and generalized by considering the geometric effect of branching angles. Based on the previous experimental results, the critical height correlation was reconstructed by use of the branch line connection angle and validation analyses were also performed using SET. The new model can be applied to vertical upward, downward and angled branch, and the accuracy of the new correlations is more improved than that of RELAP5.

• Nuclear Engineering and Technology
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• 제21권1호
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• pp.32-39
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• 1989

원형 단면을 가진 축대칭형 토카막 핵융합로에 적합할 수 있는 최적의 이상적 자기유체역학 베타값과 운전조건의 결정 방법을 제시하였다. 토로이달 전류밀도 분포로 조정되는 운전조건을 변화시키면서 이상적 자기유체역학 안정성을 유지시킬 수 있는 베타값의 한계를 계산하였다. 토로이달 전류밀도 분포에는 실험적 관찰들로부터 얻은 실험식들이 사용되었고, 베타 값의 한계를 결정하기 위해 필요로 여러 식들이 이 실험식들로부터 유도되었다. 토로이달 전류밀도의 각각 다른 분포에 대해서 다양한 베타 한계값 분포들이 얻어졌다. 이상적 자기 유체 역학 불안정성들에 의해 제안받는 최대의 베타값을 토카막의 기하학적 변수와 안전인자에 의한 scaling law로 표현하였다.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.488-493
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• 2018

The Firebrick Resistance-Heated Energy Storage (FIRES) concept developed by the Massachusetts Institute of Technology aims to enhance profitability of the nuclear power industry in the next decades. Studies carried out at Massachusetts Institute of Technology already provide estimates of the potential revenue from FIRES system when it is applied to industrial heat supply, the likely first application. Here, we investigate the possibility of operating a power plant (PP) with a fluoride-salt-cooled high-temperature reactor and a closed Brayton cycle. This variant offers features such as enhanced nuclear safety as well as flexibility in design of the PP but also radically changes the way of operating the PP. This exploratory study provides estimates of the revenue generated by FIRES in addition to the nominal revenue of the stand-alone fluoride-salt-cooled high-temperature reactor, which are useful for defining an initial design. The electricity price data is based on the day-ahead markets of Germany/Austria and the United States (Iowa). The proposed method derives from the equation of revenue introduced in this study and involves simple computations using MatLab to compute the estimates. Results show variable economic potential depending on the host grid but stress a high profitability in both regions.

• 한국소음진동공학회논문집
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• 제26권6_spc호
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• pp.697-704
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• 2016

The second shutdown drive mechanism (SSDM) that is classified into seismic category I as an active mechanical equipment shall maintain the structural integrity and its designed inherent safety functions during and/or after normal operation, anticipated operational occurrences, accidents and seismic occurrences. Therefore, not only a structural integrity assessment through numerical analyses but also a qualification test by using the prototype SSDM shall be conducted to verify the adequacy of the SSDM design. This paper describes a sort of seismic qualification test of the prototype SSDM to demonstrate that the structural integrity and operability (functionality) of SSDM are maintained during and/or after seismic excitations. From the results, this paper shows that the SSDM satisfies all design requirements without any malfunctions during and after the seismic test.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.477-488
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• 2013

The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2775-2787
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• 2021

At present, most of the widely used system codes for nuclear safety analysis are one-dimensional, which cannot effectively simulate the flow field of the reactor core or other structures. This is true even for the system codes containing three-dimensional modules with limited three-dimensional simulation function such as RELAP-3D. In contrast, the computational fluid dynamics (CFD) codes excel at providing a detailed three-dimensional flow field of the reactor core or other components; however, the computational domain is relatively small and results in the very high computing resource consuming. Therefore, the development of coupling codes, which can make comprehensive use of the advantages of system and CFD codes, has become a research focus. In this paper, a review focus on the researches of coupled CFD and thermal-hydraulic system codes was carried out, which summarized the method of coupling, the data transfer processing between CFD and system codes, and the verification and validation (V&V) of coupled codes. Furthermore, a series of problems associated with the coupling procedure have been identified, which provide the general direction for the development and V&V efforts of coupled codes.