• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.68-79
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• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2638-2651
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• 2020

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.

• Corrosion Science and Technology
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• v.19 no.3
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• pp.138-145
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• 2020

Immersion tests of Alloy 600 were conducted in simulated primary water environments of a pressurized water reactor at 325 ℃ for 10, 100, and 1000 h to obtain insight into effects of surface deformation on internal and intergranular (IG) oxidation behavior through precise characterization using various microscopic equipment. Oxidized samples after immersion tests were covered with polyhedral and filamentous oxides. It was found that oxides were abundant in mechanically ground (MG) samples the most. The number density of surface oxides increased with time irrespective of the method of surface finish. IG oxidation occurred in mechanically polished (MP) and chemically polished (CP) samples with thin internal oxidation layers. However, IG oxidation was suppressed with relatively thick internal oxidation layers in MG samples compared to MP and CP samples, suggesting that MG treatment could increase resistance to primary water stress corrosion cracking (PWSCC) from the standpoint of IG oxidation. As a result, appropriate surface treatment for Alloy 600 could prevent oxygen diffusion into grain boundaries, inhibit IG oxidation, and finally induce its high PWSCC resistance.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.419-433
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• 2016

One of the important severe accident management measures in the Light Water Reactors is water injection to the reactor core. The related phenomena are investigated by performing experiments and computer simulations. One of the most widely known is the QUENCH test-program. A number of analyses on QUENCH tests have also been performed by different computer codes for code validation and improvements. Unfortunately, any deterministic computer simulation is not free from the uncertainties. To receive the realistic calculation results, the best estimate computer codes should be used for the calculation with combination of uncertainty and sensitivity analysis of calculation results. In this article, the QUENCH-03 and QUENCH-06 experiments are modelled using ASTEC and RELAP/SCDAPSIM codes. For the uncertainty and sensitivity analysis, SUSA3.5 and SUNSET tools were used. The article demonstrates that applying the best estimate approach, it is possible to develop basic QUENCH input deck and to develop the two sets of input parameters, covering maximal and minimal ranges of uncertainties. These allow simulating different (but with the same nature) tests, receiving calculation results with the evaluated range of uncertainties.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.27-36
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• 2020

Natural convection and thermal stratification phenomena are found in large water pools that are being used as heat sinks for decay heat removal from the reactor core using passive heat removal systems. In this study, the two-phase (water and air) natural convection and thermal stratification phenomena with conjugate heat transfer in the rectangular enclosure were investigated numerically using ANSYS Fluent 17.2 code. The transient numerical simulations of these phenomena in the full-scale computational domain of the experimental facility were performed. Generation of water vapour bubbles around the heater rod and evaporation phenomena were included in this numerical investigation. The results of numerical simulations are in good agreement with experimental measurements. This shows that the natural convection is formed in region above the heater rod and the water is thermally stratified in the region below the heater rod. The heat from higher region and from the heater rod is transferred to the lower region via conduction. The thermal stratification disappears and the water becomes well mixed, only after the water temperature reaches the saturation temperature and boiling starts. The developed modelling approach and obtained results provide guidelines for numerical investigations of thermal-hydraulic processes in the water pools for passive residual heat removal systems or spent nuclear fuel pools considering the concreate walls of the pool and main room above the pool.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.1
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• pp.21-34
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• 2021

The use of Field-Programmable Gate Arrays (FPGAs) in the development of safety-related Human-Machine Interface (HMI) systems has gained much momentum in nuclear applications. Recently, one of the application areas for the Advanced Power Reactor 1400 (APR1400) is in the development of the advanced Component Interface Module (CIM) of the Engineered Safety Features Actuation System (ESFAS). Using systems engineering approach, we have developed a new FPGA-based advanced CIM software. The first step of our software development process involves the Preliminary Hazard Analysis (PHA) based on the previous CIM design. In this paper, we describe the qualitative approach used in performing the preliminary hazard analysis. The paper presents the methodology for applying a modified Hazard and Operability (HAZOP) procedure for the conduct of PHA which resulted in a qualitative risk-ranking scheme that informed the decisions for the safety criteria in the requirements specification phase. The qualitative approach provided the justification for design changes during the advanced CIM software development process.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4022-4032
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• 2021

In this study, machine learning (ML) techniques were used to model surveillance test data of nuclear power plants from an international database of the ASTM E10.02 committee. Regression modeling was conducted using various techniques, including Cubist, XGBoost, and a support vector machine. The root mean square deviation of each ML model for the baseline dataset was less than that of the ASTM E900-15 nonlinear regression model. With respect to the interpolation, the ML methods provided excellent predictions with relatively few computations when applied to the given data range. The effect of the explanatory variables on the transition temperature shift (TTS) for the ML methods was analyzed, and the trends were slightly different from those for the ASTM E900-15 model. ML methods showed some weakness in the extrapolation of the fluence in comparison to the ASTM E900-15, while the Cubist method achieved an extrapolation to a certain extent. To achieve a more reliable prediction of the TTS, it was confirmed that advanced techniques should be considered for extrapolation when applying ML modeling.

• Corrosion Science and Technology
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• v.22 no.6
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• pp.457-465
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• 2023

Irradiation-assisted stress corrosion cracking (IASCC) is one of the main degradation mechanisms of austenitic stainless steels, which are used as reactor internal materials. Slow strain rate testing (SSRT) has been widely applied to evaluate the IASCC initiation characteristics of proton-irradiated tensile specimens. Tensile specimens require low surface roughness for micro-crack observation, and electropolishing is the most important specimen pre-treatment process used for this. In this study, optimal electropolishing conditions were examined through analyzing results of polarization experiments and surface roughness measurements after electropolishing. Corrosion cell and electropolishing equipment were fabricated for polarization tests and electropolishing experiments using SSRT specimens. The experimental parameters were electropolishing time, current density, electrolyte temperature, and stirring speed. The optimal electropolishing conditions for SSRT tensile specimens made of type 316 stainless steel were evaluated as a polishing time of 180 seconds, a current density of 0.15 A/cm², an electrolyte temperature of 60 ℃, and a stirring speed of 200 RPM.