• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1916-1938
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• 2019

The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.207-216
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• 2014

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.10 s.253
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• pp.1261-1268
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• 2006

This paper reports low-cost microreactor $(10{\mu}{\ell})$ biochip for the DNA PCR (polymerase chain reaction). The microbiochip $(20mm{\times}28mm)$ is a hybrid type which is composed of PDMS (polydimethylsiloxane) layer with serpentine micochannel $(360{\mu}m{\times}100{\mu}m)$ chamber and glass substrate integrated with microheater and thermal microsensor. Undesirable bubble is usually created during sample loading to PMDS-based microchip because of hydrophobic chip surface. Created bubbles interrupt stable biochemical reaction. We designed improved microreactor chamber using microfluidic simulation. The designed reactor has a coner-rounded serpentine channel architecture, which enables stable injection into hydrophobic surface using micropipette only. Reactor temperature needed to PCR reaction is controlled within ${\pm}0.5^{\circ}C$ by PID controller of LabVIEW software. It is experimentally confirmed that SRY gene PCR by the fabricated microreactor chip is performed for less than 54 min.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.43-53
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• 2017

Increasing the operation cycle length can be an important goal in the fuel reload design of a nuclear reactor core. In this research paper, a new optimization approach, electromagnetism mechanism (EM), is applied to the fuel arrangement design of the Bushehr WWER-1000 core. For this purpose, a neutronic solver has been developed for calculating the required parameters during the reload cycle of the reactor. In this package, two modules have been linked, including PARCS v2.7 and WIMS-5B codes, integrated in a solver for using in the fuel arrangement optimization operation. The first results of the prepared package, along with the cycle for the original pattern of Bushehr WWER-1000, are compared and verified according to the Final Safety Analysis Report and then the results of exploited EM linked with Purdue Advanced Reactor Core Simulator (PARCS) and Winfrith Improved Multigroup Scheme (WIMS) codes are reported for the loading pattern optimization. Totally, the numerical results of our loading pattern optimization indicate the power of the EM for this problem and also show the effective improvement of desired parameters for the gained semi-optimized core pattern in comparison to the designer scheme.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.318-329
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• 2016

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed "Burnable absorber-Integrated Guide Thimble" (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.964-974
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• 2020

The small modular reactors (SMRs) of the integrated pressurized water reactor (IPWR) type have been widely developed owing to their enhanced safety features. The SMR-IPWR adopts passive residual heat removal system (PRHRS) to extract residual heat from the core. Because the PRHRS removes the residual heat using the latent heat of the water stored in the emergency cooldown tank, the PRHRS gradually loses its cooling capacity after the stored water is depleted. A quick restoration of the power supply is expected infeasible under station blackout accident condition, so an advanced PRHRS is needed to ensure an extended grace period. In this study, an advanced design is proposed to indirectly incorporate a dry air cooling tower to the PRHRS through an intermediate loop called indefinite PRHRS. The feasibility of the indefinite PRHRS was assessed through a long-term transient simulation using the MARS-KS code. The indefinite PRHRS is expected to remove the residual heat without depleting the stored water. The effect of the environmental temperature on the indefinite PRHRS was confirmed by parametric analysis using comparative simulations with different environmental temperatures.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3494-3515
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• 2022

This paper presents the development of a nodal diffusion code, RAST-V, and its verification and validation for VVER (vodo-vodyanoi energetichesky reactor) analysis. A VVER analytic solver has been implemented in an in-house nodal diffusion code, RAST-K. The new RAST-K version, RAST-V, uses the triangle-based polynomial expansion nodal method. The RAST-K code provides stand-alone and two-step computation modes for steady-state and transient calculations. An in-house lattice code (STREAM) with updated features for VVER analysis is also utilized in the two-step method for cross-section generation. To assess the calculation capability of the formulated analysis module, various verification and validation studies have been performed with Rostov-II, and X2 multicycles, Novovoronezh-4, and the Atomic Energy Research benchmarks. In comparing the multicycle operation, rod worth, and integrated temperature coefficients, RAST-V is found to agree with measurements with high accuracy which RMS differences of each cycle are within ±47 ppm in multicycle operations, and ±81 pcm of the rod worth of the X2 reactor. Transient calculations were also performed considering two different rod ejection scenarios. The accuracy of RAST-V was observed to be comparable to that of conventional nodal diffusion codes (DYN3D, BIPR8, and PARCS).

• 한국신재생에너지학회:학술대회논문집
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• 2011.05a
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• pp.156.1-156.1
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• 2011

It is of great importance to predict operating parameter characteristics of an integrated fuel processor by the increased life-time and system performance. In this study, computational analysis is performed to gain fundamental insights on transport phenomena and chemical reactions in reformer which consists of preheating, steam reforming, and water gas shift reaction beds. Also, a top-fired burner locates inside of the reforming system. The combustor is providing thermal energy necessary for the steam reforming bed which is a endothermic catalytic reactor. Two-dimensional numerical model of the integrated fuel processing system is introduced for the analysis of heat and mass transport phenomena as well as surface kinetics and catalytic process. A kinetic model was developed and then computational results were compared with the experimental data available in the literature. Subsequently, parameter study using the validated steam methane reforming model was conducted by considering operating parameters, i.e. steam to carbon ratio and temperature.

• Korean Journal of Poultry Science
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• v.12 no.2
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• pp.119-126
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• 1985

The purpose of the study is to review the evolutional process of the U.S. broiler industry into the integrated production and marketing systems and to examine the factors influenced such a structural change over the Past half a century. The study is also carried out under the consideration that the experience gained by the integrated U.S. broiler industry could be utilized in applying the system to any non-integrated broiler industries like one in Korea. The U.S. broiler industry has been prevailed in the typical independent operation of all stages, from factor supply through growing to dressing and marketing until such time as growers began to suffer from financial difficulties right after the end of the World Wat II due mainly to lowered price of broilers stemmed from whopping decrease of the demand, which eventually reflected to the allied agribusiness sectors. The feed business field was the first reactor to this situation so that a kind of measures was taken to cope with the problems faced in such a way to provide extended credit on feeds first and help obtain chicks or sell broilers later. This trend has been practiced for broiler related agribusiness through late 1940's to work together with growers by coordinating its function between two or more stages involved in broiler production and marketing process as a means of spreading risk and sharing profit between stages Participated. The integrated system and contract farming thus became Popular across the country through mid 1950's.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2007.06a
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• pp.39-40
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• 2007

The analysis of fault current limiting characteristics according to variation of fault current level in the integrated three-phase flux-lock type superconducting fault current limiter (SFCL), which consisted of three-phase flux-lock reactor wound on an iron core with the same turn's ratio between coil 1 and coil 2 for each single phase, was performed. To analyze the current limiting characteristics of this integrated three-phase flux-lock type SFCL, the short circuit experiments was carried out the various three-phase faults such as the single line-to-ground fault, the double line-to-ground fault, the triple line-to-ground fault. From the experimental results, the fault current limiting characteristic was improved according to increase of fault current level.