• 한국응용과학기술학회지
• /
• 제22권2호
• /
• pp.191-199
• /
• 2005

Removable protective adhesives for automobiles were synthesized by an emulsion polymerization of monomers such as n-butyl acrylate (BA), n-butyl methacrylate (BMA), acrylonitrile (AN), acrylic acid (AA) and 2-hydroxyethyl methacrylate (2-HEMA), in which AA and 2-HEMA were functional monomers. Potassium persulfate (KPS) was used as an initiator and sodium lauryl sulfate (SLS) was used as an emulsifier, and polyvinyl alcohol (PVA) was used as a stabilizer. Emulsion polymerization was carried out in a semi-batch type reactor. Tensile strength, extension, peel strength, viscosity and solid content of the synthesized adhesives were tested. The optimum physical properties of the removable protective adhesives for automobiles were obtained with the composition of 0.43 mole BA, 0.57 mole AN, 0.21 mole BMA, 0.03 mole AA, and 0.03 mole 2-HEMA.

• Nuclear Engineering and Technology
• /
• 제30권1호
• /
• pp.58-74
• /
• 1998

A PC window-based computer code, CONPAS (CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the detailed supporting event tree (DSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes five distinct, but closely related modules. Its computational feasibility to real PSAs has been assessed through an application to the UCN 3&4 full scope Level 2 PSA. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: (1) systematic uncertainty analysis / importance analysis / sensitivity analysis, (2) table / graphical display & print, (3) employment of the recent Level 2 PSA technologies, and (4) highly effective user interface. The main purpose of this paper is to introduce the key features of CONPAS code and results of its feasibility study.

• Macromolecular Research
• /
• 제14권5호
• /
• pp.517-523
• /
• 2006

The effects of particle content, size and shape on the thermal conductivity (k) and adhesion properties of thermally conductive, pressure-sensitive adhesives (PSAs) were investigated. The matrix resins were thermally crosslinkable, 2-ethylhexyl acrylic polyol and ultraviolet (UV)-curable, random copolymer consisting of acrylic oligomer and various acrylates. We found that k increased with increasing diameter and particle aspect ratio, and was further enhanced due to the reduction of the interfacial thermal barrier when the coupling agent, which increases the adhesion between particles and the matrix resin, was used. On the other hand, adhesion properties such as peel strength and tack of the thermally crosslinkable resin decreased sharply with increasing particle content. However, for UV curable resin, increased particle addition inhibited the decrease in adhesion properties.

• Nuclear Engineering and Technology
• /
• 제37권3호
• /
• pp.207-220
• /
• 2005

Nuclear power plants (NPP) are complex socio-technological systems that rely on the success of both hardware and human components. Empirical studies of plant operating experience show that human errors are important contributors to accidents and incidents, and that organizational factors play an important role in creating contexts for human errors. Current probabilistic safety assessments (PSA) do not explicitly model the systematic contribution of organizational factors to safety. As some countries, like the United States, are moving towards increased use of risk information in the regulation and operation of nuclear facilities, PSA quality has been identified as an area for improvement. The modeling of human errors, and underlying organizational weaknesses at the root of these errors, are important sources of uncertainty in existing PSAs and areas of on-going research. This paper presents a review of research into the following questions: Is there evidence that organizational factors are important to NPP safety? How do organizations contribute to safety in NPP operations? And how can these organizational contributions be captured more explicitly in PSA? We present a few past incidents that illustrate the potential safety implications of organizational deficiencies, some mechanisms by which organizational factors contribute to NPP risk, and some of the methods proposed in the literature for performing root-cause analyses and including organizational factors in PSA.

• Archives of Pharmacal Research
• /
• 제24권6호
• /
• pp.578-583
• /
• 2001

To investigate the feasibility of developing a new tenoxicam plaster, the effects of vehicles and penetration enhancers on the in vitro permeation of tenoxicam from a pressure-sensititre adhesive (PSA) matrices across the dorsal hairless mouse skin were studied. Vehicles employed in this study were propylene glycol (PC)-oleyl alcohol (OAI), PG-oleic acid (OA), and diethylene glycol monoethyl ether (DCMI)-propylene glycol monolaurate (PCML) cosolvents with/without fatty acids. In this studys amines such as triethanolamine (TEA) and tromethamine (TM) were additionally used as a solubilized. Among PSAs used, $Duro-Tak^{\circledR}$87-2510 showed much higher release rate than either $Duro-Tak^{\circledR}$ 87-2100 or $Duro-Tak^{\circledR}$87-2196. The relatively high flux rate was obtained with the formulation of DCMI-PCML (40:60, v/v) with 3% OA and 5% TM, and the flux increased as a function of the dose;the initial flux up to 12 h was $4.98{\pm}1.38{\;}{\mu\textrm{g}}/{\textrm{cm}^2}/h$ at the tenoxicam dose of $50{\;} mg/70{\;}{\textrm{cm}^2}$. This flux was much higher than that of a commercial piroxicam patch ($Trast^{\circledR}$) ($1.24{\pm}0.73{\;}{\mu\textrm{g}}/$\textrm{cm}^2/hr$) with almost only one-third that of the commercial patch. Therefore, these observations indicated that these composition of tenoxicam plaster may be practically applicable.

• 한국안전학회지
• /
• 제34권5호
• /
• pp.159-166
• /
• 2019

The seismic PSA is to probabilistically estimate the potential damage that a large earthquake will cause to a nuclear power plant. It integrates the probabilistic seismic hazard analysis, seismic fragility analysis, and system analysis and is utilized to identify seismic vulnerability and improve seismic capacity of nuclear power plants. Recently, the seismic risk of domestic multi-unit nuclear power plant sites has been evaluated after the Great East Japan Earthquake and Gyeongju Earthquake in Korea. However, while the currently available methods for system analysis can derive basic required results of seismic PSA, they do not provide the detailed results required for the efficient improvement of seismic capacity. Therefore, for in-depth seismic risk evaluation, improved system analysis method for seismic PSA has become necessary. This study develops a system analysis method that is not only suitable for multi-unit seismic PSA but also provides risk information for the seismic capacity improvements. It will also contribute to the enhancement of the safety of nuclear power plants by identifying the seismic vulnerability using the detailed results of seismic PSA. In addition, this system analysis method can be applied to other external event PSAs, such as fire PSA and tsunami PSA, which require similar analysis.

• Nuclear Engineering and Technology
• /
• 제35권2호
• /
• pp.154-164
• /
• 2003

The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through Level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities.

• Nuclear Engineering and Technology
• /
• 제53권10호
• /
• pp.3247-3255
• /
• 2021

As Probabilistic Safety Assessments (PSAs) are thoroughly conducted for the Site Operating States (SOSs) for a single unit, multi-unit Probabilistic Safety Assessments (MUPSAs) are ongoing worldwide to address new technical challenges or issues. In South Korea, the determination of the site operating states for a single site requires a logical approach with reasonable assumptions due to the fact that there are 4-8 operating units for each site. This paper suggests a simulation model that gives a reasonable expectation of the site operation states using the Monte-Carlo method as a stochastic approach and deterministic aspects such as operational policies. Statistical hypothesis tests were conducted so that the reliance of the simulation results can be guaranteed. In this study, 7 units of the Kori site were analysed as a case study. The result shows that the fraction of full power for all 7 units is nearly 0.45. For situations when more than two units are not in operation, the highest fraction combination was obtained for Plant Operation State (POS) 8, which is the stage of inspection and repairment. By entering various site operation scenarios, the simulation model can be used for the analysis of other site operation states.

• Nuclear Engineering and Technology
• /
• 제53권2호
• /
• pp.386-392
• /
• 2021

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file. The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units. A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.

• Nuclear Engineering and Technology
• /
• 제54권9호
• /
• pp.3324-3335
• /
• 2022

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.