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APR+ has been developed in Korea since 2007. APR+ adopts various advanced safety features including passive auxiliary feedwater system, four emergency diesel generators. Through the implementation of the advanced designs, APR+ increased the safety to the world best level of evolutionary reactors. The full power core damage frequency or containment failure frequency decreased significantly comparing to APR1400 that is base model of APR+. However, low-power shutdown risk has not been improved substantially. This paper suggests several design changes that optimize low-power shutdown risk. Based on the design alternatives, this paper discusses risk effectiveness of the proposed design including various factors, e.g. equipment reliability, human error, training, procedure and so on.

Since the Fukushima nuclear accident, various efforts to improve safety have been undertaken for nuclear power plants all over the world, such as the implementation of new measures to cope with severe accidents and the application of PSAs in SAMG. In the case of the domestic probabilistic safety assessment (PSA), the regulatory body is strengthening the safety goals of operating and constructing nuclear power plants through the accident management plan, including the safety objectives recommended by the USNRC. This new safety goal calls for a more realistic reflection of the “as-built, as-operated” status of nuclear power plants as well safety enhancement of nuclear power plants during operation. In this paper, we propose a practical methodology for developing a Level 1 & 2 PSA model that reflects the inherent characteristics of Korea and conforms to the new safety goals recommended by domestic regulatory agencies. To this end, we benchmarked a recent State of the Art Reactor Consequence Analysis (SOARCA) study in the United States. The selection of scenarios reflecting the inherent characteristics of Korea, the application strategies for severe accident mitigation equipment and SAMG to PSA model and revision of the Level 1 & 2 PSA model for selected scenarios are presented. The results of this study can be applied to the PSA model update conforming to the new safety goals, the revision of the Level 1 & 2 PSA reflecting the unique characteristics, and analysis of the actual radiation source term and Public Acceptance.

In nuclear power plants (NPPs), one of the important yet uncertain phenomena during a station blackout (SBO) accident is the possibility for a counter-current natural circulation flow in the reactor coolant system (RCS) pressure boundary, which could divert the core melt progression and/or the in-vessel fission product (FP) distribution and, in turn, lead to a heatup and eventual creep rupture. From the risk and accident management points of view, one of key concerns is to predict more explicitly whether such a creep rupture can occur in the RCS pressure boundary, and if it indeed can, also predict the most probable time and the most likely location. This paper presents plant-specific analysis results on the potential characteristics of natural circulation-induced creep rupture of the RCS pressure boundary that could potentially take place during an SBO accident of a reference pressurized water reactor (APR1400). MELCOR-CFD (Computational Fluid Dynamics) coupled analysis was performed to determine more precisely the relevant physical parameter that could greatly influence on the behaviour of natural circulation and the ensuing creep rupture. Multiple sensitivity analyses were also carried out to investigate the influence that particular MELCOR modeling schemes and plant conditions have in estimating the RCS creep rupture characteristics. The analysis results and relevant insights were summarized in terms of particular points of interest.

Interest is increasing in the radiological consequences of a release of aerosol and gaseous iodine, especially after the Fukushima accident and also because of new interpretations of the results of recent severe accident experiments. This work provides a brief review of the history of iodine chemistry in containment and suggests an approach to include gaseous iodine, namely in the forms of elemental iodine and organic iodide, in consequence analyses using the MACCS code. As dry deposition is an important characteristic to distinguish each chemical form of iodine when performing a consequence analysis, the mechanisms and mathematical formulas expressing dry deposition are also investigated. The proposed approach is demonstrated by performing consequence analyses with a unit release of 131I, with the resulting trends of concentration and dose for the different chemical forms of iodine presented and discussed. For the same amount of iodine release, there is a higher surface deposition of elemental iodine (I2) because it has a higher dry deposition velocity, while the air concentration of a representative organic iodide (CH3I) is higher due to its lower dry deposition velocity, which means a lower depletion of the air concentration. Despite elemental iodine having a lower air concentration, its higher dose coefficients for the inhalation pathway compensates for this when calculating doses. Further, inhaled doses increase when considering resuspension inhalation for extended durations of exposure. The approach proposed in this study is expected to be used flexibly to perform consequence analyses incorporating both aerosol and gaseous forms of iodine.

For interfacing system loss of coolant accident (ISLOCA) and multiple steam generator tube rupture (MSGTR) sequences, which are reactor building (RB) bypass accidents in CANDU-6 type reactors, off-site consequence analyses were performed based on an optimization method for the WinMACCS input model that reflects CANDU-6 nuclear power plant design-specific and nearby site-specific features, with related sensitivity analyses performed for important variables. As a result of assessment, it was confirmed that there is exceedingly low early fatality risk in the unmitigated case from MSGTR and ISLOCA accidents. For cancer fatality risk, it was confirmed that the averaged risk from the RB bypass accidents in CANDU-6 plants sufficiently satisfies NSSC safety goal (1.0E−6/year). Sensitivity analyses found surface roughness to be the most influential parameter in the MSGTR accident, with other parameters confirmed to be insignificant considering uncertainty in the off-site consequence analysis. Through the sensitivity analyses, long-term improvement items related to surface roughness, dry deposition velocities, initial radioactive clouds, the domestic public protection measures model, and dose conversion factors were derived for future off-site consequence analyses. The proposed long-term improvements for the best-estimate analysis of off-site consequence are applicable to all similar CANDU plants and can provide valuable insights into Level 3 PSA.

A steam generator tube rupture (SGTR) is a type of containment bypass accidents that if core damage is not prevented during this accident progression, has the potential to lead to a significant release of fission products (FPs) to the environment through the faulted steam generator (SG) or other possible paths in the secondary system. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during an SGTR accident of a reference pressurized water reactor (PWR) plant, and also investigates the impact of dedicated accident mitigation actions on the results of interest through the established modeling approach. To consider the integral responses of the reactor, SGs, and containment to a severe accident initiated by SGTR, detailed modeling was performed for a broad spectrum of plant safety/mitigation systems and relevant phenomena influencing the evolution of the accident. As the modeling and simulation tools for the SGTR accident analysis, MELCOR 2.2 and MAAP5 were employed to compare the results of interest. The results of the analysis demonstrate that the release of FPs into the environment could be estimated more realistically as a result of the best practice modeling for the SGTR accident and also it could be greatly decreased through dedicated mitigation actions. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

An interfacing systems loss of coolant accident (ISLOCA) is a type of containment bypass accident that may lead to a significant release of fission products (FPs) to the environment. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during the ISLOCA of a reference pressurized water reactor (PWR) plant and investigates the influence of a dedicated accident mitigation action on the results of interest through the established modeling approach. In addition to this, the key plant features and phenomenological mechanisms that might have a significant influence on the magnitude and timing of FP releases to the environment during an ISLOCA were examined through a couple of sensitivity analyses. To consider the integral responses of the reactor, containment, and auxiliary building to a severe accident initiated by an ISLOCA, detailed modeling was performed for a broad spectrum of relevant phenomena, plant systems, and structures that promote the deposition and retention of FPs along the flow path to the environment. As the modeling and simulation tools for the ISLOCA analysis, MELCOR 2.2 and MAAP5.04 were employed to compare the results of interest. The results of the analysis shows that the present best practice modeling approach proposed here has the potential to significantly decrease the release of FPs into the environment in the event of an ISLOCA. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

In CANDU-6 type reactors, a multiple steam generator tube rupture (MSGTR) accident is characterized as a reactor building (RB) bypass scenario. Although the probability of an MSGTR accident in the CANDU-6 type Wolsong plants is very low, a direct release of radioactive nuclides to the environment can cause severe radiation exposure to residents around the plants. For this reason, a best-estimate analysis has been carried out based on the currently available best-practice knowledge in the Wolsong plants. The present study is divided into three steps. First, the existing parameter file of generic CANDU-6 plants for the MAAP-ISAAC 4.03 code was revised to incorporate its design-specific features for the Wolsong plants. A comparative analysis between the existing and modified parameter files showed that the progress of a severe accident was delayed or mitigated in the case of the modified parameter file. Secondly, variables related to the steam generator decontamination factors (DFs) were improved. It was confirmed that applying the steam generator DFs based on the ARTIST test significantly reduced the release fraction of each nuclide element to the environment. Finally, severe accident management guidance (SAMG) developed for the Wolsong plants was used to manage the progression of severe accidents and to limit the release of fission products during a MSGTR accident. A mitigation action of closing the main steam safety valves (MSSVs) after meeting a SAMG entry condition can significantly reduce the amount of released fission products compared with simply supplying feedwater to a broken steam generator (SG). To meet the domestic regulatory requirement for the release of Cs-137 to the environment (100 TBq), it was identified that an operator should supply feedwater to the broken SG within 1.5 h after meeting the SAMG entry conditions. The best-estimate analysis methodology and relevant accident management strategy for the mitigation of severe accidents, which have been developed through the study, can be applied to all similar CANDU plants and provide valuable insights into Level 2 probabilistic safety assessment and development of relevant SAMG.

An interfacing system LOCA (ISLOCA) is a bypass accident that results in the release of reactor coolant directly to the outside of a reactor building through a break in part of the low-pressure boundary of the reactor coolant system. Though the core damage frequency resulting from an ISLOCA scenario in the CANDU-6 type Wolsong nuclear power plant is very low, a direct release of radioactive nuclides to the environment could lead to severe radiation exposure among residents near the plant, and cannot be overlooked. For this reason, a quantitative evaluation of the amount of radioactive nuclides to be released to the environment was carried out by applying a best-estimate methodology to the MAAP-ISAAC code 1 , and a relevant accident management strategy for it was investigated in terms of a deterministic analysis. As part of this, an auxiliary building was additionally modeled to simulate deposition effects in the auxiliary building as well as pool scrubbing effects at the submerged break location following an ISLOCA. The decontamination effect by deposition and so on within the piping following an ISLOCA was evaluated. The analysis results showed that the decontamination factor for other nuclide groups excluding inert gas, Sb and Te2 nuclide groups was about 2.5. The analysis with consideration of all these decontamination effects was identified to yield a 78.4% reduction of Cs released to the environment compared with previous conservative analysis, which did not consider the decontamination effect within the ISLOCA piping and in the auxiliary building. In the case of the Wolsong CANDU plant, prior to external cooling water injection to the reactor vault (which is considered to be a Severe Accident Management Guidance (SAMG) accident management strategy), the amount of Cs-137 released to the environment exceeds the domestic regulatory requirement (100 TBq). In contrast, a sensitivity analysis in which an operator closes the test valve installed in a medium-pressure emergency core cooling system piping path and relieves the pressure of a reactor building using a containment filtered venting system (CFVS), met the aforementioned limitation for the amount of Cs-137 when an operator closes the test valve within 1 h following the accident and operates the CFVS properly (i.e., according to its operating procedure). This means that the appropriate accident management strategy for an ISLOCA scenario in the Wolsong CANDU plant can sufficiently prevent severe radiation exposure around the plant. Moreover, it is expected that these best-estimate methodologies developed from this study can be applied to all CANDU-6 plants.

For an offsite consequence analysis, MELCOR Accident Consequence Code System (MACCS) II code is widely used as a tool. In this study, the algorithm of web-based Off-Site Consequence Analysis Program (OSCAP) using the MACCS II code was developed for an integrated leak rate test (ILRT) interval extension and Level 3 probabilistic safety assessment (PSA), and verification and validation (V&V) of the program was performed. The main input data of the MACCS II code are meteorological data, population distribution data and source term data. However, it requires lots of time and efforts to generate the main input data for an offsite consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each other as a site. To reduce efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site in Korea to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. In case of the source term data, the program includes the release fraction of each source term category resulting from Modular Accident Analysis Program (MAAP) code analysis and the core inventory data from ORIGEN code analysis. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so a user could select the defaulted source term data of each plant without handling source term input data. For V&V of the web-based OSCAP, the following analyses were performed. First of all, an offsite consequence analysis for reference plants was carried out using only the MACCS II code, and using the web-based OSCAP based on both methodologies of NUREG-1493 and NEI interim report. Secondly, the results of the offsite consequence analysis were compared with each other. Lastly, a sensitivity analysis was accomplished for the main input data of the meteorological data and the population distribution data using the web-based OSCAP. It is considered that the program can remarkably reduce man-power on performing a periodic risk assessment according to extending ILRT intervals, and Level 3 PSA. Also, the program can be applied to making accident management plan (AMP) and quantitative health objective (QHO), and to the field of public acceptance (PA).

The objective of this research is to examine the probabilistic approach to evaluating turbine ejection frequency considering common-cause failure. This paper identifies basic turbine ejection mechanisms under high and low speeds and presents a detailed probabilistic methodology (fault tree) for assessing ejection frequency. The alpha factor methodology is applied to common-cause failure evaluations. The frequencies under different test schemes are compared and the propagation of uncertainty through the fault tree model is evaluated. The following conclusions were reached: (1) the turbine blade ejection frequency due to ductile failure under high speed is around 8.005E-7/yr; (2) if common-cause failure is considered, the frequency will be increased by 11% and 33% depending on the test scheme; and (3) if the parameter uncertainties are considered, the frequency is estimated to be in the range of 9.35E-7 to 1.13E 6, with 90% confidence.

A gas-filled capillary plasma source for potential use in conjunction with laser wakefield acceleration is developed, and its discharge characteristics are presented here. In the capillary, helium or hydrogen gas (~150 torr in pressure) is injected, and a pulsed high voltage (~25 kV) is applied for electrical discharge and plasma production. The discharge properties of the helium and hydrogen gases are compared. The plasma densities inside the capillary hole are measured with a Mach-Zehnder interferometer and are obtained separately by a simple analysis using the diffusion equation and measured current profiles.

A zero-dimensional global model simulator for atmospheric pressure oxygen plasmas has been developed. The simulation model considers the configurations similar to that of plasma needle device. The simulation results show that those species of O, , and have the highest density in sequence. Electrons dissipate most of their energy through the collisions with oxygen molecules. If the input power increases, the density of most species also increases as much as three-boy collision for the creation of ozone is weakened and hence the density of ozone decreases. The body to volume ratio also affects the plasma density.

In the development of 3D graphic systems for higher resolution and more realistic modeling and rendering, graphic memories also have been playing a critical role to offer the required high bandwidth. Currently, GDDR5 SDRAM's provide with 7Gbps per pin speed [1], reaching their physical limit originated from single-ended signaling nature: noise in reference voltage and power, and channel crosstalk. Especially, the channel crosstalk takes a dominating portion in 7Gbps timing budget, becoming the main barrier for further speed improvement. Although there has been research on crosstalk canceller in memory interface [2], it imposed stringent restrictions on signal ordering and trace length in PCB and package routing, and had limited performance. Therefore, improving the efficiency of DRAM core draws more attention than pin bandwidth now.

The long gap discharge in a plasma display panel has an advantage of high luminous efficacy but has a deficiency of high firing voltage as well. A hump-shape electrode is suggested to reduce the firing voltage of a long gap discharge while maintaining high luminous efficacy. The effects of hump electrodes inserted between two sustain electrodes are investigated by experiment and with a three-dimensional fluid simulation. The experimental results show that hump structure has lower firing and sustain voltage than the conventional long gap structure. A three-dimensional fluid simulation shows that the hump electrode enhances discharge ignition while maintaining the same optical characteristics of the long gap discharge. Also, the hump structure guarantees the addressing reliability with a reduced discharge time lag.

For this study, a database model of plant reliability was developed for the effective acquisition and management of plant-specific data that can be used in various applications of plant programs as well as in Probabilistic Safety Assessment (PSA). Through the development of a web-based reliability data analysis algorithm, this approach systematically gathers specific plant data such as component failure history, maintenance history, and shift diary. First, for the application of the developed algorithm, this study reestablished the raw data types, data deposition procedures and features of the Enterprise Resource Planning (ERP) system process. The component codes and system codes were standardized to make statistical analysis between different types of plants possible. This standardization contributes to the establishment of a flexible database model that allows the customization of reliability data for the various applications depending on component types and systems. In addition, this approach makes it possible for users to perform trend analyses and data comparisons for the significant plant components and systems. The validation of the algorithm is performed through a comparison of the importance measure value (Fussel–Vesely) of the mathematical calculation and that of the algorithm application. The development of a reliability database algorithm is one of the best approaches for providing systemic management of plant-specific reliability data with transparency and continuity. This proposed algorithm reinforces the relationships between raw data and application results so that it can provide a comprehensive database that offers everything from basic plant-related data to final customized data.

The electrons injected into a laser wakefield undergo betatron oscillation and give rise to the emission of intense X-ray radiation. To investigate the generation conditions of the X-rays, the relativistic motion of an electron injected in an off-axis position has been simulated with wakefield profiles which are pre-calculated with a two-dimensional particle-in-cell code. The wakefield with a plasma density of 1.78×10 18 ㎝ -3 is generated by the laser with an intensity of 1.37×10 18 W/㎠ and a pulse width of 30 fs. From the calculation of the single particle motion, the characteristics of the betatron radiation are investigated in the time domain. As the transverse injection position increases, the power and the duration time of the radiation increase, but the width of each pulse decreases.

The extended regulatory interests in severe accidents management and enhanced safety regulatory requirements raise a need of more accurate analysis of the effect to the public health by users with diverse disciplines. This facilitates this work to develop web-based radiation health effect assessment software, RASUM, by using the MACCS2 code and HTML language to provide diverse users (regulators, operators, and public) with easy understanding, modeling, calculating, analyzing, documenting and reporting of the radiation health effect under hypothetical severe accidents. The engine of the web-based RASUM uses the MACCS2 as a base code developed by NRC and is composed of five modules such as development module, PSA training module, output module, input data module (source term, population distribution, meteorological data, etc.), and MACCS2 run module. For verification and demonstration of the RASUM, the offsite consequence analysis using the RASUM frame is performed for such as early fatality risk organ does...

4 Gb/s/pin 32 bit 512 Mb GDDR4 (Graphics Double Data Rate 4) SDRAM was implemented by using an 80 nm CMOS process. It employs a data bus inversion (DBI) coding to overcome the bottleneck of a parallel single-ended signaling, a power consumption of I/O, power supply noise, and crosstalk. Both DBI AC and DC modes are combined to a single circuit by eliminating the feedback path of a conventional DBI AC circuit while achieving high-speed operation. The proposed DBI circuit uses an analog majority voter insensitive to mismatch for small area and delay. Ronmiddot tuning further improves the voltage and time margin by adding a user-supplied offset to auto-calibrated Ronmiddot. In addition, a dual duty cycle corrector (DCC) is used to reduce duty error and jitter by averaging two outputs of two DCCs. Measured results show that DBI DC coding reduces the peak-to-peak jitter from 65.5 ps to 44.5 ps and the voltage fluctuation from 183 mV to 115 mV at the data rate of 4 Gb/s with the 2 V.

An 8 M /spl times/ 32 GDDR (graphic DDR) SDRAM operating up to 800-MHz clock (CLK) frequency is described. The GDDR SDRAM demands an effective control of CAS latency due to the large and wide number of CAS latencies at the CLK frequency. A wave-pipelined CAS latency control circuit is proposed to provide stable operation for the large and wide number of CAS latencies. The increase of CAS latency also causes a degradation of data bus efficiency at high-speed operation due to the large gap between input data (DINs) and output data (DOUTs) at the operation of write followed by read. A gapless write to read scheme improves the data bus efficiency by separating write data-path from read data-path for different banks accesses. Partial array activation commands can reduce the peak current, preventing the reduction of the data retention time of DRAM cells at high-speed operation. The GDDR SDRAM operates successfully at the CLK frequency of 800 MHz at 2.1 V and 700 MHz at 1.8 V, respectively. The power consumption is measured to be /spl sim/2 W at 1.9 V.

Non-thermal atmospheric pressure plasma (NTAPP) is an ionized gas at room temperature and has potential as a new apoptosis-promoting cancer therapy that acts by generating reactive oxygen species (ROS). However, it is imperative to determine its selectivity and standardize the components and composition of NTAPP. Here, we designed an NTAPP-generating apparatus combined with a He gas feeding system and demonstrated its high selectivity toward p53-mutated cancer cells. We first determined the proper conditions for NTAPP exposure to selectively induce apoptosis in cancer cells. The apoptotic effect of NTAPP was greater for p53-mutated cancer cells; artificial p53 expression in p53-negative HT29 cells decreased the pro-apoptotic effect of NTAPP. We also examined extra- and intracellular ROS levels in NTAPP-treated cells to deduce the mechanism of NTAPP action. While NTAPP-mediated increases in extracellular nitric oxide (NO) did not affect cell viability, intracellular ROS increased under NTAPP exposure and induced apoptotic cell death. This effect was dose-dependently reduced following treatment with ROS scavengers. NTAPP induced apoptosis even in doxorubicin-resistant cancer cell lines, demonstrating the feasibility of NTAPP as a potent cancer therapy. Collectively, these results strongly support the potential of NTAPP as a selective anticancer treatment, especially for p53-mutated cancer cells.

Summary form only given. Although various experiments have been conducted routinely in many tokamak devices, the detailed physics background of so-called, start-up including processes of the plasma breakdown, burn-through, and current ramp-up has not been identified because too many parameters are involved together non-linearly and diagnostic tools are restricted in this phase. Plasma breakdown using electron cyclotron resonance heating (ECRH) has been proposed in the Korean superconducting tokamak advanced research (KSTAR) device because the loop voltage is limited due to thick vacuum liners and engineering limits of superconducting poloidal field coils. Particle-in-cell (PIC) simulation is a very effective tool to research the plasma phenomena but difficult to be applied to three dimensional, whole-size tokamak devices because of its large computation time. We developed a three-dimensional PIC code for investigating the breakdown of the tokamak start-up with ECRH. A simplified toroidal geometry is adopted for easy calculations with a prescribed magnetic field profile calculated for the KSTAR device. This PIC code includes Monte Carlo collision (MCC) routine for hydrogen atoms. In this work, we investigated the discharge characteristics and the effect of ECRH when the breakdown occurs in a tokamak. The influence of the magnitude of the toroidal magnetic field, loop voltage, gas pressure, and incident angle and intensity of ECRH is studied for breakdown in a tokamak.

In mobile systems, the demand for the energy saving continues to require a low power memory sub-system. During the last decade, the floating-gate flash memory has been an indispensable low power memory solution. However, NOR flash memory has begun to show difficulties in scaling due to the device's reliability and yield issues. Over the past few years, phase-change random access memory (PRAM) has emerged as an alternative non-volatile memory (NVM) owing to its promising scalability and low cost process [1,2]. In this paper, a PRAM, implemented in a 58nm PRAM process with a low power double-data-rate nonvolatile memory (LPDDR2-N) interface, is presented [3].

A high efficacy alternating current plasma display panel (PDP) utilizing vertically raised multilayer bus electrodes (MBE) is simulated by using a three-dimensional fluid code for the investigation of the effects of cell structures on the discharge characteristics of a PDP cell. The raised bus electrodes provide an opposite discharge mode with a low discharge current in a relatively long gap, and thus provide higher luminance and lower current simultaneously. The luminous efficacy of an MBE-PDP with a gap distance of 350 m was found to be twice higher than that of a conventional coplanar PDP structure having an electrode gap of 60 m. The fluid simulation coupled with visible ray-trace calculation shows the light-emission patterns that vary with the electrode structures, which explain the mechanism of the improvement of the luminous efficacy. Emission spectra and spatio-temporal distributions of infra-red radiation in a PDP cell discharge were measured experimentally and are compared with simulation results.

For operating frequencies exceeding 500 MHz, the timing margin of the I/O interface is critical and requires the data input-output timing accuracy to be within 200 ps. To meet the requirement, the designed SDRAM adopts a digitally self-calibrated on-die-termination with linearity error of /spl plusmn/1% and achieves over 1.2 Gbps/pin stable operation by using window matching and latency control. The chip is fabricated in a 0.13 /spl mu/m triple-well DRAM process.

A ray-trace model has been developed for the visible lights coming out from phosphors excited by vacuum ultraviolet (UV) lights in gaseous discharges. The spatial and angular light distributions are calculated numerically by considering multiple reflections of the visible lights within a plasma-display-panel (PDP) cell. Luminance of the emitting visible lights is illustrated for two different electrode shapes of an ac-type PDP cell. This calculation can be used for the improvement of optical properties of display devices utilizing phosphors and UV lights from plasma discharges, such as flat fluorescent lamps as well as PDP.