• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.263-275
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• 2004

For the safe management of nuclear facilities, including a radioactive waste repository, data about the facility site and the surrounding environment must be collected and managed systematically. This is particularly true for a radwaste repository, which has to be institutionally controlled for a long period after closure. The objectives of this study are (1) to establish a systematical management plan for information about a radwaste repository site and its environment, and (2) to design a database management program for this information, based on the Relative Database Management System (RDBMS). The spatial data are designed by the geodatabase, which is a new object, based on the RDBMS, to manage spatial information related to the database. To meet this requirement, a new program called 'Site Information and Total Environmental data management System (SITES)' is being developed. The scope that produced from the first step of the present study for development of the SITES is introduced. The database is designed to combine spatial and attribute data, and is designed for the establishment of the Geographic Information System (GIS). The hardware and software systems are designed with consideration given to the total data management of the items within the radioactive environment.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.99-104
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• 2011

An ILRT Interval for a nuclear power plant in Korea was extended from once in five years to once in ten years. Therefore, it is necessary to evaluate risk impact for ILRT interval extensions. In this paper, input data were generated for the reference plants, KSNP, using raw data such as meteorological data, population distribution data and source term data. And, using MACCS II code the risk impact assessment was performed based on the two methodologies of NUREG-1493 and NEI Interim Report. The risk impact derived from an ILRT interval extension was identified not to be significant. It is considered to apply this study and results to making an accident management plan and safety goal, and to the field of public acceptance.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.275-283
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• 2019

Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.605-610
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• 2005

A method to estimate the relative sensitivity of a self-powered rhodium detector for an upcoming cycle is developed by combining the rhodium depletion data from a nuclear design with the site measurement data. This method can be used both by nuclear power plant designers and by site staffs of Korean standard nuclear power plants for determining which rhodium detectors should be replaced during overhauls.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2445-2453
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• 2021

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutualcomparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21^(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.373-380
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• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

• Korean Journal of Computational Design and Engineering
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• v.12 no.1
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• pp.74-85
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• 2007

During the lifecycle of a nuclear power plant many organizations are involved in KOREA. Korea Plant Engineering Co. (KOPEC) participates in the design stage, Korea Hydraulic and Nuclear Power (KHNP) operates and manages all nuclear power plants in KOREA, Dusan Heavy Industries manufactures the main equipment, and a construction company constructs the plant. Even though each organization has a digital data management system inside and obtains a certain level of automation, data sharing among organizations is poor. KHNP gets drawing and technical specifications from KOPEC in the form of paper. It results in manual re-work of definition and there are potential errors in the process. A data warehouse based on a neutral model has been constructed in order to make an information bridge between design and O&M phases. GPM(generic product model), a data model from Hitachi, Japan is addressed and extended in this study. GPM has a similar architecture with ISO 15926 "life cycle data for process plant". The extension is oriented to nuclear power plants. This paper introduces some of implementation results: 1) 2D piping and instrument diagram (P&ID) and 3D CAD model exchanges and their visualization; 2) Interface between GPM-based data warehouse and KHNP ERP system.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1148-1155
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• 2020

This work presents two different methods for quantifying and propagating the uncertainty associated with fuel composition at end of life for cask criticality calculations. The first approach, the computational approach uses parametric uncertainty including those associated with nuclear data, fuel geometry, material composition, and plant operation to perform forward depletion on Monte-Carlo sampled inputs. These uncertainties are based on experimental and prior experience in criticality safety. The second approach, the data-driven approach relies on using radiochemcial assay data to derive code bias information. The code bias data is used to perturb the isotopic inventory in the data-driven approach. For both approaches, the uncertainty in k_eff for the cask is propagated by performing forward criticality calculations on sampled inputs using the distributions obtained from each approach. It is found that the data driven approach yielded a higher uncertainty than the computational approach by about 500 pcm. An exploration is also done to see if considering correlation between isotopes at end of life affects k_eff uncertainty, and the results demonstrate an effect of about 100 pcm.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.575-584
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• 2006

As the development of total management systems for sites along with site environmental information is becoming standard, the system known as the Site Information and Total Environmental database management System (SITES) has been developed over the last two years. The first result was a database management system for storing data obtained from facilities, and a site characterization in addition to an environmental assessment of a site. The SITES database is designed to be effective and practical for use with facility management and safety assessment in relation to Geographic Information Systems. SITES is a total management program, which includes its database, its data analysis system required for site characterization, a safety assessment modeling system and an environment monitoring system. It can contribute to the institutional management of the facility and to its safety reassessment. SITES is composed of two main modules: the SITES Database module (SDM) and the Monitoring & Assessment (M&A) module [1]. The M&A module is subdivided into two sub-modules: the Safety Assessment System (SAS) and the Site Environmental Monitoring System (SEMS). SAS controls the data (input and output) from the SITES DB for the site safety assessment, whereas SEMS controls the data obtained from the records of the measuring sensors and facilities. The on-line site and environmental monitoring data is managed in SEMS. The present paper introduces the procedure and function of the M&A modules.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.226-235
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• 1999

Applying digital network technology for advanced nuclear plant requires deterministic communication for tight safety requirements, timely and reliable data delivery for operation-critical and mission-critical characteristics of nuclear plant. Communication protocols, such as IEEE 802/4 Token Bus, IEEE 802/5 Token Ring, FDDI, and ARCnet, which have deterministic communication capability are partially applied to several nuclear power plants. Although digital communication technologies have many advantages, it is necessary to consider the noise immunity from electromagnetic interference (EMI), electrical interference, impulse noise, and heat noise before selecting specific digital network technology for nuclear plant. In this paper, we consider the token frame loss and data frame loss rate due to the link error event, frame size, and link data rate in different protocols, and evaluate the possibility of failure to meet the hard real-time requirement in nuclear plant.