Energy Technology Data Exchange (ETDEWEB)

Since January, 1981, the project of development of nuclear fuel fabrication technology for Wolsung reactor (CANDU type) was undertaken by KAERI(Korea Advanced Energy Research Institute) and successfully fulfilled with loading 24 fuel bundles made by KAERI in Wolsung reactor in September, 1984. On the basis of this accumulated technology and experience, mass production plan to supply all the nuclear fuels for Wolsung reactor is under way. In this presentation, the Korean experience in the development of the nuclear fuel fabrication technology, safety and performance evaluation of KAERI fuel and the results of irradiation of KAERI fuels in Wolsung reactor will be described.

International Nuclear Information System (INIS)

The agreement between the government of Republic Of Korea (ROK) and the IAEA for the application of safeguards was signed and entered into force in 1975. The Additional Protocol (AP) to the Safeguards Agreement between the ROK and the IAEA was signed in 1999 and entered into force on 19 February 2004. Also, Implementation of Integrated Safeguards (IS) was started on 1 July, 2008 after a draw of the broader conclusion. The IAEA provides 2 hours notification for Random Interim Inspection (RII) under IS. For RII, the facility has to prepare the inspection documents in a short time. Therefore, KAERI (Korea Atomic Energy Research Institute) developed a new computerized nuclear material accounting system named KASIS (KAERI Safeguards Information treatment System) to treat the data by on-line for RII. For the efficient IS implementation, KAERI has a plan to upgrade the system to reflect the accounting approaches or reporting ...

Energy Technology Data Exchange (ETDEWEB)

Solving the maintenance and repair problems of electronic instruments, we have supported the research and development work, and reduced operation costs of the pilot plants in KAERI. In addition, we have improved the maintainability of instruments to use effectively. 18 tabs., 17 figs., 13 refs. (Author) .new.

International Nuclear Information System (INIS)

KAERI (Korea Advanced Energy Research Institute) has performed research to develop the fabrication technology of CANDU nuclear fuel since 1981. Based on the satisfactory results of in-pile and out-of-pile tests of prototype nuclear fuel and the outstanding performance of 48 KAERI-made nuclear fuels in Wolsung(CANDU) power reactor, Korean government decided KAERI to supply all the nuclear fuels for Wolsung from 1988. In order to guarantee the safety and performance of nuclear fuel manufactured in mass production scale, well-organized quality assurance system and appropriate quality control techniques should be established. To establish the QA system, KAERI reviewed various QA standards and decided to establish QA system based on the 10 CFR 50 Appendix B. Quality control techniques was also revised to fit the mass production even though quality inspection techniques have already been developed during ...

International Nuclear Information System (INIS)

Korea Advanced Energy Research Institute (KAERI) started a research project to develop the PHWR (CANDU) nuclear fuel fabrication technology in 1981. Based on the results of the intensive developmental work, several prototype fuel bundles were fabricated and tested in the Hot Test Loop at KAERI continuously in 1983 and 1984. After that, irradiation test and post-irrradiation examination were carried out for two KAERI-made fuel bundles at Chalk River Nuclear Laboratories in Canada in 1984. Since the results of in-pile and out-of-pile tests with prototype fuel bundles proved to be satisfactory, 48 additional fuel bundles were loaded in Wolsung reactor (CANDU) in 1984 and 1985, and all of them were discharged without a defect after excellent performance in the power reactor. In 1985, the Korean government decided that KAERI supplies all the fuel necessary for the Wolsung reactor. For the mass production of ...

International Nuclear Information System (INIS)

KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

Energy Technology Data Exchange (ETDEWEB)

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A ...

International Nuclear Information System (INIS)

A safety analysis code to design a pressurized water reactor and to obtain the licenses including entire proprietary rights is under development in domestic R and D project. The tasks of KAERI is to develop the constitutive relations including models for defining flow regimes and flow regime related models for inter-phase friction, wall frictions, wall heat transfer, and interphase heat and mass transfer in the two-phase three-field equations. In this paper, the process will be presented for choosing the best flow regime maps which occur in gas-liquid two-phase flow in horizontal and vertical tubes.

Energy Technology Data Exchange (ETDEWEB)

A safety analysis code, named as SPACE, for a pressurized water reactor is under development to obtain a licensing to be used for the PWR design and to hold entire proprietary rights. The task of KAERI is to develop the physical models and correlations which are required to solve the field equations. It can be divided into four parts; i) flow regime determination, ii) wall heat transfer, iii) wall and interfacial friction, iv) interfacial heat and mass transfer. This paper will describe the process to develop the models for the two-phase flow regime maps in the horizontal and vertical pipes.

International Nuclear Information System (INIS)

PEFP(Proton Engineering Frontier Project) is developing beamlines for various application fields by using the high current 100MeV proton linear accelerator. Beamlines were characterized and categorized by the beam energy, average currents, and irradiation conditions. Among these beamlines the TR104 is designed for the basic science and the research of the aerospace. Before the construction and installation of beamlines, we try to practice and prepare the detail process of installation by the precedent study of TR104 beamline in KAERI site. By this work installation process was checked and confirmed for the same condition of the PEFP site

Energy Technology Data Exchange (ETDEWEB)

The Porous media approach method was applied to the MULTID component of MARS which was developed by KAERI. This method has been widely used for analysis of large regions containing only fluid and for flows in regions with immersed solids. It can calculate flow distribution more correctly in case of flow area or volume change. Application of this method to the MULTID component in MARS was evaluated and the results were compared with those of CFD code, FLUENT. As a result, the calculated flow distributions of MARS and FLUENT show similar trend, so the effect of porous media in MARS can be applied to the other cases.

International Nuclear Information System (INIS)

The purposes of this report are (1) to perform a Verification and Validation (V and V) test for the VIPEX(Vital-area Identification Package EXpert) software and (2) to improve a software quality through the V and V test. The VIPEX was developed in Korea Atomic Energy Research Institute (KAERI) for the Vital Area Identification (VAI) of nuclear power plants. The version of the VIPEX which was distributed is 3.2.0.0. The VIPEX was revised based on the first V and V test and the second V and V test was performed. We have performed the following tasks for the V and V test on Windows XP and VISTA operating systems: ? Testing basic functions including fault tree editing ? Testing all kind of functions ? Research for update from Visual BASIC 6.0 to Visual BASIC 2008

Energy Technology Data Exchange (ETDEWEB)

Most of previous study for creep rupture of RPV lower head under severe accident condition, have been focused on global failure of RPV lower head. In contract, the local failure of the RPV nozzle region has not been studied in detail. The existence and features of nozzle failure in LAVA-ICI specimen of KAERI and LHF-4 specimen of Sandia National Lab., are observed. It is confirmed that the nozzle failure of LHF-4 specimen is due to the hoop stress in the RPV. The tensile tests in various temperatures and the creep rupture tests in various temperatures and stresses, are accomplished. The finite element analysis for LAVA-ICI experiment was confirmed, and the stress and deformation analysis results are used in LAVA-ICI experiment. 17 refs., 34 figs., 3 tabs. (Author)

Energy Technology Data Exchange (ETDEWEB)

Severe accident analysis has been performed for the Wolsung nuclear power plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks (LOAH) which is caused by loss of off-site power, diesel generators, and DC power, ISAAC(Integrated Severe Accident Analysis Code) computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanisms of calandria vessel and containment. In addition, some insights for accident management program (AMP) are also given.

International Nuclear Information System (INIS)

Nuclear power plants of 20 units of in Korea are generating about 700 MTU of spent fuels annually. The inventory of spent fuels in Korea were estimated about 10,087.07 MTU at end of 2008, and the storage space of spent fuels won't be available any more at 2016 due to the saturation of the spent fuel pools in the plants. In addition, in order to reduce carbon emission and correspond to the enormous electricity demand in Korea, 8 units of nuclear power plants are under construction and several more plants are under planning. The 100,000 MTU of spent fuel inventory are expected by the year of 2095 in Korea. Therefore, short term and long term of spent fuel management plans are under discussion and implementation in Korea. As a short term of spent fuel management strategy for the target year of 2016, central or local spent fuel dry interim storage options are mostly under discussion. As a long term of management plan, fast reactor and advanced fuel cycle R and D plan were approved by 255th ...

International Nuclear Information System (INIS)

SMART (System-Integrated Modular Advanced Reactor) is a first-of-the-kind integral reactor with 330 MW thermal power under active development by Korea Atomic Energy Research Institute (KAERI) for power generation and seawater desalination. SMART employs various design features that are not typically found in other nuclear power plants. Examples include a unique passive residual heat removal system (PRHRS), and enclosure of a pressurizer, eight helical steam generators, and eight canned reactor coolant pumps inside the reactor pressure vessel. This paper presents risk insights on the SMART reactor gained during the development of a regulatory PSA model by Korea Institute of Nuclear Safety (KINS)

International Nuclear Information System (INIS)

Zirconium alloys have anisotropic mechanical properties depending on their physical orientations and are widely used as nuclear materials such as cladding tube material. An operation condition of the nuclear reactor requires a high creep resistance, because it is subjected to long period operations, high temperature and high pressure. Generally, it takes a few days or months to do the creep experiment, so it is difficult to get a data in short period. However, there is a way to predict a creep property by using the stress-relaxation in the short term. These studies realized the stress-relaxation through a compressive test of HANA-6 (Zr-1.1Nb-0.05Cu) alloy that was developed by KAERI (Korea Atomic Energy Research Institute), and then predicted the creep property

Energy Technology Data Exchange (ETDEWEB)

To assure quality of safety critical software, software should be developed in accordance with software development procedures and rigorous software verification and validation should be performed. Software verification is the formal act of reviewing, testing of checking, and documenting whether software components comply with the specified requirements for a particular stage of the development phase[1]. New software verification methodology was developed and was applied to the Shutdown System No. 1 and 2 (SDS1,2) for Wolsung 2,3 and 4 nuclear power plants by Korea Atomic Energy Research Institute(KAERI) and Atomic Energy of Canada Limited(AECL) in order to satisfy new regulation requirements of Atomic Energy Control Boars(AECB). Software verification methodology applied to SDS1 for Wolsung 2,3 and 4 project will be described in this paper. Some errors were found by this methodology during the software development for SDS1 and were corrected by software designer. ...

Energy Technology Data Exchange (ETDEWEB)

The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, {sup 14}C, Kr, Xe, I and {sup 3}H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and {sup 14}C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion ...

International Nuclear Information System (INIS)

Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This models and correlations manual provides a complete list of detailed information of the thermal-hydraulic models used in MARS, so that this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the ...

Energy Technology Data Exchange (ETDEWEB)

This study was carried out to evaluate the biochemical and morphological effects of ionizing radiation on ovary. Immature mice (ICR, 3 week-old) were irradiated at a dose of LD{sub 80(30)} at KAERI. The ovaries were collected after 6 hours, 12 hours, 1 day, and 2 days post irradiation. To analyze the morphological changes, histological staining with hematoxylin-eosin, immuno- histochemical preparation using in situ 3'-end labeling was performed. DNA fragmentation analysis and flowcytometric evaluation of DNA extracted from whole ovary were performed. As a result of DNA fragmentation analysis, DNA fragments with 185, 370, and 555 base pairs were clearly shown at 6 hours post irradiation. The percentage of A{sub 0} cell cycle was significantly increased in the irradiated group than control. In situ 3'-end labeled follicles were increased at 6 hours post irradiation. The radiation-induced follicular atresia was taken place via an apoptotic ...

International Nuclear Information System (INIS)

Much research has been done to estimate the residual stress on a dissimilar metal weld. There are many methods to estimate the weld residual stress and FEM (Finite Element Method) is generally used due to the advantage of the parametric study. And the X-ray method and a Hole Drilling technique for an experimental method are also usually used. The aim of this paper is to develop the appropriate FEM model to estimate the residual stresses of the dissimilar overlay weld pipe. For this, firstly, the specimen of the dissimilar overlay weld pipe was manufactured. The SA 508 Gr3 nozzle, the SA 182 safe end and SA376 pipe were welded by the Alloy 182. And the overlay weld by the Alloy 52M was performed. The residual stress of this specimen was measured by using the Neutron Diffraction device in the HANARO (High-flux Advanced Neutron Application ReactOr) research reactor, KAERI (Korea Atomic Energy Research Institute). Secondly, FEM Model on the dissimilar overlay weld pipe ...

International Nuclear Information System (INIS)

Population distribution, atmospheric change, X/Q, characteristics of terrestrial ecosystem around Seoul site were surveyed. Environmental radiation and radioactivities such as gross#alpha#, gross#beta#, Cs-137, Sr-90 and H-3 of various environmental samples were analyzed. The values of environmental radiation dose tended to increase gradually in the light of the recent five years' results of environmental radiation monitoring around the nuclear power plants from 1980 to 1984, however, the changes were not significant. In addition, continuous assessment of environmental radiation monitoring on the roofs of main building and life science building at KAERI showed that the environmental radiation dose tended to increase a little during the night time. Judging from the above results, it is concluded that environmental contamination level by radioactive materials could be ignored in the case of radioisotope production or experiment using radioisotopes except the release ...

Energy Technology Data Exchange (ETDEWEB)

This R and D project is planed to supply domestic demands of Mo-99 through fission route, and consequently this project will be expected to rise up utilization of HANARO and KAERI's capability for marketing extension into domestic and oversea radiopharmaceutical market. HEU and LEU target types are decided and designed for fission Mo-99 production in domestic. Experimental study of target fabrication technology was performed and developed processing equipments. And conceptual design of target loading/unloading in/from HANARO device are performed. Tracer test of Mo-99 separation and purification process was performed, test results reach to Mo-99 recovery yield above 80% and decontamination factor above 1600. Combined Mo-99 separation and purification process was decided for hot test scheduled from next year, and performance test was performed. Conceptual design for modification of existing hot cell for fission Mo-99 production facility was performed and ...

International Nuclear Information System (INIS)

Korea Atomic Energy Research Institute(KAERI) launched a project to develop an integral reactor in 1996. The reactor called as System Integrated Modular Advanced Reactor(SMART) which is a kind of small modular reactors (SMRs). Since the early 1990s, there has been renewed interest in the development and application of small and medium sized integral reactors. 2009 assessment by the IAEA under its Innovative Nuclear Power Reactor and Fuel Cycle (INPRO) program concluded that there could be 96 SMRs in operation around the world by 2030 in its 'high' case, and 43 units in the 'low' case, none of them in the USA. The reason of the increased demand mostly comes from the fact that SMRs are thought to be more suitable for developing countries with small electrical grid capacity, insufficient infrastructure and limited investment capability than developed ones. However, it has disadvantage in the point of scale of economy. So, it should be compared the amount of this ...

Energy Technology Data Exchange (ETDEWEB)

The development of KNGR (Korean Next Generation Reactor) is now in progress. KAERI is developing KNGR which is a advanced active PWR (pressurized water reactor) and 1350 MW electric capacities and is by based on UCN(Ulchin) 3 and 4 nuclear power plant which is a Korean standard PWR. In this report, the PZR (pressurizer) and Related System Design Features for KNGR which include PZR volume, PPCS (pressurizer safety valve)were analyzed. First, the Design Parameters between KNGR compared to UCH 3 and 4 were compared, and second, advanced design features of KNGR compared to UCN 3 and 4 were analyzed. After the present analysis, it has been concluded that the safety margins for the PZR level and pressure of KNGR were more increased by the larger PZR volume than those of UCN 3 and 4, for PZR minimum water level at reactor/turbine trip and PZR maximum pressure at LOCV(loss of condenser vacuum) of KNGR were higher and lower, respectively than those of UCN 3 and 4. Also, it ...

Energy Technology Data Exchange (ETDEWEB)

Eddy current testing (ECT) method is widely used as the non-destructive evaluation (NDE) of the various kinds of material degradation occurring in nuclear power plants (NPPs) components including steam generator (SG) tubes. Detection of any defects or flaws in the steam generator tubes in the early stage is very important in maintenance of NPP for its primary role as the pressure boundary with thin wall thickness. Although the ECT technique provides lots of information for a SG management, it has a generic problem in its reliability due to a low ability in detect of small defects and some difficulty in a signal analysis. For the improvement of these shortcomings in conventional ECT, profile-MRPC (motorized rotating pancake coil) ECT technology was developed in KAERI. The key of this new technology is the development of new eddy current probes, designated as a diagnostic probe (D-probe). The D-probe is furnished with a simultaneous dual function of a defect ...

International Nuclear Information System (INIS)

The Supercritical Water Cooled Reactor(SCWR) operates in a pressure around 25MPa and temperature of 293#approx#510 .deg. C. In order to study the heat transfer behaviors and good comparisons between the various fluids, a heat transfer test loop(SPHINX) using CO_2 has been constructed in KAERI as a part of international research program, I-NERI. At a supercritical pressure, the heat transfer coefficient is much larger than that estimated from the Dittus-Boelter correlation for a relatively large flow rate with moderate wall heat flux conditions. This phenomenon was explained by the rapid variations of the physical properties near the wall with the temperature. On the contrary, the heat transfer becomes worse when the bulk fluid enthalpy is below the pseudo-critical enthalpy under a low flow rate with large heat flux conditions. This phenomenon is called 'deteriorated heat transfer', and which is explained as the modification of the shear stress distribution across ...

Science.gov (United States)

This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, ...

Energy Technology Data Exchange (ETDEWEB)

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-Based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-Informed In-Service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, ...

Energy Technology Data Exchange (ETDEWEB)

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-Based operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-Informed In-Service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, ...

Science.gov (United States)

An experiment of heat transfer to CO{sub 2}, which flows upward and downward in a circular tube with an inner diameter of 6.32 mm, was carried out with mass flux of 285-1200 kg/m{sup 2} s and heat flux of 30-170 kW/m{sup 2} at pressures of 7.75 and 8.12 MPa, respectively. The corresponding Reynolds number at the tube test section inlet ranges from 1.8 x 10{sup 4} to 3.8 x 10{sup 5}. The tube inner diameter corresponds to the equivalent hydraulic diameter of the fuel assembly sub-channel, which is being studied at KAERI. Among the tested correlations, the Bishop correlation predicted the experimental data most accurately, but only 66.9% of normal heat transfer data were predicted within {+-}30% error range. The Watts and Chou correlation, which is claimed to be valid for both the normal and deteriorated heat transfer regime, showed unsatisfactory performance. A significant decrease in Nusselt number was observed in the range of 10{sup -6}

Energy Technology Data Exchange (ETDEWEB)

The CANDU-type nuclear fuel localization project started in 1981, and mass-production system completed in 1987 through the pilot scale demonstration of fuel manufacturing. Since the completion of the mass-production system, about 24,000 fuel bundles (450 ton-U) had been delivered to Wolsung Nuclear Power Plant by the end of 1992, according to the fuel supply contracts with KEPCO. The superiority of KAERI-made nuclear fuel has been demonstrated by having achieved the highest utilization factor in the world in 1992. In 1993, as contracted, 4,824 fuel bundles well fabricated and delivered to Wolsung Nuclear Power Plant. The process improvement, quality control, safety management, safeguards of nuclear materials and various kinds of audits have also been performed in the course for fuel manufacturing. Especially in 1993, the difficulties of the reduction of participating work-force were overcome by improving the manufacturing techniques, and raising the efficiency of ...

International Nuclear Information System (INIS)

In this experiment, we measured the detection efficiency of gamma-scanning system of Irradiated Materials Examination Facility (IMEF) in Korea Atomic Energy Research Institute(KAERI) for the spent PWR fuel and activity known sources with high activity. We measured the absolute detection efficiency of gamma scanning system of IMEF to the "1"3"7Cs-source with 10.4 GBq and "6"0Co source with 25.9 GBq High-activity sources as standard sources, and a fuel burned in the KORI -1 reactor of the Kori Nuclear power plant. In analyzing, we used three peeks those were the 661.64 keV peak form "1"3"7 Cs 1173.23 and 1332.51 keV"6"0 Co sources, and 14 peaks of "1"3"4Cs and "1"5"4Eu on the spent fuel gamma spectrum which are in the energy range from 500 to 1,600 keV. We find second order equations for detection efficiency by using our experimental results. The equation show that the detection efficiency of gamma scanning system for 1 MeV gamma-ray is 2.82217x10"-"6 % for the C3 ...

Energy Technology Data Exchange (ETDEWEB)

Gamma scanning measurements are used to determine the radioactive nuclides from an irradiated fuel rod in a hot cell which is a routine task in the Irradiated Materials Examination Facility (IMEF) of the Korea Atomic Energy Research Institute (KAERI), Total gamma spectra of radioactive nuclides and the atomic ratio of {sup 134}Cs to {sup 137}Cs in a few segments of a single fuel rod are measured by the present gamma scanning system. For the purpose of determining the radioactive nuclide's distribution in detail within a fuel rod quantitatively, we attempt to upgrade the present gamma scanning system in this paper, therefore, an investigation of the feasibility of the tomographic gamma scanning (TGS) technique, which is one of the promising nondestructive assay (NDA) methods is proposed. The TGS technique is an efficient method and it has an accurate precision for the characterization of several fuel parameters. In this paper, a principle and an emission ...

International Nuclear Information System (INIS)

The characteristics of per-fluorocarbon compounds (PFC) are colorless, non-toxic, easily vaporized and nonflammable. Also, some of them are liquids of a high density, low surface tension, low latent heat and low specific heat. These particular chemical and physical properties of fluoro-organic compounds permit their use in very different fields such as electronics, medicine, tribology, nuclear and material science. The Sonatol process was developed under a contract with the DOE. The Sonatol process uses an ultrasonic agitation in a PFC solution that contains a fluorinated surfactant to remove radioactive particles from surfaces. Filtering the suspended particles allows the solutions to be reused indefinitely. They applied the Sonatol process to the decontamination of a heterogeneous legacy Pu-238 waste that exhibited an excessive hydrogen gas generation, which prevents a transportation of such a waste to a Waste Isolation Pilot Plant. Korea Atomic Energy Research Institute ...