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Sample records for 69-group kaeri wims

  1. ZZ WIMKAL-88, 69-Group KAERI WIMS Library for Thermal Reactors

    1 - Description of program or function: Format: WIMS; Number of groups: 69 energy groups; Nuclides: The following nuclides are included: H-1, H-1 (in H2O), H-1 (in ZrH), H-2, H-2 (in D2O), He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C, C (in graphite), N-14, 0-16, F, Na, Al, Si, P-31, S-32, K, Ti, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Zr, Zircaloy-2, Nb-93, Mo, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd, In-113, In-115, Sn, Gd, Dy-164, Er-166, Er-167, Lu-176, Hf, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Au-197, Pb. The following fission products are included: Kr-83, Mo-95, Tc-99, Ru-101, Ru-103, Rh-103, Rh-105, Pd-105, Pd-108, Ag-109, Cd-113, In-115, I-127, I-135, Xe-131, Xe-135, Cs-133, Cs-134, Cs-135, Nd-143, Nd-145, Pm-147, Pm-148g, Pm-148m, Pm-149, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Eu-154, Eu-155, Gd-157, Pseudo F.P. The following actinides are included: Th-232, Pa-233, U-233, U-234, U-235, U-236, U-237, U-238, Np-237, Np-238, Np-239, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242g, Am-242m, Am-243, Cm-242, Cm-244. Origin: ENDF/B-V or IV and JENDL-2(Rev.1); and from ENDL-84, where data was not available in ENDF/B. Weighting spectrum: Within-group weighting fluxes were computed. Neutron cross section library for thermal reactor design analysis with 69 energy WIMS groups structure. 2 - Method of solution: The library was produced using NJOY and by processing nuclides from ENDF/B-V or IV and JENDL-2 (Rev.1); and from ENDL-84, where data was not available in ENDF/B. Transport cross sections were computed using P1 scattering matrix data. Within-group weighting fluxes for actinides were computed on an ultra-fine group basis for accurate intermediate resonance self-shielding. A more explicit representation was adapted for the fission-product chain. A more extensive representation of the actinide burnup chain including Th cycle was selected

  2. Testing WIMS-D4M cross sections and the ANL ENDF/B-V 69 group library. Results from global diffusion and Monte Carlo calculations compared with measurements in the Romanian 14-MW TRIGA reactor

    Bretscher, M.M.

    1993-12-31

    The WIMS-D4 code has been modified (WIMS-D4M) to produce microscopic isotopic cross sections in ISOTXS format for use in diffusion and transport calculations. Beginning with 69-group libraries based on ENDF/B-V data, numerous cell calculations have been made to prepare a set of broad group cross sections for use in diffusion calculations. Global calculations have been made for two control rod states of the Romanian steady state TRIGA reactor with 29 fresh HEU fuel clusters. Detailed Monte Carlo calculations also have been performed for the same reactor configurations using data based on ENDF/B-V. Results from these global calculations are compared with each other and with the measured excess reactivities. Although region-averaged macroscopic principal cross sections obtained from WIMS-D4M are in good agreement with the corresponding Monte Carlo values, problems exist with the high energy (E > 10 keV) microscopic hydrogen transport cross sections.

  3. The developments of CWIMS code and its 69-group library

    Because of the limitation of original WIMS 69-group library, the reaction cross-sections and scattering matrices in the epithermal energy ranges are only given for one temperature. According to the requirement of user and for wide applications, the suitable adjustments of WIMS library were done, and the new WIMS library---CWIMS library is temperature-dependent in the whole energy ranges. Meanwhile the WIMS/D4 code was modified according to the new WIMS format library. Some auxiliary codes of new version WIMS/D4--- CWIMS, such as CSCN-Select and Collapse the CWIMS library and W10T2-Change CWIMS library from BCD to binary or from binary to BCD format were designed. In order to demonstrate the reliability of the CWIMS library and CWIMS code, five thermal assemblies -- TRX-1 and 2, BAPL-1,2 and 3 were calculated by using the CWIMS code and its own library. The calculated results were compared with those of experiments and old WIMS library

  4. Evaluation of WIMS-D/4 nuclear data library used on TRIGA reactor calculation

    The 69 groups constants of H in ZrH, 166Er and 167Er generated by NJOY and GASKET codes are inserted into WIMS nuclear data library WIMS-CNDC and WIMS-NINT libraries used on RTIGA reactor calculation are obtained. In order to check WIMS-CNDC and WIMS-NINT libraries, the scattering cross-section is compared with that in WIMS-IJS library. The group constant, K∞ and temperature coefficient are calculated by using WIMS-CNDC, WIMS-NINT and WIMS-IJS. The results show the both libraries are suitable for calculation of TRIGA reactor

  5. WIMS-D4M user manual

    The Winfrith Improved Multigroup Scheme (WIMS) code has been used extensively throughout the world for power and research reactor lattice physics analysis. There are many WIMS versions currently in use. The D4 version selected by the RERTR program was originally developed in 1980). It was chosen for the accurate lattice physics capability and an unrestricted distribution privilege. The code and its 69-group library tape 166259 generated in Winfrith were obtained from the Oak Ridge National Laboratory Radiation Shielding Information Center (RSIC) in 1992. Since that time the RERTR program has added three important features. The first was the capability to generate up to 20 broad-group bumup-dependent macroscopic or microscopic ISOTXS cross sections for each composition of the unit cell, a new ENDF/B-V based nuclear data library, and a new Supercell option. As a result of these modifications and other minor ones, the code is now named WIMS-D4M. A supplementary reference guide can be obtained from the RSIC that contains detailed explanations of all user options, library contents, along with several sample problems. Primary applications of WIMS for research reactor modeling do not require an extensive knowledge of all WIMS user options. This user guide is primarily addressed to the needs of the research reactor community although the code can be used for most thermal reactor lattices. The guide is written based on the experience of the RERTR staff with WIMS-D4M and will discuss only the most needed options for research reactor analyses

  6. WIMS-D4M user manual

    Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Costescu, C.I. [Univ. of Illinois, Champaign, IL (United States)

    1995-07-01

    The Winfrith Improved Multigroup Scheme (WIMS) code has been used extensively throughout the world for power and research reactor lattice physics analysis. There are many WIMS versions currently in use. The D4 version selected by the RERTR program was originally developed in 1980). It was chosen for the accurate lattice physics capability and an unrestricted distribution privilege. The code and its 69-group library tape 166259 generated in Winfrith were obtained from the Oak Ridge National Laboratory Radiation Shielding Information Center (RSIC) in 1992. Since that time the RERTR program has added three important features. The first was the capability to generate up to 20 broad-group bumup-dependent macroscopic or microscopic ISOTXS cross sections for each composition of the unit cell, a new ENDF/B-V based nuclear data library, and a new Supercell option. As a result of these modifications and other minor ones, the code is now named WIMS-D4M. A supplementary reference guide can be obtained from the RSIC that contains detailed explanations of all user options, library contents, along with several sample problems. Primary applications of WIMS for research reactor modeling do not require an extensive knowledge of all WIMS user options. This user guide is primarily addressed to the needs of the research reactor community although the code can be used for most thermal reactor lattices. The guide is written based on the experience of the RERTR staff with WIMS-D4M and will discuss only the most needed options for research reactor analyses.

  7. Generation of 69-group cross section library based on JEF data for TRIGA reactor calculations and its validation by analyzing the benchmark lattices of thermal reactors - 095

    A new executable, identified as NJOY99.0 has been created to generate the 69-group cross-section library for the reactor lattice transport code WIMS. The new code incorporates modifications in the WIMSR module of NJOY to generate the 69-group library, which will be used for TRIGA reactor calculations. The basic evaluated nuclear data file JEF-2.2 was used to generate the 69-group cross-section library in WIMS format. The results for TRX-1, TRX-2, BAPL-1, BAPL-2, and BAPL-3 benchmarks obtained by using the generated 69-group cross-section library from JEF-2.2 were analyzed. The following integral parameters were considered for the validation of the 69-group library: finite medium effective multiplication factor (keff), Ratio of epithermal to thermal 238U captures (ρ28), Ratio of epithermal to thermal 235U fission (δ25), Ratio of 238U fission to 235U fission (δ28) and Ratio of 238U captures to 235U fissions (C*). The TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project (WLUP) Stage-I by Ravnik. The calculated results of the integral parameters of TRX and BAPL Benchmark Lattices obtained by using the new version of code WIMSD-5B were found to be in good agreement with the experimental values. Besides, The TRX and BAPL calculation results showed that JEF-2.2 is reliable for thermal reactor calculations and validated the 69-group library, which will be used for the neutronic calculation of the TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh. (authors)

  8. WIMS Library updating

    At the end of 1990 the WIMS Library Update Project (WLUP) has been initiated at the International Atomic Energy Agency. The project was organized as an international research project, coordinated at the J. Stefan Institute. Up to now, 22 laboratories from 19 countries joined the project. Phase 1 of the project, which included WIMS input optimization for five experimental benchmark lattices, has been completed. The work presented in this paper describes also the results of Phase 2 of the Project, in which the cross sections based on ENDF/B-IV evaluated nuclear data library have been processed. (author)

  9. WIMS-E

    This report describes the WIMS-E Scheme for Neutronics Calculations. This scheme was set up to extend the existing calculational facilities in the neutronics field so as to cover a wider range of requirements and to permit more rapid changes to meet future requirements. It consists of a set of computer programs. (U.K.)

  10. KAERI photonuclear library

    Chang, Jong Hwa; Lee, Young Ouk; Han, Yin Iu

    2000-03-01

    This report contains summary information and figures depicting the KAERI photonuclear data library that extends up to 140 MeV of incident photon. The library consists of 143 isotopes from C-12 to Bi-209, providing the photoabsorption cross section and the emission spectra for neutron, proton, deuteron, triton, alpha particles, and all residual nuclides in ENDF6 format. The contents of this report and ENDF-6 format data library are available at http://atom.kaeri.re.kr/.

  11. Safeguards Implementation at KAERI

    The main objective of the safeguards implementation activities is to assure that there are no diversions of declared nuclear material and/or no undeclared activity. The purpose of safeguards implementation activities is the assistance facility operators to meet the safeguards criteria set forth by the Atomic Energy Safety Acts and Regulations. In addition, the nuclear material and technology control team has acted as a contact point for domestic and international safeguards inspection activities and for the relevant safeguards cooperation. Domestic inspections were successfully carried out at the KAERI nuclear facilities pursuant to the domestic laws and regulations in parallel with the IAEA safeguards inspections. It is expected that safeguards work will be increased due to the pyro-related facilities such as PRIDE, ACPF and DUPIC, for which the IAEA is making an effort to establish safeguards approach. KAERI will actively cope with the plan of the NSSC by changing its domestic inspection regulations on the accounting and control of nuclear materials

  12. Safeguards Implementation at KAERI

    Jung, Juang; Lee, Sung Ho; Lee, Byung-Doo; Kim, Hyun-Sook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The main objective of the safeguards implementation activities is to assure that there are no diversions of declared nuclear material and/or no undeclared activity. The purpose of safeguards implementation activities is the assistance facility operators to meet the safeguards criteria set forth by the Atomic Energy Safety Acts and Regulations. In addition, the nuclear material and technology control team has acted as a contact point for domestic and international safeguards inspection activities and for the relevant safeguards cooperation. Domestic inspections were successfully carried out at the KAERI nuclear facilities pursuant to the domestic laws and regulations in parallel with the IAEA safeguards inspections. It is expected that safeguards work will be increased due to the pyro-related facilities such as PRIDE, ACPF and DUPIC, for which the IAEA is making an effort to establish safeguards approach. KAERI will actively cope with the plan of the NSSC by changing its domestic inspection regulations on the accounting and control of nuclear materials.

  13. Multigroup cross section library; WIMS library

    The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings

  14. WIM calibration and data quality management

    D P G de Wet

    2010-01-01

    Weigh-in-motion (WIM) scales are installed on various higher order roads in South Africa to provide traffic loading information for pavement design, strategic planning and law enforcement. Some WIM systems produce anomalies that cannot be satisfactorily explained even by highly experienced professionals. Much of the problem relates to the difficulty in determining the appropriate calibration factors to correct systematic measurement error for WIM systems and the inadequacy of data quality man...

  15. WIMS-CRNL: A user's manual for the Chalk River version of WIMS

    This report describes the preparation of the input for WIMS-CRNL, the Chalk River version of the WIMS lattice code. Also included are notes on the operation of the code, contents of the associated libraries, and the relation of WIMS-CRNL to other versions of the code

  16. Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code

    Our aim in this report is to compare of calculation results, obtained with the use of different libraries, which are in the variant of the WIMS7B code. We had the three libraries: the 1986 library is based on the UKNDL files, the two 1996 libraries are based on the JEF-2.2 files, the one having the 69 group approximation, the other having the 172 group approximation. We wanted also to have some acquaintance with the new option of WIMS-7B - CACTUS. The variant of WIMS-7B was placed at our disposal by the code authors for a temporal use for 9 months. It was natural to make at comparisons with analogous values of TVS-M, MCU, Apollo-2, Casmo-4, Conkemo, MCNP, HELIOS codes, where the other different libraries were used. In accordance with our aims the calculations of unprofiled and profiled assemblies of the VVER-1000 reactor have been carried out by the option CACTUS. This option provides calculations by the characteristics method. The calculation results have been compared with the K∞ values obtained by other codes in work. The conclusion from this analysis is such: the methodical parts of errors of these codes have nearly the same values. Spacing for Keff values can be explained of the library microsections differences mainly. Nevertheless, the more detailed analysis of the results obtained is required. In conclusion the calculation of a depletion of VVER-1000 cell has been carried out. The comparison of the dependency of the multiply factor from the depletion obtained by WIMS-7B with different libraries and by the TVS-M, MCU, HELIOS and WIMS-ABBN codes in work has been performed. (orig.)

  17. The '1981' WIMS nuclear data library

    The WIMS nuclear datra library currently used for all WIMS calculations at Winfrith is known as the '1981' library. It contains data in 69 energy groups for 126 nuclides including fuel, moderator and structural materials, fission products and miscellaneous detectors. This document gives a brief summary of the data, concentrating on 2200 m/sec cross sections, Maxwellian averages and resonance integrals, but also including other important data and references where possible. (author)

  18. Pyroprocessing technology development at KAERI

    Lee, Han Soo; Park, Geun Il; Kang, Kweon Ho; Hur, Jin Mok; Kim, Jeong Guk; Ahn, Do Hee; Cho, Yung Zun; Kim, Eung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-08-15

    Pyroprocessing technology was developed in the beginning for metal fuel treatment in the US in the 1960s. The conventional aqueous process, such as PUREX, is not appropriate for treating metal fuel. Pyroprocessing technology has advantages over the aqueous process: less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, compact equipment, etc. The addition of an oxide reduction process to the pyroprocessing metal fuel treatment enables handling of oxide spent fuel, which draws a potential option for the management of spent fuel from the PWR. In this context, KAERI has been developing pyroprocessing technology to handle the oxide spent fuel since the 1990s. This paper describes the current status of pyroprocessing technology development at KAERI from the head-end process to the waste treatment. A unit process with various scales has been tested to produce the design data associated with the scale up. A performance test of unit processes integration will be conducted at the PRIDE facility, which will be constructed by early 2012. The PRIDE facility incorporates the unit processes all together in a cell with an Ar environment. The purpose of PRIDE is to test the processes for unit process performance, operability by remote equipment, the integrity of the unit processes, process monitoring, Ar environment system operation, and safeguards related activities. The test of PRIDE will be promising for further pyroprocessing technology development

  19. WIM calibration and data quality management

    D P G de Wet

    2010-10-01

    Full Text Available Weigh-in-motion (WIM scales are installed on various higher order roads in South Africa to provide traffic loading information for pavement design, strategic planning and law enforcement. Some WIM systems produce anomalies that cannot be satisfactorily explained even by highly experienced professionals. Much of the problem relates to the difficulty in determining the appropriate calibration factors to correct systematic measurement error for WIM systems and the inadequacy of data quality management methods. The author has developed a post-calibration method for WIM data, called the Truck Tractor (TT method, to correct the magnitude of recorded axle loads in retrospect. In addition, it incorporates a series of data quality checks. The TT method is robust, accurate and adequately simple for use on a routine basis for a wide variety of South African WIM systems. The calibration module of the TT method (i.e. the procedure to determine the calibration factor, kTT has been accepted by SANRAL and incorporated into the model it uses to quantify the cost of overloading on toll concessions. Some of the data quality checking concepts are also being considered for further use and threshold values for tests are being refined by SANRAL for this purpose.

  20. XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections

    1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections

  1. WIMS nuclear data library and its updating

    This report gives a brief overview of the status of reactor physics computer code WIMS-D/4 and its library. It presents the details of WIMS-D/4 Library Update Project (WLUP), initiated by International Atomic Energy Agency (IAEA) with the goal of providing updated nuclear data library to the user of WIMS-D/4. The WLUP was planned to be executed in several stages. In this report the calculations performed for the first stage are presented. A number of benchmarks for light water and heavy water lattices proposed by IAEA have been analysed and the results have been compared with the average of experimental values, the IAEA reference values and the average of calculated results from different international laboratories. (author) 8 figs

  2. The WIMS-E module W-PONE

    This report describes the WIMS-E module W-PONE for adding P1 matrices to a WIMS-E interface. W-PONE has recently been rewritten in Fortran 77 and a number of improvements have been made. The report includes data input and the method of use within the WIMS-E integrated scheme. (author)

  3. Description of WIMS Library Update Project (WLUP)

    WIMS-D is one of the few reactor lattice codes that are in the public domain and therefore are available on non-commercial terms, for research and power nuclear reactor calculations. The main weakness of the WIMS-D package is its multi-group constants library, which is based on very old data. Relatively good performance of WIMS-D is attributed to a series of empirical adjustments to the multi-group data. However, the adjustments are not always justified by more accurate and recent experimental measurements. In view of the recently available new, or revised, evaluated nuclear data files it was felt that the performance of WIMS-D could be improved by updating its library. The WIMS-D Library Update Project (WLUP) was initiated in the early 1990's and finished in 2001. The International Atomic Energy Agency (IAEA) supported its co-ordination, but the project itself consisted of voluntary contributions from a large number of participants. In due course, several benchmarks for testing the library were identified and analyzed, the WIMSR module of the NJOY code system was upgraded, a detailed parametric study was performed to investigate the effects of various data processing input options on integral results and, the data processing methods for the main reactor materials were optimized. The final product, available on CD-ROM from NDS-IAEA includes: 69 and 172 group WIMSD libraries prepared from the selected evaluated data files, IAEA-TECDOC with detailed documentation, Processing inputs, Benchmark inputs and, the system of auxiliary codes developed under the project. (author)

  4. Current Status of Pyroprocessing Development at KAERI

    Hansoo Lee

    2013-01-01

    Full Text Available Pyroprocessing technology has been actively developed at Korea Atomic Energy Research Institute (KAERI to meet the necessity of addressing spent fuel management issue. This technology has advantages over aqueous process such as less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, and compact equipments. This paper describes the pyroprocessing technology development at KAERI from head-end process to waste treatment. The unit process with various scales has been tested to produce the design data associated with scale-up. Pyroprocess integrated inactive demonstration facility (PRIDE was constructed at KAERI and it began test operation in 2012. The purpose of PRIDE is to test the process regarding unit process performance, remote operation of equipments, integration of unit processes, scale-up of process, process monitoring, argon environment system operation, and safeguards-related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

  5. Decommissioning of nuclear research facilities at KAERI

    At the Korea Atomic Energy Research Institute (KAERI), two research reactors (KRR-1 and KRR-2) and one uranium conversion plant (UCP) are being decommissioned. The main reason of the decommissioning was the diminishing utilities; the start of a new research reactor, HANARO, and the higher conversion cost than that of international market for the UCP. Another reason of the decommissioning was prevention from spreading radioactive materials due to the deterioration of the facilities. Two separate projects have already been started and are carried out as planned. The KAERI selected several strategies, considering the small scale of the projects, the internal standards in KAERI, and the future prospects of the decommissioning projects in Korea. In this paper, the current status of the decommissioning including the waste management and the technology development will be explained

  6. A study for the KAERI research tunnel

    Major goal of the R and D on the KAERI Research Tunnel in 1997 are 1) concept development of the KAERI research tunnel and its major units 2) computer simulation of facilities 3) study on thermo-hydro mechanical coupling in the vicinity of a waste repository 4) effect of excavated distrubed zone. In addition supplementary site investigation to understand the distribution of stresses in the site was done along with long term monitoring of the water table. (author). 44 refs., 16 tabs., 36 figs

  7. Analysis of uranium dioxide and uranium metal lattices using different multi-group cross section sets in WIMS-D/4 format

    Thermal reactor design calculations are being performed in India using the WIMS/D-4 multi group cross section library, obtained in late 60's, reflecting the status of the basic nuclear data and processing technology then available. Significant improvements in basic evaluated data files such as ENDF/B-IV to VI and JEF data files etc. have been made in the past four decades and the multigroup libraries have been updated world over using improved and comprehensive nuclear data processing code systems. A few of such updated multigroup cross sections in WIMS/D-4 format are available from KAERI and NEA data bank sources. This paper presents the analysis of a set of enriched UO2 and U-metal uniform critical lattice experiments. These include TRX(4), BAPL (3) and B and W (17) lattice, 64 enriched UO2 lattices complied in NEACRP-U-190 report, 56 enriched UO2 lattices and 61 U-metal lattices which were used for validating the WIMKAL-1988 library. Calculated reaction rate values from the participants of WIMS library update project (WLUP) are available for TRX, BAPL lattices. Integral data measured in the lattices of TRX, BAPL, B and W and NEACRP compilations are available in the open literature. Different calculational methods like J± and Pij, and resonance interpolation schemes were examined in the theoretical analysis. Possible shortcomings of the WIMS-D/4 multigroup cross section library currently being used are also identified. (author)

  8. Generation of a WIMS-D/4 multigroup constants library based on the JENDL-3.2 nuclear data and its validation through some benchmark experiments analysis

    Rahman, M. [Institute of Nuclear Science and Technology, Savar, Dacca (Bangladesh); Takano, Hideki

    1996-11-01

    A new 69 group library of multigroup constants for the lattice code WIMS-D/4 has been generated with an improved resonance treatment, processing nuclear data from JENDL-3.2 by NJOY91.108. A parallel ENDF/B-VI based library has also been constructed for intercomparison of results. Benchmark calculations for a number of thermal reactor critical assemblies of both uranium and plutonium fuels have been performed with the code WIMS-D/4.1 with its three different libraries: the original WIMS library (NEA-0329/10) and the new ENDF/B-VI and JENDL-3.2 based libraries. The results calculated with both ENDF and JENDL based libraries show similar tendency and are found in better agreement with the experimental values. Benchmark parameters are further calculated with the comprehensive lattice code SRAC95. The results from SRAC95 and WIMS-D/4.1 (both with JENDL-3.2 based libraries) agrees well to each other as well as to the other previously published values. (author)

  9. WIMS-ANL 4.0, Deterministic Code System for Lattice Calculation

    1 - Description of program or function: The WIMS-ANL code is an extension of the Winfrith WIMS-D4 code for lattice cell computations. This code has been tailored to address some of the problem areas encountered in dealing with research reactor fuels, experiment, reflector and control regions. The SUPERCELL option eliminates some of the limitations of the traditional SPECTROX solution and supports the solution of more complex geometries with a more detailed spatial mesh and multiple resonance materials. The code generates both macroscopic and microscopic cross sections in the ISOTXS format with any selected number of energy groups. The user can specify which fission product isotopes are to be explicitly included in the microscopic burnup dependent ISOTXS library. Fission product library data can be generated for use with the MCNP code and burnup dependent applications. The cross section library data provided are based on ENDF/B version VI (69 group) and V (69 and 172 group) data. A revised 172 group library based on ENDF/B-VI is being generated with newer data and additional isotopes. This library will be made available at a later time. The code is variably dimensioned so that other group structures could be used. The source code and output format have been completely revised to reflect current coding practices and to permit display of the results on typical desk top monitors. The content of the output displayed is completely under the user's control. 2 - Methods:The methods of solution in WIMS-ANL remain unchanged from those used in the original WIMS-D4 code with the same resonance treatment and a choice of collision probability and DSN solutions for the simple lattice cell. The SUPERCELL option provides for the selection of supporting auxiliary cells that might represent the various different elements and varying spectra of the final SUPERCELL model. The resonance treatments where applicable are carried out in the auxiliary cells. These data are combined in the

  10. WIMS library update project: first stage extension

    This paper reports the results of nine structural lattices obtained through the WIMS-TRACA computer program. This work was performed by request of the managers of the WLU/IAEA project, for the extension of the first stage. These benchmark lattices include regular arrays with heavy water and data of the thorium cycle. Besides K∞ and Keff (employing the experimental buckling to account for the leakages) spectrum index and ratio at reaction rates are also determined for comparison with the experimental values. The input data for each lattice, are given in the appendix to help exploring possible differences in the results. (author). 4 refs, 1 fig, 11 tabs

  11. The neutron radiography programme at KAERI

    The first KAERI neutron radiography facility, which was installed at the research reactor KRR-2(2MW) in early 1980's to utilize for the inspection of the nuclear and non-nuclear objects, was closed at the end of 1995. As a continued programme, a new neutron radiography facility has been installed at HANARO with various upgrades. In this article, its design features, performance characteristics and utilization programme are outlined.

  12. The technological innovation case of the KAERI

    Choi, J. I. [Habat Univ., Daejeon (Korea, Republic of); Jang, S. K. [Sungkonghoe Univ., Seoul (Korea, Republic of); Hong, K. P. [Baekseok Univ., Chunan (Korea, Republic of); Lee, E. S. [National Fusion Research Institue, Daejeon (Korea, Republic of)

    2008-01-15

    The research aims to investigate what key success factors (KSFs) of technological innovation in KAERI are, and to suggest how these findings are utilized for KAERI. In order to achieve these goals we have employed case study based on in-depth interview and literature review. And there are two fields of research in KAERI: one is nuclear energy-related research, the other is non energy-related research. The former is 'nuclear fuel cladding tube' which is an industrial product and being regarded as catch-up (or imitative) mode of technological innovation: the latter is 'HemoHIM', herbal composition of health functional food, which is consumer goods and regarded as creative (or innovative) mode of technological innovation. We found some KSFs in these two research and development cases in KAERI: firstly, to train researcher to be a 'product champion' who can fill in the gap of 'death valley' between pure research and commercialization: secondly, to build researchers' competency in order to catch up advanced countries' technological competencies. Thirdly, to amend institutional rules and regulations for commercializing processes of R and D outcomes, notably 'R and D joint venture by Government Research Institute (GRI) and private sector' fourthly, to enhance the capabilities of external management for researchers' technological innovation competency. And finally, we recommend using successful R and D cases as educational materials when training young researchers for sharing old generations' experiences and tacit knowledge.

  13. Programmes associated with WIMS library tapes

    A description of, and details of data preparation for the programmes WIMLIB, WIMCON, WINGAL, WIMRES and WIMMIX are given. All these are KDF9 programs with the exception of WIMGAL which is available both on KDF9 tape writing program, and may be used either to write a completely new library tape or to produce a revised version of an existing WIMS library tape. WIMCON takes an existing library tape together with appropriate condensation spectra in order to produce a condensed library tape in fewer energy groups. WIMGAL, WIMRES and WIMMIX are programs for preparing data for WIMLIB from GALAXY ITS2 tapes from resonance parameters, and by mixing other WIMLIB data, respectively. (U.K.)

  14. The WIMS characteristics method in a subgroup resonance treatment

    A brief overview is given of the subgroup resonance capture as implemented in the WIMS code. Recent developments to the general geometry characteristics solution module in WIMS, known as CACTUS, may be used in combination with WIMS, subgroup modules to derive broad group shielded cross sections for almost any geometry in two dimension. This application is described, together with some sensitivity studies for simple pin cell case, and also an example of its use for a more complex geometry. (author). 9 tabs., 4 figs

  15. Overview of the Radioecological Research at KAERI

    Choi, Yong Ho; Lim, Kang Muk; Kim, Byung Ho; Keum, Dong Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper presents a brief history of the research and a summary of the data production. During the past 30 years, a comparatively large amount of radioecological data for food crops was produced at KAERI. Some of the data have been used for the off-site dose calculation or dynamic food-chain model validation in one way or another. A considerable amount of KAERI data was included in an IAEA's handbook and underlying TECDOC. Further studies should be conducted to have sufficient numbers of parameter values to realistically cover various environmental and agricultural conditions. It is desirable for as many of the produced data as possible to be used by the dose assessor. Not only the data producer but also the dose assessor needs to make an effort for a greater amount of the domestic data to be used in estimating the public dose for Koreans. Radioecology is a scientific discipline for studying the movement and accumulation of radionuclides within ecosystems composed of air, soil, water and living organisms including humans. It started in the late 1940s in the USSR and the early 1950s in the USA for the purpose of assessing the environmental impact of the radionuclides released by military uses of fissile material. With an increase in the peaceful use of nuclear energy, radioecologists took a great interest in the environmental impact assessment of nuclear power plants and other nuclear fuel cycle facilities. Radiation doses to the public by the planned and ongoing operations of such nuclear installations should be estimated for both normal operation and an accident. These estimations are made using assessment models which require parameter values to quantify various transfer processes of radionuclides in the ecosystem. In KAERI, radioecological research has been conducted for the past 30 years with an emphasis put on the production of data on the transfer of radionuclides to major food crops.

  16. Overview of the Radioecological Research at KAERI

    This paper presents a brief history of the research and a summary of the data production. During the past 30 years, a comparatively large amount of radioecological data for food crops was produced at KAERI. Some of the data have been used for the off-site dose calculation or dynamic food-chain model validation in one way or another. A considerable amount of KAERI data was included in an IAEA's handbook and underlying TECDOC. Further studies should be conducted to have sufficient numbers of parameter values to realistically cover various environmental and agricultural conditions. It is desirable for as many of the produced data as possible to be used by the dose assessor. Not only the data producer but also the dose assessor needs to make an effort for a greater amount of the domestic data to be used in estimating the public dose for Koreans. Radioecology is a scientific discipline for studying the movement and accumulation of radionuclides within ecosystems composed of air, soil, water and living organisms including humans. It started in the late 1940s in the USSR and the early 1950s in the USA for the purpose of assessing the environmental impact of the radionuclides released by military uses of fissile material. With an increase in the peaceful use of nuclear energy, radioecologists took a great interest in the environmental impact assessment of nuclear power plants and other nuclear fuel cycle facilities. Radiation doses to the public by the planned and ongoing operations of such nuclear installations should be estimated for both normal operation and an accident. These estimations are made using assessment models which require parameter values to quantify various transfer processes of radionuclides in the ecosystem. In KAERI, radioecological research has been conducted for the past 30 years with an emphasis put on the production of data on the transfer of radionuclides to major food crops

  17. Radiological Emergency Response System of KAERI

    The Act of Physical Protection and Radiological Emergency came into effect in Feb. 2004. This act requires to the nuclear industries that the situation of the radiological emergency should be monitored by some proper equipment. To monitor the radiological emergency based on the act, KAERI, Korea Atomic Energy Research Institute, has been installing RERS, Radiological Emergency Response System, and establishing the implementation plan on the radiological emergency response. This paper describes the hardware and the operation of the RERS in view of the radiological emergency response

  18. Development of KAERI LBLOCA realistic evaluation model

    A realistic evaluation model (REM) for LBLOCA licensing calculation is developed and proposed for application to pressurized light water reactors. The developmental aim of the KAERI-REM is to provide a systematic methodology that is simple in structure and to use and built upon sound logical reasoning, for improving the code capability to realistically describe the LBLOCA phenomena and for evaluating the associated uncertainties. The method strives to be faithful to the intention of being best-estimate, that is, the method aims to evaluate the best-estimate values and the associated uncertainties while complying to the requirements in the ECCS regulations. (author)

  19. Waste management in decommissioning projects at KAERI

    Two decommissioning projects are being carried out at the KAERI (Korean Atomic Energy Research Institute), one for the Korea research reactors, KRR-1 and KRR-2, and another for the uranium conversion plant (UCP). The concept of the management of the wastes from the decommissioning sites was reviewed with relation to the decommissioning strategies, technologies for the treatment and the decontamination, and the characteristics of the waste. All the liquid waste generated from the KRR-1 and KRR-2 decommissioning site is evaporated by a solar evaporation facility and all the liquid waste from the UCP is treated together with the lagoon sludge waste. The solid wastes from the decommissioning sites are categorized into three groups; not contaminated, restricted releasable and radioactive waste. The not-contaminated waste will be reused and/or disposed of an industrial disposal site, and the releasable waste will be stored for a future disposal at the KAERI. The radioactive waste is packed into containers, and it will be stored at the decommissioning sites till it is sent to a national repository site. The reduction of the radioactive solid waste is one of the strategies for the decommissioning projects and could be achieved by a repeated decontamination. By the achievement of a minimization strategy, the amount of radioactive waste was reduced and the disposal cost will be reduced, but the cost for the manpower, and for a direct handling of the materials as well as for the administration was increased

  20. Upgrades to the WIMS-ANL code

    The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries

  1. Upgrades to the WIMS-ANL code

    The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries. (author)

  2. Over view of nuclear fuel cycle examination facility at KAERI

    Nuclear fuel cycle examination facilities at the Korea Atomic Energy Research Institute (KAERI) consist of two post-irradiation examination facilities (IMEF and PIEF), one chemistry research facility (CRF), one radiowaste treatment facility (RWTF) and one radioactive waste form examination facility (RWEF). This paper presents the outline of the nuclear fuel cycle examination facilities in KAERI. (author)

  3. Over view of nuclear fuel cycle examination facility at KAERI

    Lee, Key-Soon; Kim, Eun-Ga; Joe, Kih-Soo; Kim, Kil-Jeong; Kim, Ki-Hong; Min, Duk-Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-09-01

    Nuclear fuel cycle examination facilities at the Korea Atomic Energy Research Institute (KAERI) consist of two post-irradiation examination facilities (IMEF and PIEF), one chemistry research facility (CRF), one radiowaste treatment facility (RWTF) and one radioactive waste form examination facility (RWEF). This paper presents the outline of the nuclear fuel cycle examination facilities in KAERI. (author)

  4. A Study on the Revitalizing of technology commercialization in KAERI

    The TEC training program should be implemented for researches who want to commercialize their own technologies. To build creative organization culture is essential for technology commercialization. Collaboration strategy is related to analyze how KAERI is catching up their technological capabilities in nuclear technology, and what the success factors of KAERI in technology commercialization are.

  5. Analysis of the burnup credit benchmark with an updated WIMS-D Library

    The OECD/NEA Burnup Credit Benchmark was analyzed with the WIMSD5B code using a fully updated library based on ENDF/B-VI Revision 5 data. Parts-1A and 1B were considered. The criticality prediction tested in Part-1A was in very good agreement with the reference result. A slight trend to overestimate the absorption rate by the fission products was noted, which can be explained by spectral effects resulting from the coarseness of the WIMS-D 69-group energy grid. The isotopic composition prediction tested in Part-1B was within the uncertainty interval of the reference results, except for 109 Ag at lower burnup and 155 Gd in all the cases. For 109 Ag the cause of the discrepancy was the use of old fission yield data in generating the reference solution. Similarly for 155 Gd the difference was due to old 155 Eu capture cross sections. Compared to the measurements, a serious underprediction of Sm isotopes is observed. This could be due to problems in the measured values or in the nuclear data of Sm precursors. We conclude that our processing methods do not introduce significant errors to the basic nuclear data. Care should be taken in the interpretation of the reference average benchmark solution due to a possible bias towards the ENDF/B-V evaluated nuclear data files

  6. Advanced SFR Concept Design Studies at KAERI

    Advanced SFR design concepts have been developed which satisfy the Gen IV technology goals at KAERI. Two types of reactor core were developed for breakeven and TRU burner and both cores do not have blankets to enhance proliferation resistance. The Advanced SFR is a pool-type reactor that improves system safety through slow system transients. The heat transport system adopts two double wall tube Steam Generators and a passive Residual Heat Removal System PDRC. To secure the economic competitiveness of an SFR, the diameter of the reactor vessel of the Advanced SFR is designed to be 14.5 m, which is a very compact size compared to other designs. Also, various R and D activities have been performed in order to prepare some analysis tools and to support the development of design concepts. (author)

  7. Recent adjustments to the WIMS nuclear data library

    With few exceptions the WIMS library data have been unchanged since the 1960s and most of the validation of WIMS is based on datasets that were in existence at that time. The main aim of the present study was to re-assess the U235 thermal data. Other proposed data modifications include further adjustments to the U238 fast and resonance data, a correction to the Pu239 resonance integral and a consistent choice of fission spectrum. A variety of experiments from solid uranium metal to dilute solutions of U235, representative of the enormous range on which WIMS validation is based, have been analysed with the proposed data and the overall agreement is now generally very satisfactory. (U.K.)

  8. The WIMS-E module W-HEAD

    A description is given of the computer program W-HEAD, which is part of the WIMS-E Scheme for neutronics calculations, together with a detailed description of the input data to W-HEAD and the theory behind its use of resonance integrals. When a sequence of WIMS-E modules is submitted to a computer, W-HEAD is normally the first member of this sequence. It is connected to other parts of the WIMS-E scheme through a standard interface. W-HEAD calculates group averaged microscopic cross sections corrected for resonance self-shielding and the interaction effects of overlapping resonances. It makes use of equivalence theorems to relate the heterogeneous lattice cell problem to an equivalent homogeneous one. (author)

  9. The collision probability modules of WIMS-E

    This report describes how flat source first flight collision probabilities are calculated and used in the WIMS-E modular program. It includes a description of the input to the modules W-FLU, W-THES, W-PIP, W-PERS and W-MERGE. Input to other collision probability modules are described in separate reports. WIMS-E is capable of calculating collision probabilities in a wide variety of geometries, some of them quite complicated. It can also use them for a variety of purposes. (author)

  10. Preliminary analysis of the KAERI RCCS Experiment Using GAMMA+

    This paper describes the analysis of the KAERI RCCS experiment. GAMMA+ code was used for analysis of the RCCS 1/4-scale natural cooling experimental facility designed and built at KAERI to verify the performance of the natural circulation phenomenon. The results obtained from the GAMMA+ analysis showing the temperature profiles and flow rates at steady state were compared with the results from the preliminary experiments conducted in this facility. GAMMA+ analysis for the KAERI RCCS experimental setup was carried out to understand its natural circulation behavior. The air flow rate at the chimney exit achieved by experiments was from to be almost same as that of GAMMA+

  11. Development of tritium technologies at KAERI

    Korea has been operating a CANDU nuclear power plant since 1983. Tritium generated in the heavy water of the plant is removed by the Wolsong TRF (Tritium Removal Facility) and measurement campaigns of tritium near the power plant have shown the efficiency of the TRF system. The HANARO reactor uses heavy water as both reflector and moderator. In HANARO the tritiated water removal system consists of compressors, condensers, and adsorption beds. A tritium behavior analysis code (TRIBAC) for a Very High Temperature Gas-Cooled Reactor (VHTR) is under development at KAERI. The TRIBAC computer software has been equipped with models for tritium production, purification, and leakage, as well as chemisorption and tritium behavior, in the hydrogen production system. Korea takes part into the ITER program and is responsible for the supply of an SDS (Tritium Storage and Delivery System). Within this program Korea has launched an experimental program to study the physico-chemical properties of metal and their hydrides in which hydrogen isotope gases can be stored and removed safely

  12. Development of tritium technologies at KAERI

    Chung, H.; Koo, D.; Lee, J.; Park, J.; Yim, S.P.; Yoon, C.; Lim, J.; Choi, W.; Ahn, H.; Kang, H.; Kim, I.; Paek, S. [KAERI-UST, Yuseong, Daejeon (Korea, Republic of); Yunn, S.H.; Jung, K.J. [NFRI, Yuseong, Daejeon (Korea, Republic of)

    2015-03-15

    Korea has been operating a CANDU nuclear power plant since 1983. Tritium generated in the heavy water of the plant is removed by the Wolsong TRF (Tritium Removal Facility) and measurement campaigns of tritium near the power plant have shown the efficiency of the TRF system. The HANARO reactor uses heavy water as both reflector and moderator. In HANARO the tritiated water removal system consists of compressors, condensers, and adsorption beds. A tritium behavior analysis code (TRIBAC) for a Very High Temperature Gas-Cooled Reactor (VHTR) is under development at KAERI. The TRIBAC computer software has been equipped with models for tritium production, purification, and leakage, as well as chemisorption and tritium behavior, in the hydrogen production system. Korea takes part into the ITER program and is responsible for the supply of an SDS (Tritium Storage and Delivery System). Within this program Korea has launched an experimental program to study the physico-chemical properties of metal and their hydrides in which hydrogen isotope gases can be stored and removed safely.

  13. The WIMS-E module W-PROC

    The program W-PROC is a module of the WIMS-E Scheme for neutronics calculations. W-PROC calculates collision probabilities for a system consisting of spherical grains packed in annular geometry, such as the fuel in a high temperature gas cooled reactor. This report describes the modelling approximations made and gives instructions for using the program. (U.K.)

  14. Interaction of NRCs with their environment - KAERI's experience

    Main players in KAERI's environment are the Government, nuclear industry (essentially nuclear power related), Academic community and the public. The Board of Trustees of KAERI has members from three important ministries of the Government and this Board formulates the nuclear R and D programme. The current programme plan covers a period of 1996-2006. The Korean nuclear industry has grown out of the core groups within KAERI. Until 1996, certain key areas in the design of nuclear steam supply system, nuclear fuel and nuclear waste management were still a part of KAERI responsibilities. However, with the growth of the nuclear power programme to 14 GW(e) (16 reactors), and more reactors under construction and plan, a decision has been taken to shift these activities to the industry, along with the personnel (600). The Government has also decided to secure financial resources for R and D by a contribution of 0.1 cents/kw·h from the nuclear utilities to a fund. In 1998 this fund collected 90 million US$ and 75% was made available to KAERI. So there is a very strong linkage between the Government, KAERI and the nuclear industry. With the academic community, KAERI takes post-graduate and post doctoral research students, gives R and D projects to the universities and has joint projects in some areas like fusion research. With public, KAERI has followed the policy of openness. It has made specific efforts to convey more easily understood benefits of radioisotopes and radiation. Also, communication is quite often targeted at specific groups rather than public at large. This policy has helped in the public acceptance of nuclear power which provided 41% of the electricity in 1998. (author)

  15. Web-based sorption database (KAERI-SDB)

    Radionuclide sorption data is necessary for the safety assessment of radioactive waste disposal. However the accessibility to the nuclide sorption database is limited. The web-based sorption database (KAERI-SDB) was developed to provide sorption data in a convenient way. The development of the KAERI-SDB was achieved by improving the performance of pre-existing sorption DB programme (SDB-21C) and incorporating the user requirement. The KAERI-SDB was designed that users can access it by using a web browser. Main functions of the KAERI-SDB include (1) log-in/join, (2) search and store of sorption data and (3) scatter plot chart and index chart. It is expected that the KAERI-SDB is widely applied to the safety assessment of radioactive waste disposal by enhancing the accessibility to experts and practitioner related the nuclear industry and governmental administration. It is also expected that reliabilities for the radioactive waste disposal increased by opening the web-based sorption DB to public

  16. Use of weigh-in-motion (WIM) data for site-specific LRFR bridge rating

    ZHAO, HUA; Uddin, Nasim; Waldron, Christopher J.; O'Brien, Eugene J.

    2012-01-01

    In this paper, truck weigh-in-motion (WIM) data are used to develop live load factors for use on Alabama state-owned bridges. The factors are calibrated using the same statistical methods that were used in the original development of AASHTO’s Load and Resistance Factor Rating (LRFR) Manual. This paper describes the jurisdictional and enforcement characteristics in the state, the WIM data filtering, sorting, and quality control, as well as the calibration process. Large WIM data sets from five...

  17. WIMS/ABBN library based on the fond-2 evaluated files

    Description of a new system WIMS/ABBN is presented. It includes updated libraries of the WIMS/D4 code. The report involves sources and method of their creation, analysis and comparison of results at different tests also. Second path of the work is the development of the burnup library at WIMS system. As the result it was made a new system of burnup which new capabilities are shown. (authors)

  18. Assessment of the WIMS-D5 applicability to CANDU reactors

    The purpose of this study is to develop a WIMS/CANDU code for a lattice calculation on the basis of WIMS-D5 code for the safety analysis of CANDU reactors. To assess the WIMS-D5 applicability to a CANDU reactor, a lattice model was developed For CANDU-6 reactors at the Wolsong site. As for the benchmark of the code validation, the code-to-code comparison was performed between the WIMS-D5 code with both the 69- and 172-energy groups of ENDF/B-VI nuclear data library and the WIMS-AECL code with the 89-energy group. The comparison studies of the reactor physics parameters such as void reactivity', coolant/fuel/moderator temperature coefficients were conducted with the change of the internal isotopic composition due to the fuel burning-up using both WIMS-AECL and POWDERPUFS-V (PPV) codes. The results show that the present results between the WIMS-D5 code and WIMS-AECL code agreed well with those of the PPV at the beginning of the fuel horn-up phase. As burning-up progresses, the results of WIMS-D5 show a large deviation from those of PPV for CANDU 6 reactors. (author)

  19. Neutronic Calculations of TRIGA MARK-II with WIMS Cluster Options

    Neutronic calculations for RTP are made by using WIMS by utilizing several techniques. In this study, we explore the cluster options available in WIMS. In order to use this technique, the RTP core are split into several annulus containing both water and fuel. This enables us to determine the average flux at each annulus. This paper will demonstrate the required input card and general procedure for preparing WIMS input using the cluster option. Comparison of flux and multiplication factor between WIMS and experimental data are made and the amount of error estimated. (author)

  20. Development of Operational Parameters for Advanced Voloxidation Process at KAERI

    KAERI has been developing a voloxidation process as a head-end process of pyroprocessing technology with INL (Idaho National Laboratory). The work scope of KAERI is to develop the operation parameters for advanced voloxidation process at KAERI using surrogate materials and SIMFUEL. In order to evaluate operation conditions of an advanced voloxidation process, oxidation and vaporization behavior of metals and Cs compounds was investigated in terms of thermal treatment atmosphere and temperature by using thermodynamic data. And also, the oxidation and vaporization behavior of semi-volatile fission products with process pressure and temperature was investigated using surrogate materials. Particle size control for U3O8 powder was investigated using SIMFUEL and a rotary voloxidizer. According to analysis of KAERI works, the operation conditions for advanced voloxiation process may be consisted of the following four steps: 1) oxidation of UO2 pellet into U3O8 powder at 500 .deg. C in oxidative atmosphere, 2) additional oxidation of noble metal alloy and vaporization of high vapor pressure of fission products at 700 .deg. C in oxidative atmosphere, 3) granulation of U3O8 powder and vaporization of Cs compounds at 1200 .deg. C in an atmosphere of argon, and 4) reduction of UO2+x granules into UO2 granules at 1000 .deg. C in an atmosphere of 4%H2-Ar. This report will be used as a useful means for determining the operation parameters for advanced voloxidation process

  1. KAERI's challenge to steady production of radioisotopes and radiopharmaceuticals

    The Korea Atomic Energy Research Institute (KAERI) is a national organization in Korea, and has been doing many research and development works in radioisotope production and applications for more than 30 years. Now KAERI regularly produces radioisotopes (I-131, Tc-99m, Ho-166) for medical use and Ir-192 for industrial use. Various I-131 labeled compounds and more than 10 kinds of Tc-99m cold kits are also produced. Our multi-purpose reactor, named HANARO, has been operative since April of 1995. HANAKO is an open tank type reactor with 30 MW thermal capacity. This reactor was designed not only for research on neutron utilization but for production of radioisotopes. KAERI intended to maximize the radioisotope production capability. For this purpose, radioisotope production facilities (RIPF) have been constructed adjacent to the HANARO reactor building. There are four banks of hot cells equipped with manipulators and some of the hot cells were installed according to the KGMP standards and with clean rooms. In reviewing our RI production plan intensively, emphasis was placed on the development of new radiopharmaceuticals, development of new radiation sources for industrial and therapeutic use, and steady production of selected radioisotopes and radiopharmaceuticals. The selected items are Ho-166 based pharmaceuticals, fission Mo-99/Tc-99m generators. solution and capsules of I-131, and Ir-192 and Co-60 for industrial use. The status and future plan of KAERI's research and development program will be introduced, and will highlight programs for steady production. (author)

  2. Revised transport cross-sections for the WIMS library

    WIMS transport cross-sections above 4 eV are formed by a column-sum correction in which an assumed current spectrum is used to weight the P1 scattering data for a given isotope. Revised weighting spectra lead to improved transport cross-sections for the principal moderators: the effect on calculations of k-infinity is small but leakage calculations, for the homogenised cell, are now in close agreement with corresponding B1 calculations using explicit P1 data. Energy condensation of the B0 (transport corrected) equations appears to be more valid than the procedure used to condense the B1 equations. (author)

  3. The Wims-Traca code for the calculation of fuel elements. User's manual and input data

    The set of modifications and new options developped for the Wims-D code is explained. The input data of the new version Wims-Traca are described. The printed output of results is also explained. The contents and the source of the nuclear data in the basic library is exposed. (author)

  4. On the use of WIMS-7 for calculations on accelerator-driven systems

    De Kruijff, W.J.M.; Freudenreich, W.J.M

    1998-02-01

    The WIMS-7 code package has successfully been applied for a simple benchmark of a lead-cooled accelerator-driven system (ADS). With WIMS-7 it is possible to model a fixed source and to calculate the multiplication in a subcritical system. The calculations have shown that WIMS-7 is capable of treating this benchmark of a homogenized lead-cooled system with a fast neutron spectrum. The results described in this report are very promising and stimulate further investigation of WIMS-7 to study ADS-applications and lead-cooled reactor cores. It is useful to have a more extensive validation of WIMS-7 for lead-cooled ADS. In this report we have only considered a simple homogenized system. In the near future the application of WIMS-7 will be twofold. First, WIMS-7 can be applied to calculate the neutron spectrum in an accelerator-driven system in order to perform transmutation studies with a burnup code. Second, WIMS-7 can be used to study in more detail the neutronics of accelerator-driven systems. This is useful in order to learn more about the physics of accelerator-driven systems. 6 refs.

  5. On the use of WIMS-7 for calculations on accelerator-driven systems

    The WIMS-7 code package has successfully been applied for a simple benchmark of a lead-cooled accelerator-driven system (ADS). With WIMS-7 it is possible to model a fixed source and to calculate the multiplication in a subcritical system. The calculations have shown that WIMS-7 is capable of treating this benchmark of a homogenized lead-cooled system with a fast neutron spectrum. The results described in this report are very promising and stimulate further investigation of WIMS-7 to study ADS-applications and lead-cooled reactor cores. It is useful to have a more extensive validation of WIMS-7 for lead-cooled ADS. In this report we have only considered a simple homogenized system. In the near future the application of WIMS-7 will be twofold. First, WIMS-7 can be applied to calculate the neutron spectrum in an accelerator-driven system in order to perform transmutation studies with a burnup code. Second, WIMS-7 can be used to study in more detail the neutronics of accelerator-driven systems. This is useful in order to learn more about the physics of accelerator-driven systems. 6 refs

  6. DUPIC nuclear fuel manufacturing and process technology development at KAERI

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and has been under post irradiation examination. (Hong, J. S.)

  7. The Status of Implementation on Additional Protocol at KAERI

    Under the Additional Protocol, a State is required to provide the IAEA with further information on nuclear related activities, all buildings on a site, etc. through an expanded declaration and further access rights at a nuclear site and any location included in the expanded declaration. This paper describes the implementation status on the expanded declaration and the complementary access at KAERI under the AP. This paper reviewed the implementation status of the AP at KAERI, and there are some practical issues to prepare the expanded declarations as mentioned above. From the view point of the effective and efficient processing of the expanded declaration, it will be necessary to discuss the criterion for the definition of the nuclear fuel cycle-related R and D to be declared under the AP with IAEA

  8. A Study on the Export Control System at KAERI

    The current non-proliferation regime requires strengthening the export control from Korea to foreign countries. This means that the ministries related to export control deeply emphasize the prohibition of the illegal proliferation in the domestic society as well as international society. The principle of export control for non-proliferation of WMD is to control the transfer of the strategic items/technology to the countries which intend to develop the WMD in accordance with the multilateral agreements of the Nuclear Supply Group (NSG), Wassenaar Agreement (WA), Austrian Group (AG) and Missile Technology Control Regime (MTCR). Among them, export controls at KAERI are deeply related to the guidelines of the NSG, an international nuclear export control regime. Since the new concept of an export system was launched in Jan. 2014, KAERI needs to consider new approaches to meet the requirement of the revised domestic law and regulation. To cope with this environmental change, this paper suggests new approaches to effectively conduct the export control at KAERI

  9. Evaluation of ENDF/B-VI library with WIMS-AECL/RFSP Code system

    The object of this research is the evaluation of the cross-section charicteristics of ENDF/B-VI WIMS-AECL library against ENDF/B-V library previously used in the validation of WIMS-AECL code. validation of WIMS-AECL code had been carried out through the Phase-B post simulation of Wolsong Units 2, 3 and 4 before. Discrepancies between the calculated and measured values were thought to be mainly from observation errors and partly from the ENDF/B-V library. Till now, there had been various validation calculations for ENDF/B-VI library in the field of PWR but not in CANDU-PHWR. We herein, evaluated the ENDF/B-VI WIMS-AECL library for Wolsong Unit 4 by comparing the results with previous ones of ENDF/B-V for the same reactor unit with same WIMS/RFSP code system. It can be summarized that the Phase-B post simulation results of WIMS/RFSP with ENDF/B-VI are better than those of ENDF/B-V, because of less difference between calculated and measured values. There must be further study with different core conditions, however, for the exact evaluation of ENDF/B-VI WIMS-AECL library including calculations of many other physical parameters and the treatment of isotopes which is not in ENDF/B-VI but in ENDF/B-V

  10. Calculation of doppler coefficient of reactivity by WIMS code

    The Doppler coefficient of reactivity is an important factor in prediction of several transients in light water reactors. Some of the past studies raised the question about the 10% uncertainty that traditionally was taken in calculations of Doppler coefficient by LWR lattice code. In order to bridge the gap of lack of accurate benchmark problem to evaluate the accuracy of Doppler effect, Mosteller et al. proposed a computational benchmark problem of Doppler coefficient to evaluate the accuracy and consistency of LWR lattice physics code. In this paper we present the results obtained from WIMS-D4 lattice code and compare it with those obtained by CELL-2 lattice code part of the EPRI-PRESS reactor physics package. The results obtained from the Monte Carlo code MCNP-3A served as reference for both cases, and was taken from ref 1. (authors). 4 refs., 2 figs., 1 tab

  11. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d4 code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  12. Risorse pedagogiche interattive: accesso e condivisione. Una soluzione WIMS+SAML

    Cazzola, M

    2011-01-01

    Il poster presenta l'integrazione di WIMS (WWW Interactive Multipurpose Server) con Shibboleth, realizzata presso l'Università di Milano-Bicocca al fine di gestire l'autenticazione e l'autorizzazione degli utenti istituzionali sulla piattaforma.

  13. Contributions to and expectations from the CRP - KAERI (Republic of Korea). KAERI activity plan for MONJU CRP

    Full text: Thank you very much for your invitation to the IAEA CRP on 'Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel'. In KAERI we have developed a system analysis code as well as a CFD code for analysis of fluid flow and heat transfer in the upper plenum of liquid metal reactor. To perform a benchmark analysis of sodium natural convection in MONJU reactor vessel will be a valuable chance for KAERI to validate the code by comparing the prediction results directly with the measured plant data. It will also be a basis to refine the reactivity feedback model of SSC-K code because the reactivity model due to a expansion of CRDL is closely related to the temperature distribution in hot pool. The analysis of natural convection (thermal stratification) in the upper plenum of liquid metal reactor is also a challenging problem for the CFD code developers since the thermal stratification is not well modelled by the usual turbulence model like the k-epsilon model. Thus, KAERI will certainly participate in the benchmark analyses, and the analyses will be carried out in two ways; (1) Analysis of the Natural Convection with the CFD code - Carry out computation for the benchmark problem using the CFD code. - Validation of the CFD code using the experimental data. - Investigation of the choice of turbulence model on the accuracy of the solution. (2) Analysis of the Natural Circulation with the System Analysis Code SSC-K - Analyses of the natural convection experiments with the SSC-K two-dimensional pool model. - Evaluation and validation of the SSC-K pool model with respect to the measured data and other analyses. - Identification of other issues and R and D needs. (author)

  14. Establishment of database system for management of KAERI wastes

    Radioactive wastes generated by KAERI has various types, nuclides and characteristics. To manage and control these kinds of radioactive wastes, it comes to need systematic management of their records, efficient research and quick statistics. Getting information about radioactive waste generated and stored by KAERI is the basic factor to construct the rapid information system for national cooperation management of radioactive waste. In this study, Radioactive Waste Management Integration System (RAWMIS) was developed. It is is aimed at management of record of radioactive wastes, uplifting the efficiency of management and support WACID(Waste Comprehensive Integration Database System) which is a national radioactive waste integrated safety management system of Korea. The major information of RAWMIS supported by user's requirements is generation, gathering, transfer, treatment, and storage information for solid waste, liquid waste, gas waste and waste related to spent fuel. RAWMIS is composed of database, software (interface between user and database), and software for a manager and it was designed with Client/Server structure. RAWMIS will be a useful tool to analyze radioactive waste management and radiation safety management. Also, this system is developed to share information with associated companies. Moreover, it can be expected to support the technology of research and development for radioactive waste treatment

  15. Nuclear data evaluation and group constant generation for reactor analysis

    A new 69-group nuclear data library for WIMS-KAERI code was generated using the ENDF/B-V, IV, JENDL-2, and ENDL-84 data and NJOY which is nuclear data processing code. Thermal reactor benchmark problems recommended by the Cross Section Evaluation Working Group at BNL were analyzed using this new library and WIMS-KAERI code. Using 14 benchmark problems the calculated average value and standard deviation for effective multiplication factors were 1.00303 and 0.00514, respectvely.(Author)

  16. Design of radioactive wastes management integration system (RAWMIS) in KAERI

    An Radioactive Wastes Management Integration Systeme(RA WMIS) for the safe management of radioactive waste and spent fuel in KAERI is developed to collect basic information, provide the framework for national regulation, and efficiency in the management of radioactive waste and spent fuel. This system can also provide end-users access to information such as a statistical documents and integrated data from various waste generators to meet increased researchers needs and interests. We use result to find out entities of the number of 18 cases similar system study in the inside and outside of the country and analyze works in the radioactive waste treatment facility. We design database schema, entity-relationship diagram and prototyping input/output item. This system will be support to the study for radioactive material valance and inventory

  17. Reliability analysis of digital I and C systems at KAERI

    This paper provides an overview of the ongoing research activities on a reliability analysis of digital instrumentation and control (I and C) systems of nuclear power plants (NPPs) performed by the Korea Atomic Energy Research Institute (KAERI). The research activities include the development of a new safety-critical software reliability analysis method by integrating the advantages of existing software reliability analysis methods, a fault coverage estimation method based on fault injection experiments, and a new human reliability analysis method for computer-based main control rooms (MCRs) based on human performance data from the APR-1400 full-scope simulator. The research results are expected to be used to address various issues such as the licensing issues related to digital I and C probabilistic safety assessment (PSA) for advanced digital-based NPPs. (author)

  18. MASTER- an indigenous nuclear design code of KAERI

    KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers

  19. Summary of the Safety Culture Activities in HANARO of KAERI

    The definition of safety culture in HANARO takes the IAEA's definition and it is the assembly of characteristics of attitudes in the HANARO center and individuals which establishes that, as an overriding priority, the HANARO safety issues receive the attention warranted by their significance. Since the power operation of HANARO started in 1996, HANARO has been operated for about 11 years and its degree of utilization and the number of experimental facilities have increased. This achievement is partly due to the spread of safety culture to the operators and the reactor users. In this paper, the safety culture activities done by the HANARO center of KAERI are described, and its efforts necessary for an improvement of it are presented

  20. Status of the Decommissioning Engineering System Code Development of KAERI in 2014

    Various information systems have been developed and used at decommissioning sites for planning a project, record keeping for a post management and cost estimation. KAERI is the only expert group which has decommissioning experiences and KAERI is trying to develop computer code to converge all the data which has been accumulated during KRR-1 and 2 and UCP (Uranium Conversion Plant) decommission. KAERI DES supports two kinds of platform; web-based or application oriented program. This paper describes current status and features of KAERI DES application. As a responsible leading group of Korean decommissioning research field, KAERI has been developing DES application program, which is going to be an important mile stone of decommission industry in Korea. User friendly graphical interface and lots of actual data let people well understood on decommission cost evaluation. It is expected that continuous effort and funds will be delivered to this research

  1. Status of the Decommissioning Engineering System Code Development of KAERI in 2014

    Jin, Hyung Gon; Park, S. K.; Park, H. S.; Song, C. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Various information systems have been developed and used at decommissioning sites for planning a project, record keeping for a post management and cost estimation. KAERI is the only expert group which has decommissioning experiences and KAERI is trying to develop computer code to converge all the data which has been accumulated during KRR-1 and 2 and UCP (Uranium Conversion Plant) decommission. KAERI DES supports two kinds of platform; web-based or application oriented program. This paper describes current status and features of KAERI DES application. As a responsible leading group of Korean decommissioning research field, KAERI has been developing DES application program, which is going to be an important mile stone of decommission industry in Korea. User friendly graphical interface and lots of actual data let people well understood on decommission cost evaluation. It is expected that continuous effort and funds will be delivered to this research.

  2. NJOY installation on μVAX-II at IJS verification for WIMS library applications

    The code NJOY-87 recently became available. A distribution tape was obtained from RSIC and the code was successfully installed on the μVAX-II machine at the Jozef Stefan Institute. Before the test cases could be executed some minor corrections were required. The results differ slightly from the reference solution, particularly the self shielded scattering matrices. For LWR core applications the differences do not seem important but for other applications they should be examined more closely. In NJOY verification for WIMS library applications the emphasis is on the cross section definitions and the processing errors. The proposed procedure is to create the WIMS library using two independent codes, compare the cross section where possible and analyse in detail the results of the WIMS calculations for some standard benchmark lattice

  3. Use of WIMS-E lattice code for prediction of the transuranic source term for spent fuel dose estimation

    A recent source term analysis has shown a discrepancy between ORIGEN2 transuranic isotopic production estimates and those produced with the WIMS-E lattice physics code. Excellent agreement between relevant experimental measurements and WIMS-E was shown, thus exposing an error in the cross section library used by ORIGEN2

  4. A study on the mid- and long-term strategies of KAERI's future vision

    In this study, KAERI's role, expected changes of KAERI's management environments and its implementational directions, strategical direction of nuclear R and D, and KAERI's future vision and sustainable growth strategies beyond the up-to date successful achievements were analysed. The purpose of this study is investigating recent rapid changes of KAERI's political and management environments to establish future vision and growth strategies of KAERI in the 21st centuries. KAERI has performed its' mission as a government funded research organization successfully and significantly contributed promotion of national nuclear industry and capabilities through manpower development and self-reliance of nuclear power technologies and academic advancement in the fields of nuclear energy. That way, it has contributed to supply stable energy and develop economy and industry as well. In order to respond properly to newly emerged missions, integral and systematic institutional efforts are required to secure more research findings from the central and local Government and industries as well. High-quality human resources having creative expertise, experiences and skills are pre-requisite for securing competitiveness of nuclear technologies and industries. So it is essential to request the Governmental support and establish the manpower development plan in long term bases. KAERI is now standing at the turning moment to take off from the catch-up strategy of the advanced nuclear technologies (KAERI 1.0) into the innovative and creative vision and challenges, that is to say, KAERI 2.0, to establish an new technological culture, respond to social requirements and seek the international leading role

  5. A general introduction to the use of the WIMS-E modular program

    This report describes the WIMS-E Scheme for Neutronics Calculations. This was originally set up to extend the existing calculational facilities in the neutronics field so as to cover a wider range of requirements and to permit more rapid changes to meet future requirements. It consists of a family of compatible reactor physics programs which pass data to each other by way of sets of standardised files. These separate programs have also been combined to form the Integrated WIMS-E Modular Program which contains them as overlay segments together with a controlling routine which selects modules for execution in response to users' commands. (author)

  6. Validation of a new library of nuclear constants of the WIMS code; Validacion de una nueva biblioteca de constantes nucleares del Codigo WIMS

    Aguilar H, F. [Departamento de Experimentacion, Gerencia del Reactor, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-10-15

    The objective of the present work is to reproduce the experimental results of the thermal reference problems (benchmarks) TRX-1, TRX-2 and BAPL-1 to BAPL-3 with the WIMS code. It was proceeded in two stages, the first one consisted on using the original library of the code, while in the second one, a library that only contains the present elements in the benchmarks: H{sup 1}, O{sup 16}, Al{sup 27}, U{sup 235} and U{sup 238} was generated. To generate the present nuclear data in the WIMS library, it was used the ENDF/B-IV database and the Data processing system of Nuclear Data NJOY, the library was generated using the FIXER code. (Author)

  7. Recent Progress on the DUPIC Fuel Fabrication Technology at KAERI

    Jung-Won Lee; Ho-Jin Ryu; Geun-Il Park; Kee-Chan Song [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-ku, Daejeon, 305-353 (Korea, Republic of)

    2008-07-01

    Since 1991, KAERI has been developing the DUPIC fuel cycle technology. The concept of a direct use of spent PWR fuel in Candu reactors (DUPIC) is based on a dry processing method to re-fabricate Candu fuel from spent PWR fuel without any intentional separation of the fissile materials and fission products. A DUPIC fuel pellet was successfully fabricated and the DUPIC fuel element fabrication processes were qualified on the basis of a Quality Assurance program. Consequently, the DUPIC fuel fabrication technology was verified and demonstrated on a laboratory-scale. Recently, the fuel discharge burn-up of PWRs has been extended to reduce the amount of spent fuel and the fuel cycle costs. Considering this trend of extending the fuel burn-up in PWRs, the DUPIC fuel fabrication technology should be improved to process high burn-up spent fuels. Particularly the release behavior of cesium from the pellet prepared with a high burn-up spent fuel was assessed. an improved DUPIC fuel fabrication technology was experimentally established with a fuel burn-up of 65,000 MWd/tU. (authors)

  8. Thorium fuel cycle concept for KAERI's accelerator driven system project

    Korea Atomic Energy Research Institute (KAERI) has been carrying out accelerator driven system related research and development called HYPER for transmutation and energy production. HYPER program is aiming to develop the elemental technologies for the subcritical system by 2001 and build a small bench scale test facility (∼5MW(th)) by the year 2006. Some major features of HYPER have been developed and employed, which are on-power fueling concepts, a hollow cylinder-type metal fuel, and Pb-Bi as a coolant and spallation target material. Another fuel cycle concept for HYPER has been also studied to utilize thorium as a molten salt form to produce electricity as well as to transmute TRU elements. At the early stage of the fuel cycle, fissile plutonium isotopes in TRU will be incinerated to produce energy and to breed 233U from thorium. Preliminary calculation showed that periodic removal of fission products and small amount of TRU addition could maintain the criticality without separation of 233Pa. At the end of the fuel cycle, the composition of fissile plutonium isotopes in TRU was significantly reduced from about 60% to 18%, which is not attractive any more for the diversion of plutonium. Thorium molten salt fuel cycle may be one of the alternative fuel cycles for the transmutation of TRU. The TRU remained at the end of fuel cycle can be incinerated in HYPER having fast neutron spectrums. (author)

  9. Overview of Digital I and C PSA Research in KAERI

    This paper provides an overview of the ongoing research activities on probabilistic safety assessment (PSA) of digital instrumentation and control (I and C) systems in nuclear power plants (NPPs) performed by Korea Atomic Energy Research Institute (KAERI). The research activities are performed mainly on the methodology aspect and the model aspect of the digital I and C PSA. The methodology aspect includes the development of a new safety-critical software reliability analysis method by integrating the advantages of existing software reliability analysis methods, the fault coverage estimation method based on the fault injection experiments, and a new human reliability analysis method for computer-based main control rooms (MCRs) based on human performance data from the APR-1400 simulator. The model aspect includes the development of the digital-based safety-critical I and C systems such as digitalized reactor protection system (RPS) and engineered safety features component control system (ESF-CCS). The research results are expected to be used to address various issues such as the licensing issues related to digital I and C PSA for advanced digital-based NPPs

  10. Current activities of post-irradiation examination at KAERI

    A wide range of post-irradiation examination (PIE) for the nuclear fuels irradiated at NPPs with different design characteristics have been carried out at PIEF at KAERI. The examination was conducted to evaluate the irradiation performances as well as the fuel integrities. The input data leading to the design upgrades of the nuclear fuels have mostly been obtained from the PIE of the irradiated fuels. A comprehensive non-destructive and destructive examination equipment are incorporated with the hot cell examination system. The main activity of PIEF is concentrated on the commercial nuclear fuel examination as the IMEF focused on the HANARO irradiated fuel and material examination. Recently, the above mentioned two facilities put great concentrations on the examination of the structural components of the fuel assembly such as skeleton, spacer grid and hold down spring elements to cope with the safety requirements of fuel integrities to meet a highly extended burn up conditions. In this paper, a brief and general activity of the both facilities and the future scope of work are introduced. (author)

  11. Recent Progress on the DUPIC Fuel Fabrication Technology at KAERI

    Since 1991, KAERI has been developing the DUPIC fuel cycle technology. The concept of a direct use of spent PWR fuel in Candu reactors (DUPIC) is based on a dry processing method to re-fabricate Candu fuel from spent PWR fuel without any intentional separation of the fissile materials and fission products. A DUPIC fuel pellet was successfully fabricated and the DUPIC fuel element fabrication processes were qualified on the basis of a Quality Assurance program. Consequently, the DUPIC fuel fabrication technology was verified and demonstrated on a laboratory-scale. Recently, the fuel discharge burn-up of PWRs has been extended to reduce the amount of spent fuel and the fuel cycle costs. Considering this trend of extending the fuel burn-up in PWRs, the DUPIC fuel fabrication technology should be improved to process high burn-up spent fuels. Particularly the release behavior of cesium from the pellet prepared with a high burn-up spent fuel was assessed. an improved DUPIC fuel fabrication technology was experimentally established with a fuel burn-up of 65,000 MWd/tU. (authors)

  12. DUPIC fuel fabrication using spent PWR fuels at KAERI

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter III, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details

  13. Coupling of Wims-AECL and Origen-S for depletion calculations - 357

    One of the more powerful tools for isotope depletion calculations in neutron-irradiated material is the SCALE (Standardized Computer Analyses for Licensing Evaluation) module ORIGEN-S, maintained and developed by Oak Ridge National Laboratory. ORIGEN-S takes as input, in addition to a material description, a problem-dependent cross section library in which relative reaction rates for each nuclear process have been pre-evaluated. Creating different libraries for different stages of burnup, and for different materials, allows the 'point' code phenomenology of ORIGEN-S to be extended to more complicated geometries. To this end, AECL (Atomic Energy of Canada Limited) has coupled its successful 2-D neutron transport solver WIMS-AECL 2.5d to ORIGEN-S to create the coupled code 'WOBI' (WIMS-ORIGEN Burnup Integration). This code has been validated against PIE (post irradiation examination) results for CANDUTM reactors and for light-water reactors, and is extensively used at AECL to calculate exit compositions and decay heats for high and low enriched uranium fuels at the NRU (National Research Universal) research reactor located at the Chalk River Laboratories. In addition, because of the significantly expanded list of reactions available in ORIGEN-S, WOBI is more useful for advanced fuel cycle studies than WIMS-AECL alone. This paper discusses the validation results, and verification of WOBI against simple WIMS-AECL and ORIGEN-S stand-alone models. (authors)

  14. LWR-WIMS, a computer code for light water reactor lattice calculations

    LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)

  15. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable

  16. Burn up calculations for ETRR 1 and ETRR 2 reactors with wims and origen codes

    For ETRR -1 and ETRR - 2 research reactor, the 235 U depletion is determined with wims and origen codes the two calculated results show good agreement with each other. The buildup of different fission products (important from both the safety and protection point of view) is also calculated. The radioactivity and decay heat of the spent fuel is determined up to 30 years

  17. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors; Validacion del sistema de codigos WIMS-SNAP para calculos en reactores nucleares tipo TRIGA-MARK II

    Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu

    2000-07-01

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  18. Proceedings of 2008 KAERI/JAEA joint seminar on advanced irradiation and PIE technologies

    Under the Arrangement for Cooperation in the field of peaceful uses of Nuclear Energy between the Korea Atomic Energy Research Institute (KAERI) and the Japan Atomic Energy Agency (JAEA), the 2008 KAERI-JAEA Joint Seminar on Advanced Irradiation and PIE (post-irradiation examination) Technologies has been held at KAERI in Daejeon, Korea, from November 5 to 7, 2008. This seminar was organized by the PIE and Radwaste Division, Research Reactor Engineering Division, and HANARO Management Division in KAERI. It was also the first time to hold the seminar under the agreement signed September 4, 2008. This triennial seminar is the sixth in series of bilateral exchange of irradiation technologies. Since the first joint seminar on Post Irradiation Examination Technology between JAERI and KAERI held at JAERI Oarai center, Japan in 1992, it has been a good model of international cooperation program between KAERI and JAEA in the field of neutron irradiation uses. At the fifth seminar in 2005, irradiation technology field was included to the joint seminar, moreover in this time it is expanded to the research reactor management field for covering whole areas of irradiation using in research reactors. The seminar was divided into three technical sessions; the sessions addressed the general topics of 'research reactor management', 'advanced irradiation technology' and 'post-irradiation examination technology'. Total 46 presentations were made, and active information exchange was done among participants. This proceeding is containing the papers or manuscripts presented in the 2008 KAERI-JAEA Joint Seminar on Advanced Irradiation and PIE Technologies. The 46 of the presented papers indexed individually. (J.P.N.)

  19. Proceedings of the third JAERI-KAERI joint seminar on post irradiation examination technology

    Between the Department of JMTR of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Cycle Research Group of the Korea Atomic Energy Research Institute (KAERI), it has been periodically carried out the collaboration on technical information exchange by specialists and scientists, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI. And JAERI-KAERI joint seminar has been held every three years. The 1st and 2nd JAERI-KAERI Joint Seminars were held in November 1992 at JAERI and in September 1995 at KAERI, respectively. The 3rd JAERI-KAERI Joint Seminar was held on 25 and 26 March, 1999 at the Oarai Research Establishment of JAERI. In this seminar, total participants of 84 were joined from JAERI, KAERI, Hanyang University, Japan Nuclear Cycle Development Institute, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. Contributed presentations were in three sessions; Current status and future perspectives on PIE, PIE techniques and Evaluation of PIE data. Re-assembling technique for JOYO fuel, Nd-YAG laser welding technique, grain boundary analysis using FEG-TEM, lift time estimation of PWR Rod Cluster Control Assembly (RCCA) rodlet and failure analysis of Korea Nuclear Power Plant (KNP) fuel have been widely noticed as topic items on PIE. And some comments from PIE user, were pointed out that the nano-PIE technique, the flexibility to ad-hoc demands on testing space or utilization, and the international collaboration were very important for the next generation's PIE. The 34 of the present papers are indexed individually. (J.P.N.)

  20. Proceedings of the third JAERI-KAERI joint seminar on post irradiation examination technology

    NONE

    1999-09-01

    Between the Department of JMTR of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Cycle Research Group of the Korea Atomic Energy Research Institute (KAERI), it has been periodically carried out the collaboration on technical information exchange by specialists and scientists, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI. And JAERI-KAERI joint seminar has been held every three years. The 1st and 2nd JAERI-KAERI Joint Seminars were held in November 1992 at JAERI and in September 1995 at KAERI, respectively. The 3rd JAERI-KAERI Joint Seminar was held on 25 and 26 March, 1999 at the Oarai Research Establishment of JAERI. In this seminar, total participants of 84 were joined from JAERI, KAERI, Hanyang University, Japan Nuclear Cycle Development Institute, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. Contributed presentations were in three sessions; Current status and future perspectives on PIE, PIE techniques and Evaluation of PIE data. Re-assembling technique for JOYO fuel, Nd-YAG laser welding technique, grain boundary analysis using FEG-TEM, lift time estimation of PWR Rod Cluster Control Assembly (RCCA) rodlet and failure analysis of Korea Nuclear Power Plant (KNP) fuel have been widely noticed as topic items on PIE. And some comments from PIE user, were pointed out that the nano-PIE technique, the flexibility to ad-hoc demands on testing space or utilization, and the international collaboration were very important for the next generation's PIE. The 34 of the present papers are indexed individually. (J.P.N.)

  1. Proceedings of 2012 JAEA/KAERI joint seminar on advanced irradiation and PIE technologies

    Under the 'Arrangement for Corporation in the field of peaceful uses of Nuclear Energy between the Japan Atomic Energy Agency (JAEA) and the Korean Atomic Energy Research Institute (KAERI)', the 2012 JAEA/KAERI Joint Seminar on Advanced Irradiation and PIE (post-irradiation examination) Technologies has been held at Mito, Japan from March 28 to 30, 2012. This triennial seminar is the seventh in series of bilateral exchange of irradiation and PIE technologies and research reactor management. Since the first joint seminar on the PIE Technology between JAERI (Japan Atomic Energy Research Institute, former agency of JAEA) and KAERI was held at JAERI Oarai Research Institute, Japan in 1992, the international cooperation program between JAEA and KAERI has been actively carried out in the field of neutron irradiation. At the fifth seminar in 2005 and sixth in 2008, the irradiation technology and the research reactor management fields were included, respectively, to the joint seminar, and it covers whole areas of irradiation using research reactors. In this seminar total 37 presentations were made in three technical sessions, which are 'research reactor management', 'advanced irradiation technology' and 'post-irradiation examination technology', and active information exchange was done among participants. Papers or manuscripts presented in the 2012 JAEA/KAERI Joint Seminar on Advanced Irradiation and PIE Technologies are contained in the proceedings. (author)

  2. The WIMS-E modules W-PRES and W-RES

    The WIMS-E modules W-PRES and W-RES produce cross sections corrected for the effects of resonance absorption in complicated geometries. They can do this in any geometry for which collision probabilities can be calculated, and hence supplement the WIMS-E module W-HEAD which is limited to one dimensional slab or annular geometry. Resonance interaction may be taken into account for a mixture of resonance absorbers. However, it is convenient and usually adequate to calculate most resonances in W-HEAD and only the U-238 absorption resonances in W-RES. This report describes the method used and the data requirements of the programs. (U.K.)

  3. Evaluation of the performance of mini-WIMS in design calculations for SGHWR's

    In order to use the WIMS code for SGHWR design calculations it is desirable to reduce the computing time to a minimum. To this end, a study has been made of the effects of using condensed data libraries with few groups in the main transport routine and with coarse mesh representations. The results of initial lattice calculations are given in considerable detail for a set of SGHW experimental cores. The effects of condensation on attainable burnup and irradiated fuel composition for natural and enriched power reactor lattices have also been studied. Comparisons between detailed and condensed WIMS calculations are the main theme of the report but METHUSELAH and experimental results are included whenever possible. (author)

  4. Design of a Capacitive Flexible Weighing Sensor for Vehicle WIM System

    Qing Li

    2007-08-01

    Full Text Available With the development of the Highway Transportation and Business Trade, vehicle weigh-in-motion (WIM technology has become a key technology and trend of measuring traffic loads. In this paper, a novel capacitive flexible weighing sensor which is light weight, smaller volume and easy to carry was applied in the vehicle WIM system. The dynamic behavior of the sensor is modeled using the Maxwell-Kelvin model because the materials of the sensor are rubbers which belong to viscoelasticity. A signal processing method based on the model is presented to overcome effects of rubber mechanical properties on the dynamic weight signal. The results showed that the measurement error is less than ���±10%. All the theoretic analysis and numerical results demonstrated that appliance of this system to weigh in motion is feasible and convenient for traffic inspection.

  5. Validation of a new library of nuclear constants of the WIMS code

    The objective of the present work is to reproduce the experimental results of the thermal reference problems (benchmarks) TRX-1, TRX-2 and BAPL-1 to BAPL-3 with the WIMS code. It was proceeded in two stages, the first one consisted on using the original library of the code, while in the second one, a library that only contains the present elements in the benchmarks: H1, O16, Al27, U235 and U238 was generated. To generate the present nuclear data in the WIMS library, it was used the ENDF/B-IV database and the Data processing system of Nuclear Data NJOY, the library was generated using the FIXER code. (Author)

  6. Thermal neutron streaming effects and WIMS analysis of the Penn State subcritical graphite pile

    This analysis was performed on the Pennsylvania State University (PSU) subcritical reactor to find more accurate values for such nuclear parameters as the thermal fuel utilization factor, thermal diffusion length in the graphite, migration area, keff, etc. The analysis involved using the Winfrith Integrated Multigroup Scheme (WIMS) code as well as various hand calculations to find and compare those parameters. The data found in this analysis will be used by future students in the Penn State laboratory courses

  7. Calculations of WWER cells and assemblies by WIMS-7B code

    A study of the nuclear data libraries of the WIMS-7B code have been performed in calculations of computational benchmark problems. The benchmarks cover pin cell, single fuel assembly with several different fuel types, moderator densities. Fuel depletion is performed to a burnup of 60 MWd/kgNM in the WWER-1000 pin cell. The results of the analysis of the benchmark with different code systems have been compared and indicated good agreement among the different methods and data. (Authors)

  8. Wenders in Motion: A Study On The Way Of Characterization In Wim Wenders’ Road Movies

    Erman Pehlivan

    2013-06-01

    Full Text Available Famous for his road movies, WimWenders, has a deeper understanding of the word; motion. More than often,Wenders tells the story of wanderers by putting them in a constant state oftravel.  These travels alter thewanderers’ characters and play a key role in Wenders’ storytelling. This paperstudies Wim Wenders’ way of characterization in three parts. The first partstates out keywords to define Wim Wenders’ wanderers: movement/motion, thejourney, homesickness, spatial levels, the Ozu connection. And it mainlyfocuses on his road trilogy: Im Lauf der Zeit (1976, FalscheBewegung (1975, and Alice in den Städten (1974, whileexplaining the stated keywords. The second part only focuses on Paris, Texas(1984 and its main character Travis, who might be seen as the ultimatewanderer whom all other Wenders’ characters blend in to. Thethird and the final part takes Der Himmel Über Berlin (1987 as ‘avertical road movie’ hoping for finding a cure for the worldwide homesicknessthat all the Wenders’ wanderers suffer.  Also there are a handful of musicreferences hidden through the paper in homage to director’s love for rock ‘n’roll music.

  9. The KAERI 10 MeV Electron Linac - Description and Operational Manual

    The objective of this technical report is to guide the right operation and maintenance of the KAERI electron linac system. The KAERI electron linac system consists of 2 MeV injector based on 176 MHz Normal conducting RF (Radio Frequency)cavity and 10 MeV main accelerator based on 352 MHz Superconducting RF cavity, electron beamlines (injection and extraction). Since a electron accelerator generates hazard radiation, this system is located at the shielded room in basement and we can operate the system using the remote control system. It includes the description and the operational manual as well as the detailed technical direction for trouble shooting

  10. The KAERI 10 MeV Electron Linac - Description and Operational Manual

    Lee, Byung Cheol; Park, Seong Hee; Jung, Young Uk; Han, Young Hwan; Kang, Hee Young

    2005-06-15

    The objective of this technical report is to guide the right operation and maintenance of the KAERI electron linac system. The KAERI electron linac system consists of 2 MeV injector based on 176 MHz Normal conducting RF (Radio Frequency)cavity and 10 MeV main accelerator based on 352 MHz Superconducting RF cavity, electron beamlines (injection and extraction). Since a electron accelerator generates hazard radiation, this system is located at the shielded room in basement and we can operate the system using the remote control system. It includes the description and the operational manual as well as the detailed technical direction for trouble shooting.

  11. The Status of Development on a Web-Based Nuclear Material Accounting System at KAERI

    Lee, Byungdoo; Kim, Inchul; Lee, Seungho; Kim, Hyunjo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The Integrated Safeguards (IS) has been applied to 10 nuclear facilities and 1 location outside facility (LOF) at the Korea Atomic Energy Research Institute (KAERI) since July 2008. One of the major changes in the implementation of safeguards under the IS is to apply the concept of a Random Interim Inspection (RII) instead of an interim inspection. The RII plan is notified within a few hours under the IS. It is thus difficult for facility operators to prepare the inspection documents within a short time if they do not periodically manage and process the nuclear material accounting data at each facility. To resolve these issues, KAERI developed a Web-based accounting system with the function of a near real-time accounting (NRTA) system to effectively and efficiently manage the nuclear material accounting data produced at the nuclear facilities and cope with a short notice inspection under the IS, called KASIS (KAeri Safeguards Information treatment System). The facility operators must input the accounting data on the inventory changes, which are the transfers of nuclear materials among the nuclear facilities and the chemical/physical composition changes, into the KASIS. KAERI also established an RFID system for controlling and managing the transfer of nuclear material and/or radioactive materials between the nuclear facilities for the purpose of nuclear safety management, and developed the nuclear material accounting system with the functions of inventory management of nuclear material at the facility level.

  12. Korea Atomic Energy Research Institute (KAERI) in the 21st century

    Abstract. KAERI (Korea Atomic Energy Research Institute), a national nuclear research institute in the Republic of Korea, celebrated its fortieth anniversary last April. It has played a key role in the Korean nuclear history such that it: initiated and promoted the peaceful uses of nuclear energy in the Republic of Korea; maintained nuclear expertise on whole spectrum of nuclear field through conducting nuclear R and D programs, operating nuclear research facilities, and training and educating specialized nuclear personnel; founded a cornerstone of Korean nuclear industry by participating in the establishment of a nuclear engineering company and a nuclear fuel company and localizing nuclear fuel and reactor technology; and contributed to nuclear safety regulation by incubating a specialized nuclear regulatory body. Recently, to concentrate on nuclear R and D on advanced technology, KAERI went through management reform such as: the transfer of nuclear engineering divisions responsible for NSSS design and nuclear fuel design to nuclear industry in 1996; and the downsizing of manpower in 1998. Currently KAERI is in the challenging stage in terms of its missions and manpower. In the coming 21st century, KAERI is required to maintain the current R and D momentum and also to conduct priority-based research requiring concentrated effort. (author)

  13. A study on the establishment of nuclear R and D plan and policy in KAERI

    In this study, all the R and D projects of KAERI based on the national 'mid- and long-term nuclear energy R and D program' were classified into 'top-priority projects', 'core projects', 'basic and fundamental projects' and 'preceding projects' according to the time-depending importance and the characteristics of R and D projects. This study also suggested the differential support of R and D resources among projects on the basis of the new classification. It should be required the cooperation and support of the domestic nuclear family for KAERI to be regenerated into the internationally excellent institute through the new R and D policy. It is also necessary to be made a proper condition for the effective accomplishment off the new R and D policy, for example, an implementation of the policy under the full responsibility of KAERI. It is expected that the results of this study will be used as basic data for establishment of the KAERI's long-term vision and the planning of the national nuclear energy R and D program, and reflected in the national nuclear energy R and D policy

  14. The Status of Development on a Web-Based Nuclear Material Accounting System at KAERI

    The Integrated Safeguards (IS) has been applied to 10 nuclear facilities and 1 location outside facility (LOF) at the Korea Atomic Energy Research Institute (KAERI) since July 2008. One of the major changes in the implementation of safeguards under the IS is to apply the concept of a Random Interim Inspection (RII) instead of an interim inspection. The RII plan is notified within a few hours under the IS. It is thus difficult for facility operators to prepare the inspection documents within a short time if they do not periodically manage and process the nuclear material accounting data at each facility. To resolve these issues, KAERI developed a Web-based accounting system with the function of a near real-time accounting (NRTA) system to effectively and efficiently manage the nuclear material accounting data produced at the nuclear facilities and cope with a short notice inspection under the IS, called KASIS (KAeri Safeguards Information treatment System). The facility operators must input the accounting data on the inventory changes, which are the transfers of nuclear materials among the nuclear facilities and the chemical/physical composition changes, into the KASIS. KAERI also established an RFID system for controlling and managing the transfer of nuclear material and/or radioactive materials between the nuclear facilities for the purpose of nuclear safety management, and developed the nuclear material accounting system with the functions of inventory management of nuclear material at the facility level

  15. KAERI's challenge to steady production of radioisotopes and radiopharmaceuticals

    Park, J.H.; Han, H.S.; Park, K.B. [Korea Atomic Energy Research Institute, Taejon (Korea)

    2000-10-01

    The Korea Atomic Energy Research Institute (KAERI) is a national organization in Korea, and has been doing many research and development works in radioisotope production and applications for more than 30 years. Now KAERI regularly produces radioisotopes (I-131, Tc-99m, Ho-166) for medical use and Ir-192 for industrial use. Various I-131 labeled compounds and more than 10 kinds of Tc-99m cold kits are also produced. Our multi-purpose reactor, named HANARO, has been operative since April of 1995. HANAKO is an open tank type reactor with 30 MW thermal capacity. This reactor was designed not only for research on neutron utilization but for production of radioisotopes. KAERI intended to maximize the radioisotope production capability. For this purpose, radioisotope production facilities (RIPF) have been constructed adjacent to the HANARO reactor building. There are four banks of hot cells equipped with manipulators and some of the hot cells were installed according to the KGMP standards and with clean rooms. In reviewing our RI production plan intensively, emphasis was placed on the development of new radiopharmaceuticals, development of new radiation sources for industrial and therapeutic use, and steady production of selected radioisotopes and radiopharmaceuticals. The selected items are Ho-166 based pharmaceuticals, fission Mo-99/Tc-99m generators. solution and capsules of I-131, and Ir-192 and Co-60 for industrial use. The status and future plan of KAERI's research and development program will be introduced, and will highlight programs for steady production. (author)

  16. Validation of MCNP and WIMS-AECL/DRAGON/RFSP for ACR-1000 applications

    Bromley, Blair P.; Adams, Fred P.; Zeller, Michael B.; Watts, David G.; Shukhman, Boris V.; Pencer, Jeremy [AECL - Chalk River Laboratories, Chalk River (Canada)

    2008-07-01

    This paper gives a summary of the validation of the reactor physics codes WIMS-AECL, DRAGON, RFSP and MCNP5, which are being used in the design, operation, and safety analysis of the ACR-1000{sup R}. The standards and guidelines being followed for code validation of the suite are established in CSA Standard N286.7-99 and ANS Standard ANS-19.3-2005. These codes are being validated for the calculation of key output parameters associated with various reactor physics phenomena of importance during normal operations and postulated accident conditions in an ACR-1000 reactor. Experimental data from a variety of sources are being used for validation. The bulk of the validation data is from critical experiments in the ZED-2 research reactor with ACR-type lattices. To supplement and complement ZED-2 data, qualified and applicable data are being taken from other power and research reactors, such as existing CANDU{sup R} units, FUGEN, NRU and SPERT research reactors, and the DCA critical facility. MCNP simulations of the ACR-1000 are also being used for validating WIMS-AECL/ DRAGON/RFSP, which involves extending the validation results for MCNP through the assistance of TSUNAMI analyses. Code validation against commissioning data in the first-build ACR-1000 will be confirmatory. The code validation is establishing the biases and uncertainties in the calculations of the WIMS-AECL/DRAGON/RFSP suite for the evaluation of various key parameters of importance in the reactor physics analysis of the ACR-1000. (authors)

  17. Assessment of a Bridge WIM System on Integral Concrete Bridges and on Steel Orthotropic Decks

    Ieng, Sio Song; SCHMIDT, Franziska; ROMBONI, Frédéric; Jacob, Bernard

    2011-01-01

    Bridge-Weigh-In-Motion uses bridges as a scale to weigh vehicles. Practically, this is done by measuring the strains in that bridge, and relating them to the weight and dimensions of a truck called “calibration trucks” whose shape and axle weights are well known. This article summarizes different B-WIM experiments the institute IFSTTAR (formerly called LCPC) realized and the lessons drawn from this experience. First, the system has been tested on frame-type bridges with integral s...

  18. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  19. A WIMS-NESTLE reactor physics model for an RBMK reactor

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

  20. A WIMS-NESTLE reactor physics model for an RBMK reactor

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given. (author)

  1. Internal structure of spiral arms traced with [CII]: Unraveling the WIM, HI, and molecular emission lanes

    Velusamy, T; Goldsmith, P F; Pineda, J L

    2015-01-01

    The spiral arm tangencies are ideal lines of sight in which to determine the distribution of interstellar gas components in the spiral arms and study the influence of spiral density waves on the interarm gas in the Milky Way. We present a large scale (~15deg) position-velocity map of the Galactic plane in [CII] from l = 326.6 to 341.4deg observed with Herschel HIFI. We use [CII] l-v maps along with those for Hi and 12CO to derive the average spectral line intensity profiles over the longitudinal range of each tangency. Using the VLSR of the emission features, we locate the [CII], HI, and 12CO emissions along a cross cut of the spiral arm. In the spectral line profiles at the tangencies [CII] has two emission peaks, one associated with the compressed WIM and the other the molecular gas PDRs. When represented as a cut across the inner to outer edge of the spiral arm, the [CII]-WIM peak appears closest to the inner edge while 12CO and [CII] associated with molecular gas are at the outermost edge. HI has broader ...

  2. CEDPA Africa Office honors Kenya M.P. Phoebe Asiyo, WIM 1.

    Kairu, M

    1993-10-01

    The discussion highlights the accomplishments of Phoebe Asiyo, who was an alumna of the first Women in Management (WIM) workshops of CEDPA Africa and a Member of Parliament from Western Kenya. Phoebe Asiyo was honored by over 100 guests at a CEDPA Africa Regional Office reception in Nairobi. CEDPA's president remarked that the Honorable Mrs. Asiyo had worked very hard on women's issues in her lifetime with enthusiasm, fresh ideas, and challenges to continue working with and for women. Mrs. Asiyo was elected to Parliament in 1992 as one of six other women after involvement in the Kenya campaign for democracy. Since independence, the number of women legislators has been the highest. Her legislative efforts were encouraged and sustained by involvement in the 1978 WIM workshop, after which she was elected to Parliament (1980). Between 1980 and 1988, there were only two women members: (Mrs. Asiyo and another). Between 1988 and 1992, Mrs. Asiyo served as ambassador to UNIFEM, the UN Development Fund for Women. Mrs. Asiyo considers that women's contribution to political life has been to provide the kind of leadership that empowers and enables the poor and grassroots communities to take control of their lives and their communities. Leaders in power are held accountable for their impact on an impoverished population, with new standards initiated by women. This type of leadership leads to long-term betterment in living conditions. PMID:12345285

  3. Analysis of Moderator Temperature Reactivity Coefficient of the PWR Core Using WIMS-ANL

    The Moderator Temperature Reactivity Coefficient (MTRC) is an important parameter in design, control and safety, particularly in PWR reactor. It is then very important to validate any new processed library for an accurate prediction of this parameter. The objective of this work is to validate the newly WIMS library based on ENDF/B-VI nuclear data files, especially for the prediction of the MTRC parameter. For this purpose, it is used a set of light water moderated lattice experiments as the NORA experiment and R1-100H critical reactors, both of reactors using UO2 fuel pellet. Analysis is used with WIMSD/4 lattice code with original cross section libraries and WIMS-ANL with ENDF/B-VI cross section libraries. The results showed that the moderator temperatures reactivity coefficients for the NORA reactor using original libraries is - 5.039E-04 %Δk/k/℃ but for ENDF/B-VI libraries is - 2.925E-03 %Δk/k/℃. Compared to the designed value of the reactor core, the difference is in the range of 1.8 - 3.8 % for ENDF/B-IV libraries. It can be concluded that for reactor safety and control analysis, it has to be used ENDF/B- VI libraries because the original libraries is not accurate any more. (author)

  4. WEDRO, Data Processing Routines for WIMS-D/4 WIMSE File

    1 - Description of program or function: The WIMSE Data Processing Routines (WEDRO) are a set of routines developed to process the WIMS-E file which is produced by the computer code WIMS-D/4.1. The data manipulation functions of WEDRO-1.1 include the following: - Spatial homogenization of cross-sections using flux-volume weighting - Energy condensation of the cross-sections with three variants of the Travelli formula available for collapsing microscopic transport cross sections - Lumping of fission products to a user specified scheme - Computation of Normalised Generalised Equivalence Theory (NGET) discontinuity factors for one-dimensional slab and two-dimensional pin cell problems - Generation of cross section files in three different applications formats (e.g. the WORKING file format in the AMPX and SCALE code packages). 2 - Method of solution: Data manipulation. 3 - Restrictions on the complexity of the problem: WEDRO-1.1 is a variably dimensioned code and there is thus no restriction on the number of energy groups etc. The size of the problem is only restricted by the core storage available

  5. The development of KAERI management information system (II) -The development of Time Sheet Management System-

    The purpose of this report is to describe the work done for the development, operation and maintenance of Time Sheet Management System. This work is a part of the development KAERI management information system. Manpower management is essential to cope with the external circumstances promptly and to maximize the productivity of the organization. This work aims at setting up a basis for the manpower management system. It is widely recognized that neither timely decision making nor competitive edge can be secured with the traditional management technology in so a rapidly changing situations home and abroad, which can be characterized by openness and informality. The necessity of efficient and scientific man-power management by time-study has emerged on the reorganization of KAERI by expanding matrix system in order to enhance the R and D productivity. (Author)

  6. WLUP3.0, 69 and 172 Group Cross Section Libraries for WIMS

    Description or function: WLUP contains validated WIMS-D formatted cross section libraries in 69 and 172 energy group structures for nuclear reactor calculations. Materials from recently released evaluated nuclear data libraries are included. The NJOY nuclear data processing system was applied for generating the cross section files following the models and conventions built into the WIMS-D lattice code. The relevant features for the WIMS users are: - Energy group structures: 69 and 172 energy groups. - List of materials: WIMS ID, general information, source of data. - Cross sections: 69 and 172 group plots. - Resonance data: WIMS ID, temperature, background cross sections. - Goldstein-Cohen factors: Goldstein-Cohen lambda values. - Thermal scattering data: thermal scattering laws and P1 matrixes. - Fission spectrum: fission spectrum data. - Burnup data: burnup chains. - Fission product yields: fission yield tables. - Pseudo lumped fission product: Description of pseudo fission product. - Energy release by fission: table of energy released by fission. - Dosimetry data: dosimetry reactions, source of data. - Averaging flux and current spectra: flux and current spectra plots (Numerical data on NJOY inputs). - WIMSD5B updates: WIMSD5B extensions and updates. - Processing methods: Brief description on processing methods. Moderators: 1-H-H2O, 1-H-ZrH, 1-D-D2O, 4-Be, 6-C, 8-O-16. Structural materials: 2-He-3, 2-He-4, 3-Li-6, 3-Li-7, 5-B-10, 5-B, 7-N, 9-F, 11-Na, 12-Mg, 13-Al, 14-Si, 15-P, 16-S, 17-Cl, 20-Ca, 22-Ti, 23-V, 24-Cr, 25-Mn, 26-Fe, 27-Co-59, 28-Ni, 29-Cu, 40-Zr, 41-Nb-93, 42-Mo, 47-Ag, 48-Cd, 49-In, 50-Sn, 51-Sb-121, 51-Sb-123, 63-Eu, 72-Hf, 73-Ta, 74-W, 82-Pb. Burnable materials: 5-B-10, 5-B-11, 72-Hf-176, 72-Hf-177, 72-Hf-178, 72-Hf-179, 72-Hf-180. Fission products: 36-Kr-83, 42-Mo-95, 43-Tc-99, 44-Ru-101, 44-Ru-103, 44-Ru-106, 45-Rh-103, 45-Rh-105, 46-Pd-105, 46-Pd-107, 46-Pd-108, 47-Ag-109, 48-Cd-113, 49-In-115, 51-Sb-125, 52-Te-127, 53-I-127, 53-I-135, 54-Xe

  7. Blasting Impact by the Construction of an Underground Research Tunnel in KAERI

    The underground research tunnel, which is under construction in KAERI for the validation of HLW disposal system, is excavated by drill and blasting method using high-explosives. In order not to disturb the operation at the research facilities such as HANARO reactor, it is critical to develop a blasting design , which will not influence on the facilities, even though several tens of explosives are detonated almost simultaneously. To develop a reasonable blasting design, a test blasting at the site should be performed. A preliminary analysis for predicting the expected vibration and noise by the blasting for the construction of the underground research tunnel was performed using a typical empirical equation. From the study, a blasting design could be developed not to influence on the major research facilities in KAERI. For the validation of the blasting design, a test blasting was carried out at the site and the parameters of vibration equation could be determined using the measured data during the test blasting. Using the equation, it was possible to predict the vibration at different locations at KAERI and to conclude that the blasting design would meet the design criteria at the major facilities in KAERI. The study would verify the applicability of blasting method for the construction of a research tunnel in a rock mass and that would help the design and construction of large scale underground research laboratory, which might be carried out in the future. It is also meaningful to accumulate technical experience for enhancing the reliability and effectiveness of the design and construction of the HLW disposal repository, which will be constructed in deep underground by drill and blasting technique

  8. Proceedings of 2005 JAEA-KAERI joint seminar on advanced irradiation and PIE technologies

    In this seminar, total participants of over 100 were jointed from JAEA, KAERI, Hanyang University, Chungnam National University, Kyung Hee University, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. The technical development and experimental data on the irradiation test and PIE were aggressively discussed in this seminar. Contributed presentations were 35 in three sessions; Current status and future program on irradiation test and PIE (10 presentations), Development of irradiation and PIE technologies (15 presentations) and Evaluation of irradiation and PIE data (10 presentations). Development of instrumented capsule technologies for HANARO irradiation, current PIE activities in each hot laboratory of both countries, development of irradiation capsules in JMTR for the Irradiation Assisted Stress Corrosion Cracking (IASCC) study, development of irradiation and PIE techniques for the safety research on the high burnup fuel, utilization plan of JOYO and development of MOX fuel containing americium have been widely noticed as topic items on irradiation and PIE technologies. This proceedings is containing papers presented in the 2005 JAEA-KAERI Joint Seminar. It also indicates the current status of the aggressive information exchange activity on two fields of irradiation test and PIE technologies between JAEA and KAERI under the Arrangement for the Implementation of Cooperative Research Program mentioned above. The 35 of the presented papers are indexed individually. (J.P.N.)

  9. A computer program with graphical user interface to plot the multigroup cross sections of WIMS-D library

    As a result of the IAEA Co-ordinated Research Programme entitled 'Final Stage of the WIMS Library Update Project', new and updated WIMS-D libraries based upon ENDF/B-VI.5, JENDL-3.2 and JEF-2.2 have become available. A project to prepare an exhaustive handbook of WIMS-D cross sections from old and new libraries has been taken up by the authors. As part of this project, we have developed a computer program XnWlup with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualization of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. The current features of the software, on-line help manual and future plans for further development are described in this paper

  10. Traffic weigh-in-motion (WIM measurements and validation of the Texas perpetual pavement structural design concept

    Lubinda F. Walubita

    2011-01-01

    Full Text Available Over the past few years, the State of Texas has used perpetual pavement (PP structures on its heavily trafficked highways, where the expected 20-year truck-traffic estimate of 80 kN ESALs (equivalent single axle loads is in excess of 30 million. As a means to validate the Texas PP structural design concept and to make optimal future truck-traffic design recommendations, traffic Weigh In-Motion (WIM measurements were conducted and analyzed for two PP projects. The findings indicated that the initial 80 kN ESAL traffic design estimates for PP were comparable to the projections based on the actual measured WIM traffic data. However, underestimation of the hot mix asphalt layer dynamic moduli resulted in conservative designs for the PP structures. In addition, based on the successful use of the automated WIM data stations for traffic data collection, the paper highlights possible applications and advantages (as compared to conventional manual collection of traffic data of using detailed WIM traffic data information for future analyses of both highway operation and pavement structural design.

  11. A study for developing training courses of the nuclear training center -with priority given to the training goals of KAERI-

    The final goal of this project, which covers 3 years (from 1992 to 1994), is to develop personnel training courses of the Korea Atomic Energy Research Institute (KAERI) and to derive the most desirable training system therefrom. To achieve this final goal successfully the first year's research was designed and has been carried on; firstly, to analyze the on-going issues and what kind of reform measures should be introduced to both the input and conversion processes of KAERI to efficiently achieve the organization goals, secondly, to derive personnel training goals of KAERI based on the analyses. First, this study introduced the viewpoint of systems approach for organization analysis, and defined that the productivity of an organization mainly depends on manpower quality of the input section and efficiency of the conversion process. Next, general organization theories and characteristics of research and development organization were studied, and derived that in research and development organization the expertise of a specialist should be regarded as the main value rather than his position, and the atmosphere should be human-centered, being free and democratic rather than authoritarian. And the study emphasizes more flatted structure of organization, necessity of sense of Management By Objectives (MBO), future planning capability, quality of manager with democratic leadership as criteria for the analysis of research and development organization. Finally, analyzing organization structure and behavior of KAERI based on the criteria, the study derived the ends-means hierarchy of personnel training of KAERI and discussed the necessity of organization reform of KAERI. (Author)

  12. Application Of WIMS Code To Calculation Kartini Reactor Parameters By Pin-Cell And Cluster Method

    Analysis UZrH fuel element parameters calculation in Kartini Reactor by WIMS Code has been done. The analysis is done by pin cell and cluster method. The pin cell method is done as a function percent burn-up and by 8 group 3 region analysis and cluster method by 8 group 12 region analysis. From analysis and calculation resulted K∼ = 1.3687 by pin cell method and K∼ = 1.3162 by cluster method and so deviation is 3.83%. By pin cell analysis as a function percent burn-up at the percent burn-up greater than 59.50%, the multiplication factor is less than one (k∼ < 1) it is mean that the fuel element reactivity is negative

  13. Distributed fuel-management computation using RFSP, WIMS-AECL and PVM

    The Parallel Virtual Machine (PVM) software package was used to build an interface between RFSP and WIMS-AECL to enable history-based, local-parameter reactor fuel-management simulations in which batches of lattice-cell transport and burnup calculations can be made in parallel. The interface is based on the master/slave crowd-computation model. For slave computers numbering from one to twenty, the overhead spent by the one master preparing input for the slaves and processing their outputs was observed to be small in comparison with the computing time spent by the slaves themselves. Anticipating the availability of a much larger network of slaves in the future, two potential computational bottlenecks that might arise are described, and possible remedies for them are outlined. (author)

  14. User Manual for XnWlup2.0, A Software to Visualize Nuclear Data for Thermal Reactors in WIMS-D Libraries

    A project to prepare an exhaustive handbook of WIMS-D cross sections for thermal reactor applications comparing different WIMS-D compatible nuclear data libraries originating from various countries has been successfully implemented. A computer software, called XnWlup2.0, with graphical user interface for MS Windows has been developed at BARC. This report summarizes the salient features of this new software for the users of WIMS-D libraries. Several sample outputs produced by the software are presented to illustrate the powerful use of this software for routine use in reactor physics analyses. (author)

  15. New Beam Line Design of TRIAC as a Stable Heavy-Ion Accelerator at KAERI

    Lee, Cheol Ho; Chang, Dae Sik; Oh, Byung Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of); Seo, Chang Seog; Yun, Chong Cheoul [Institute for Basic Science, Daejeon (Korea, Republic of); Jeong, Sun Chan [dHigh Energy Accelerator Research Organization, Tsukuba-shi (Japan)

    2012-05-15

    KEK (High Energy Accelerator Research Organization) TRIAC (Tokai Radioactive Ion Accelerator Complex) was a radioactive isotope accelerator which can provide beams of uranium fission fragments with the maximum energy of 1.1 MeV/nucleon produced by protons of 30 MeV and 1 {mu}A (30 W in beam power, actually deposited in the production target) from the JAEA Tandem Accelerator. Because of the critical limitations in the reaccelerated energy and intensity of available RIBs (Radioactive ion beams), TRIAC considered an upgrade program seriously, but it was canceled. Finally the complex had been closed at the end of 2010, and it was transferred to KAERI (Korea Atomic Energy Research Institute) after being disassembled to promote a new availability in Korea. KAERI team has a plan to reassemble this device as a stable ion beam accelerator with a minimized change for the low energy beam line including the ion source and the target system. The new stable ion accelerator will be used not only for the basic research but also for the application of heavy ion beams. Before the reassembling of TRIAC at KAERI, new layout of the beam line should be designed, and checked by beam optics simulation. The operation conditions and beam optics characteristics of the new beam line components can be understood with this simulation. The works that should be done before reassembling as a new machine have been done in this study. The beam optics calculations were preferentially carried out with arbitrary order beam physics code COSY INFINITY (COSY) or beam envelope code TRANSPORT

  16. New Beam Line Design of TRIAC as a Stable Heavy-Ion Accelerator at KAERI

    KEK (High Energy Accelerator Research Organization) TRIAC (Tokai Radioactive Ion Accelerator Complex) was a radioactive isotope accelerator which can provide beams of uranium fission fragments with the maximum energy of 1.1 MeV/nucleon produced by protons of 30 MeV and 1 μA (30 W in beam power, actually deposited in the production target) from the JAEA Tandem Accelerator. Because of the critical limitations in the reaccelerated energy and intensity of available RIBs (Radioactive ion beams), TRIAC considered an upgrade program seriously, but it was canceled. Finally the complex had been closed at the end of 2010, and it was transferred to KAERI (Korea Atomic Energy Research Institute) after being disassembled to promote a new availability in Korea. KAERI team has a plan to reassemble this device as a stable ion beam accelerator with a minimized change for the low energy beam line including the ion source and the target system. The new stable ion accelerator will be used not only for the basic research but also for the application of heavy ion beams. Before the reassembling of TRIAC at KAERI, new layout of the beam line should be designed, and checked by beam optics simulation. The operation conditions and beam optics characteristics of the new beam line components can be understood with this simulation. The works that should be done before reassembling as a new machine have been done in this study. The beam optics calculations were preferentially carried out with arbitrary order beam physics code COSY INFINITY (COSY) or beam envelope code TRANSPORT

  17. Development of Temperature Measurements and Calorimetry for the Neutral Beam Test Stand Operation at KAERI

    Operation of the Neutral Beam Test Stand(NB-TS) at Korea Atomic Energy Research Institute(KAERI) now reaches to 80 kV-20A for about 10 seconds. Experiments with this kind of enormous power and energy necessarily entail many temperature measurements at various locations of the system, and most of the beam line components require to be monitored of their temperatures. We have been implementing temperature measurement utilizing K-Type and T-Type thermocouples(TCs) and a Pt-100 resistance temperature detector for the instrumentation and control and for establishing calorimetry during the operation of the NB-TS facility

  18. The development of KAERI management information system -First year: The development of manpower information management system-

    The purpose of this report is to describe the implementation of the management information system for manpower. This job is the first year's for development KAERI management information system. It is important to properly manage a manpower to cope with the external circumstances promptly and to maximize the productivity of the organization. This report aims at basic management of manpower and uses multimedia to keep abreast with the times and introduces the concept of GUI (Graphic User Interface) to user for ease access. (Author)

  19. Experiences with NJOY and ENDF pre-processing system on PC-486 and benchmarks of WIMS library generated from ENDF/B-VI

    The objective of the present study is to develop WIMS-CITATION computation system using PC-486 for nuclear analysis of advanced Pressurized Water Reactors, whose burnups are designed to achieve more than 60,000 MWD/MTU, and CANDU reactors. In order to achieve the objective, it is essential to generate the proper WIMS library. The possibility of employing NJOY code and ENDF Pre-Processing System are under investigation. In order to verify the acceptability of the WIMS library generated from ENDF/B-VI on PC-486, new libraries are generated and compared with the libraries generated using CYBER and SUN Work Station for testing benchmark results of five light water and five heavy water lattices. The result shows that PC-486-based WIMS library is acceptable even though there exist some minor problems. (author). 11 refs, 16 figs, 9 tabs

  20. Verification of KAERI-DySCo for a dynamic simulation of VHTR-based SI hydrogen production facilities 1: sulfuric acid multistage distillation column module

    KAERI has developed the dynamic code (KAERI-DySCo) to analyze the start-up behaviors of the SI process components. This study focuses on the verification of a simulation module for the sulfuric acid multi-stage distillation column in the KAERI-DySCo. In agreement with the steady state values measured experimentally by KIST, it has been finally confirmed that the SAMDC, which is one of the simulation modules in KAERI-DySCo for the dynamic simulation code of VHTR-based SI hydrogen production facilities, is a feasible simulation module for calculating the start-up dynamic behavior of a sulfuric acid multistage distillation column

  1. Neutronic study of nuclear reactors. Complete calculation of TRIGA MARKII reactor and calculations of fuel temperature coefficients. (Qualification of WIMS code)

    The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)

  2. Evaluation for KAERI 6x6 Reflood Test Using TRACE Code

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed by the Korean nuclear industries. The SPACE is a best-estimated two phase three-field thermal-hydraulic analysis code to analyze the performance of pressurized water reactors and is under a licensing review by the regulatory body. For a new code, various SET/IET assessments should be performed to identify the accuracy of code/model. Among the SETs to evaluate the effect of reflood heat transfer, the KAERI 6x6 reflood test was evaluated by the only SPACE code. The 6x6 reflood test facility (ATHER) has been constructed at KAERI to investigate quantitatively the mechanism of reflood phenomena during the reflood phase of LBLOCA and to evaluate the effect of droplet flow on core cooling during the reflood phase. In this study, the ATHER test was assessed independently by the TRACE code. The objectives of this study are to identify the prediction capability of TRACE code and to utilize the prediction results for the review of SPACE code. The TRACE V5.0 patch 4 was used in this calculation. The calculation for the 6x6 reflood test (ATHER) was performed with the TRACE code. From the calculation results, the major behavior of the wall temperature could be predicted well. However, the further study will be needed to resolve the differences of quenching behaviors and to understand the reflood heat transfer model of TRACE code

  3. Performance analyses and tests on the KAERI devised spacer grids for PWRs

    Spacer grid which is one of the most important structural components in a pressurized light water reactor fuel assembly supports the fuel rods laterally and vertically. Based on design experiences and by scrutinizing the design features of foreign advanced nuclear fuels and the foreign patents of spacer grids, KAERI has devised its own spacer grid shapes and has acquired patents. In this study, a performance evaluation of two new spacer grid shapes devised by KAERI was carried out from the mechanical/structural and thermo hydraulic view points. And also performance evaluation of two commercial spacer grid shapes was carried out for the sake of a comparison. The comparisons included the spring characteristics, fuel rod vibration characteristics, fretting wear resistance, impact strength characteristics, CHF enhancement, and the pressure drop of the spacer grid shapes. The comparison results have shown that the performances of the new spacer grid shapes are better or at least not worse than those of the commercial spacer grid shapes. (author)

  4. KAERI/BNFL/COGEMA joint cooperation on environmentally friendly nuclear fuel cycle option study in Korea

    Through the project of 'KAERI/BNFL/COGEMA joint cooperation on environmentally friendly nuclear fuel cycle option study in Korea', the followings were studied. 1. Evaluation of environmental friendliness and of economic feasibility on the thermal neutron reactor type nuclear fuel cycle. 2. Evaluation of environmental friendliness on the future type nuclear fuel cycle. 3. Perspective of middle and long term electric power supply and of nuclear power plant constructionstes, development of device for the pretreatment of solid wastes, treatment of DU waste, analysis of the contaminated soil waste, induction of optimum conditions of coring from the 200L flexible waste forms, and long-term leaching test of 200L drum's waste form for the development of waste treatment and volume reduction technology, the characterization of waste formsenerated at KAERI. Therefore, radwastes are disposed of in a disposal site as solidified waste forms for its complete isolation from the human environment. The physicochemical properties of waste forms and the radionuclide concentration in waste forms should be evaluated for the radiological and structural safety of a disposal site, radionuclide type and solidification matrix, and it is difficult to carry out tests(for example, compressive strength, leaching rate, etc)with a full-scale waste forms. The waste classification and acceptance criteria is the result of technology development for characterization of waste and solidified waste forms. This treatment is carry out to low-cost and low-absorbed dose

  5. Chest Wall Thickness Measurements and the Dosimetric Implications for Male Radiation Workers at the KAERI

    Using ultrasound techniques, the Korea Atomic Energy Research Institute has measured chest wall thicknesses of a group of male workers at the Korea Atomic Energy Research Institute. A site-specific biometric equation has been developed for these workers. Chest wall thickness is an important modifier on lung counting efficiency. These data have been put into the perspective of the ICRP recommended dose limits for occupationally exposed workers:100 mSv in a 5-year period with a maximum of 50 mSv in any one year. For measured chest wall thicknesses of 1.9 cm to 4.1 cm and a 30 min counting time, the achievable MDAs for natural uranium in the KAERI lung counter vary from 5.75 mg to 11.28 mg. These values are close to, or even exceed, the predicted amounts of natural uranium that will remain in the lung (absorption type M and S) after an intake equal to the Annual Limit on Intake corresponding to a committed dose of 20 mSv. This paper shows that the KAERI lung counter probably cannot detect an intake of Type S natural uranium in a worker with a chest wall thickness equal to the average value (2.7 cm) under routine counting conditions

  6. Calculation analysis of Wims/D4-Batan-2DIFF neutron spectrum on RSG-GAS with cadmium ratio

    The calculation analysis of WIMS/D4-BATAN-2DIFF neutron spectrum was performed by comparison the calculation result of cadmium ratio with the experiment result on CIP, IP2, IP3 and IP4 irradiation positions of RSG GAS tenth core. The foils of Au, Mn and Co were used for determination of the measured and calculated cadmium ratios. Spectrum calculation was done in 69 energy group with 541 energy group (till 10 MeV) cross section of foil absorption reaction. The difference values between cadmium ratio calculation and experiment result for all cases were in interval of 11.0%-26.3% which are out of measurement deviation range. From these result, it concluded that the use of WIM /D4 in generating group constant is not sufficient to obtain the neutron spectrum, especially for non-fuel region

  7. Fabrication test of an engineering model cryo-sorption pump of KAERI test stand for KSTAR NBI

    The neutral beam injection system for KSTAR tokamak requires a pumping speed of > 2 x 103 m3/s to evacuate hydrogen/deuterium gases in the beam line chamber. In order to develop the KSTAR NBI system in KAERI test stand, that does not have a liquid He plant for the cryo-condensation pump, the cryo-sorption pump is being developed. An engineering model cryo-sorption pump, that will be a module of the pump for KAERI test stand, are designed, fabricated, and tested. The basic concept of the design is to obtain a maximum pumping speed with one refrigerator and minimum depth. The measured pumping speed of the engineering model is 80 m3/s for hydrogen at the panel temperature of 12 K. This pump will be used for the KAERI NBI Test Stand. (author)

  8. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

  9. Development of a 14.5 GHz electron cyclotron resonance ion source at KAERI

    A 14.5 GHz electron cyclotron resonance ion source has been designed and fabricated at KAERI (Korea Atomic Energy Research Institute) to produce multi-charged ion beams (particularly C6+ ion beams) for medical applications. The magnet system has normal conductor solenoid coils and a permanent magnet hexapole. A welded tube with aluminium and stainless steel is used as an ECR plasma chamber to improve the production of secondary electrons. A klystron supplies microwave energy to the plasma. A movable beam extractor with an 8 mm aperture covers various ion species and charge numbers of the beam. Fabrication and initial experimental results on ECR plasma and beam extraction are discussed in this paper.

  10. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    Park, Cheol; Lee, B. C.; Chae, H. T.; Kim, H.; Lee, C. S.; Choi, C. O.; Jun, B. J. [KAERI, Taejon (Korea, Republic of); Vien, Luong Ba; Dien, Nguyen Nhi [Vietnam Atomic Energy Commission, Hanoi (Viet Nam)

    2004-05-01

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon.

  11. Progress in the KAERI high energy nuclear data library : proton-induced neutron emission spectra

    Proton-induced neutron yields and emission spectra up to a few hundreds MeV are important nuclear data in the particle transport of the accelerator-driven system (ADS) and in the space shielding for trapped protons and solar energetic particle events. Within the framework of KAERI high energy nuclear data library evaluation, energy-angle spectra of secondary neutrons produced from the proton-induced neutron production reaction, (p, xn), of C-12, Al-27, Fe-56, and Pb-208 for energies below 400 MeV are evaluated based upon model calculations, guided and benchmarked by existing experimental data. Theoretical calculations were performed with the optical model analysis for the direct reactions and transmission coefficients, Hauser-Feshbach model for the equilibrium emission, and the exciton model for the preequilibrium emission, using the ECIS-GNASH code system. (author)

  12. A Study on the Management of Intellectual Property for the Potential Markets of KAERI

    The intellectual property law of the Republic of South Africa is similar to that of Korea except for a few regulations. In Republic of South Africa, the rights of joint inventor are limited, there is no request for examination, and the allowance of patent is generally determined within 18 months from the application date. Risky patents or applications are not found in Republic of South Africa. However, KAERI needs ceaselessly to search and investigate patents or patent applications in Republic of South Africa. Finally, we propose to build a patent management team within an operation division to respond swiftly to possible market changes. The operation-oriented patent management team will efficiently secure competitive patents and effectively realize a profit from the competitive patents

  13. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon

  14. A neutron guide installation status and its first performance test result at KAERI

    Cho, S. J.; Cho, Y. G.; Lee, C. H.; Lee, K. H.; Kim, K. P.

    2011-04-01

    A neutron guide system that includes neutron guides, a main shutter, and a vacuum system was successfully installed at the HANARO research reactor of the Korea Atomic Energy Research Institute (KAERI) last year, and is now operating with 5 cold neutron instruments. The neutron flux and spectrum were measured by using gold wire and a disc chopper. The total measured neutron fluxes for various position are about 10-25% lower than the calculated fluxes, which is probably caused by neutron guide misalignment, larger gap between neutron guides, low reflectivity, imperfect cold neutron source data, and so on. But the measured neutron fluxes of the neutron guides are very high. The status of the neutron guide installation and its first performance test result is described in this paper.

  15. The features and solution methodologies of the KAERI nuclear design code MASTER

    KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER Uncertainty Topical Report which includes global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations was transmitted in June 1996 as part of a license agreement to Korea Institute of Nuclear Safety (KINS). The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification and validation results are in details presented in the separate paper. (author)

  16. Plant application uncertainty evaluation of LBLOCA analysis using RELAP5/MOD3/KAERI

    Lee, Sang Yong; Chung, Bub Dong; Hwang, Tae Suk; Lee, Guy Hyung; Chang, Byung Hoon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the revised ECCS rule has been developed and this report describes the application of new REM to large break LOCA. A computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/ MOD3.1 was used as the best estimated code for the analysis and Kori unit 3 and 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by using Monte-Carlo sampler for each response surface. This study shows that plant application uncertainty can be quantified and demonstrates the applicability of the new realistic evaluation methodology. (Author) 29 refs., 40 figs., 8 tabs.

  17. Visualization system of the KAERI-DySCo for dynamic simulation of the VHTR-SI process

    The Korea Atomic Energy Research Institute-Dynamic Simulation Code (KAERI-DySCo) based on the Visual C++ which is an integration application software to analyze the dynamic behavior of the Sulfur Iodine (SI) process, which is a nuclear hydrogen process that is coupled to a Very High Temperature Gas Cooled Reactor (VHTR) through an Intermediate Heat Exchanger (IHX) generates a large number of raw data during its execution process. The generated raw data include various types of data such as thermodynamics data, mole flow rates of input and output material and their temperature and pressure information. In order to effectively classify and monitor the generated raw data, data visualization is required. As tools of the data visualization, the Chart FX and the Spread 7.0, which are commercial components, have been used; this has been used to process database of the raw data in the form of numerical value, and that has been used to present the raw data in the form of a graphical chart. The Chart FX based on the NET platform is embedded in the KAERI-DySCo in form of a Dynamic Link Library (DLL) by using a Windows Forms Control Hosting method. On the other hand, the Spread 7.0 is embedded in the KAERI-DySCo by using an ActiveX control method. As results of the data visualization, a main window of the KAERI-DySCo and its sub window have been introduced. A start-up dynamic simulation result of a HIx distillation column by using the KAERI-DySCo has been also introduced as an example of data visualization output. (author)

  18. Analysis of the Rowlands uranium oxide pin-cell benchmark with an updated WIMS-D library

    The Rowlands uranium oxide light water reactor pin-cell numerical benchmark results from the literature were analysed to obtain a self consistent set which can be used as reference. The materials relevant to the benchmark from the JEF-2.2 evaluated nuclear data file were processed with the NJOY code and the WIMS-D multigroup library was updated. An input for WIMSD-5A was prepared. Integral parameters, which include reaction rates and multiplication factors for the pin cell at different temperatures, moderator density and leakage were calculated. The results were compared to the previously defined reference values

  19. Validation of nuclear data for heavy water reactor lattices using WIMS and WIMKAL-88 nuclear data libraries

    Integral measurements of various types provide valuable data to assess the adequacy of the cross sections used in predicting the nuclear characteristics of reactors. In this context measurements of reactivity, relative reaction rates and neutron balance assume fundamental importance. We have analysed these parameters for heavy water moderated systems by using WIMS and WIMKAL-88 cross section libraries both of which have 69 energy groups. The analysis has been carried out by the lattice analysis code CLUB. It employs a method based on combination of interface current formalism and collision probability (CP) method. 6 refs, 1 tab

  20. Three-dimensional mechanical stability analysis for the underground research tunnel in KAERI

    Kwon, S. K.; Kim, K. S.; Park, J. W.; Joe, W. J

    2004-01-01

    For the disposal of high-level radioactive waste, it is required to develop a disposal concept. The geological disposal research group in KAERI had developed a reference disposal concept and now is studying for developing a Korean disposal concept. In order to develop a Korean disposal concept, the validation of disposal concept and performance safety analysis are essential. For the validation of disposal concept, it is necessary to construct an underground research laboratory. For doing that, the research team for the validation of disposal concept studies for constructing an underground research facility in KAERI. In order to design the underground research facility with computer simulation, disposal concept, rock characteristics, and topology should be considered. In this study, the reference disposal concept, which is necessary to be considered in the design of underground research tunnel, will be introduced first. After then, the important factors related to the underground research tunnel will be discussed. In the case of the site, where the underground tunnel are expected to be located, the surface topology is varying with thick weathered zone. In order to make the research modules in deep location with limited tunnel excavation, it is recommended to excavate a declined access tunnel. This study is for investigating the influence of different geological conditions, tunnel slope, tunnel size, and sequential excavation. In this study, mechanical stability analysis for the conceptual design of the underground research tunnel for the validation of Korean disposal concept had been carried out. Investigation of the influence of important parameters on mechanical stability was performed. The results from this study will be utilized for the geological investigation at the site, where an underground research tunnel will be located, as well as the design of the tunnel. Important conclusions from the study are as followings: (1) If the underground research tunnel is

  1. The Treatment Procedure for a Volume Reduction of the Spent HEPA Filters in KAERI

    Spent filter wastes of about 2,200 units have been stored in the radioactive waste storage facility of the Korea Atomic Energy Research Institute since its operation. Among these spent filter wastes, a HEPA filter account for about 95 %. All these HEPA filter wastes generated at KAERI have been stored inside a poly bag in accordance with the original form without any treatment of them. Therefore, in order to secure a space in a radioactive waste storage facility approaching its saturation, it is necessary to treat them by a compaction in view of a radioactive waste treatment and storage, and finally to repack the compacted spent filters into a regular drum for sending them to a final disposal site. To do that, the spent HEPA filter wastes were classified according to their generation facility, their generation date and their surface dose rate by investigating the inventory of them. And also, a nuclide assessment of them was conducted by taking a representative sample at the spot of a high dose rate at the intake surface and the outlet surface of a spent HEPA filter without a dismantlement, before compacting them. At present, for the spent HEPA filter wastes after a radionuclide assessment, a compaction treatment of them is now being conducted by using the shaping and compacting equipment developed at KAERI. Thus, to put a HEPA filter with a hexahedral form of a 610(W) x 610(H) x 305(T) mm into a regular drum (DOT-17H) with an inner diameter of about 572 mm, a columnar shaping with a capacity of 15 tons was conducted. From this shaping, a shaped HEPA filter waste with a diameter of about 500 mm was directly put into a regular drum. And then, the compaction treatment of a shaped HEPA filter with a capacity of about 60 tons was conducted by vertically compacting it. As a result, a volume reduction rate of a spent HEPA filter waste by a shaping and compacting of it accounted for about 1/8 when compared to its original form. (authors)

  2. Testing of a JEF-1 based WIMS-D cross section library for migration area and k-infinity predictions for LWHCR lattices

    The cell code WIMSD4 is used for the analysis of PROTEUS-LWHCR experiments. A library for this code which is based on the European evaluation JEF-1 was produced at EIR using the Los Alamos NJOY system with its module WIMSR and the Canadian management code WILMA. In general, this library delivered more accurate eigenvalues and reaction rates than the WIMS-Standard and WIMS81 libraries did in comparison to experimental values from PROTEUS-LWHCR Cores 1-3. However, large discrepancies (up to about 10%) occured between calculated migration areas (M2). Additional investigations have been undertaken to clarify this problem, since theoretical M2-values are needed for deducing k-infinity in the experiments. This has been done in the context of calculations for a reference LWHCR test lattice. The following major reasons for these deviations were found. First, the self-scattering term in non-moderators (P0 matrix) in the JEF-1 library was not transport corrected. Second, Standard and JEF-1 libraries use infinite dilute cross sections for 238U, whereas the WIMS81 library uses fully shielded cross sections. Third, the standard library uses the 'row' formula for the transport correction, whereas the 'inflow' formula is applied in the case of JEF-1 and WIMS81 libraries. Lastly, oxygen and 238U scattering cross sections in the fast energy range are smaller in the case of the WIMS81 library. Differences in calculated k-infinity values between the currently used library and WIMS81 (up to 3%) come (in order of importance for the reference LWHCR lattice) mainly from resonance cross sections for 240Pu capture, 238U capture and 239Pu fission. Recommendations have been made for generating a new JEF-1 library using updated versions of WIMSR and WILMA. (author)

  3. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Kozier, K.S

    1999-05-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)

  4. Generation of handbook of multi-group cross sections of WIMS-D libraries by using the XnWlup2.0 software

    A project to prepare an exhaustive handbook of WIMS-D cross section libraries for thermal reactor applications comparing different WIMS-D compatible nuclear data libraries originating from various countries has been successfully designed. To meet the objectives of this project, a computer software package with graphical user interface for MS Windows has been developed at BARC, India. This article summarizes the salient features of this new software and presents significant improvements and extensions in relation to its first version [Ann Nucl Energ 29 (2002) 1735

  5. Analysis of climate change scenarios in an olive orchard microcatchment in Spain using the model WIMMED

    Guzmán, Enrique; Aguilar, Cristina; José Polo, María; Taguas, Encarnación V.

    2015-04-01

    Olive orchards constitute traditional systems in the Mediterranean Basin. In Andalusia, Southern Spain, more than 1.5Mha are dedicated to olive crop land use, which represent a production of 1Mt of olive oil per year. This is a strategic economic sector with environmental and social relevance. In the context of climate change in Andalusia, the Intergovernmental Panel on Climate Change has highlighted that an increase of temperatures and rainfall intensities as well as the reduction of cumulated rainfall might be expected. This may mean serious detrimental economic and environmental risks associated to floods and the reduction of available water resources which would be convenient to quantify. The objective of this work is to analyse the rainfall-runoff relationships in an olive orchard catchment by the application of the distributed hydrological model WIMMED (Herrero et al., 2009) simulating the effects of climate change, with a special emphasis on extreme events. Firstly, the model was calibrated and validated with 9 maximum annual events of a datasets from 2005-2012 obtained in an olive orchard catchment in Spain (Taguas et al., 2010). In this stage, only the saturated hydraulic conductivity and soil moisture in saturation were adjusted after a sensitivity analysis where 68 simulations were carried out. A good agreement was obtained between observed and simulated hydrographs. The mean errors and the root mean square errors were 0.18 mm and 2.19 mm for the calibration and 0.18 and 1.94 mm, for the validation. Finally, the catchment response to the increase of intensity and temperature and the reduction of cumulated rainfall were simulated for the maximum event of the series. The results showed a rise of 11% of the runoff coefficient quantifying the possible impact of climate change. REFERENCES Herrero J, Polo M., Moñino A., Losada MA (2009). An energy balance snowmelt model in a Mediterranean site. J. Hydrol. 371, pp. 98-107 Taguas EV, Peña A, Ayuso JL, Yuan Y

  6. Status of neutron beam facilities at HANARO and a thermal neutron guide project of KAERI

    After successful installation of cold neutron facilities at HANARO such as neutron guides, cold neutron source including cold neutron instruments, now 14 cold and thermal neutron spectrometers are operating, and 5 instruments are under commissioning. The neutron guides with complicated shapes placed in the beam plug and the main shutter also in the curved part were delivered by a guide provider but the rest guides such as the guides in the guide bunker and the guide hall area were fabricated by KAERI. All the guides are coated with M=2 supermirror having different cross-sections and curvatures were operating with a high performance, where 10 cold neutron spectrometers will open to outside users. For a planning of a new project called ‘thermal guide facilities development’, the neutron guide system design started late last year, which was carried out to optimize the layout of the instruments and to calculate the neutron flux at sample position. At this meeting, the simulation results of the thermal neutron guide beam lines, status of in-house neutron guide development and specifications of some instruments will be presented.

  7. KAERI results on BFS-62 3A critical experiment analysis (Phase 5)

    This presentation is reporting on the KAERI's results on the BFS-62-3 A critical experiment analysis (Phase-5 ). In Phase-5 Model a homogeneous full core model is employed. Transport and diffusion calculations in R-Z model are carried out. R-Z model is used to test the transport effect and to provide the spectral weighting in generation of effective group XS. Based on the Hex-Z model, the sodium void reactivity effects (SVRE) are calculated for the voided regions. Cross Section Library KAFAX was based on the nuclear data file: JEF-2.2, ENDF-B/VI, prepared in MATXS format with multi-groups. Effective Cross Section(XS) Generation was done by cell XS calculation and group collapsing from 80 to 25 and 9 groups. BFS-62-3A Critical Experiment was modelled in R-Z and Hex-Z geometry. Results of Sodium Void Reactivity Effects include: Effect of Axial Mesh Size Change; Effect of Different Regionwise Spectrum Weighting in XS Collapsing; and differences caused by using JEF-2.2 and ENDF-B/VI libraries. A summary of KAERl Results is presented

  8. KAERI's technology development program of chemical decontamination for nuclear power reactors

    The activated corrosion products formed on the internal surface of primary coolant system of nuclear power plants can be removed by chemical decontamination. Dilute chemical decontamination method is widely used in consideration of keeping base metal integrity and producing relatively small amount of resulting radwastes. The application of chemical decontamination to PWRs is limited at present mainly to the channel heads of steam generators, but a growing necessity of entire NSSS decontamination is expected to accelerate the development and demonstration of the technology so that the commercial application of the technology will be realized in early 1990s. In Korea, nine nuclear power plants of PWR type except one will be available by 1989. The first chemical decontamination of the steam generator channel head of this nuclear power plant was done in 1984 by a foreign technology. KAERI's chemical decontamination technology development program funded by the Ministry of Science and Technology was started in 1983 to establish the technical guidelines and criteria and to obtain the technical self reliance. It is described. (Kako, I.)

  9. First lasing of the KAERI millimeter-wave free electron laser

    Lee, B.C.; Jeong, Y.U.; Cho, S.O. [Korea Atomic Energy Research Institute, Taejon (Korea, Democratic People`s Republic of)] [and others

    1995-12-31

    The millimeter-wave FEL program at KAERI aims at the generation of high-power CW laser beam with high efficiency at the wavelength of 3{approximately}10 mm for the application in plasma heating and in power beaming. In the first oscillation experiment, the FEL has lased at the wavelength of 10 mm with the pulsewidth of 10{approximately}30 {mu}s. The peak power is about 1 kW The FEL is driven by a recirculating electrostatic accelerator having tandem geometry. The energy and the current of the electron beam are 400 keV and 2 A, respectively. The FEL resonator is located in the high-voltage terminal and is composed of a helical undulator, two mesh mirrors, and a cylindrical waveguide. The parameters of the permanent-magnet helical undulator are : period = 32 mm, number of periods = 20, magnetic field = 1.3 kG. At present, with no axial guiding magnetic field only 15 % of the injected beam pass through the undulator. Transport ratio of the electron beam through the undulator is very sensitive to the injection parameters such as the diameter and the divergence of the electron beam Simulations show that, with unproved injection condition, the FEL can generate more than 50 kW of average power in CW operation. Details of the experiments, including the spectrum measurement and the recirculation of electron beam, are presented.

  10. Surface Decontamination of System Components in Uranium Conversion Plant at KAERI

    A chemical decontamination process using nitric acid solution was selected as in-situ technology for recycle or release with authorization of a large amount of metallic waste including process system components such as tanks, piping, etc., which is generated by dismantling a retired uranium conversion plant at Korea Atomic Energy Research Institute (KAERI). The applicability of nitric acid solution for surface decontamination of metallic wastes contaminated with uranium compounds was evaluated through the basic research on the dissolution of UO2 and ammonium uranyl carbonate (AUC) powder. Decontamination performance was verified by using the specimens contaminated with such uranium compounds as UO2 and AUC taken from the uranium conversion plant. Dissolution rate of UO2 powder is notably enhanced by the addition of H2O2 as an oxidant even in the condition of a low concentration of nitric acid and low temperature compared with those in a nitric acid solution without H2O2. AUC powders dissolve easily in nitric acid solutions until the solution pH attains about 2.5 ∼ 3. Above that solution pH, however, the uranium concentration in the solution is lowered drastically by precipitation as a form of U3(NH3)4O9 . 5H2O. Decontamination performance tests for the specimens contaminated with UO2 and AUC were quite successful with the application of decontamination conditions obtained through the basic studies on the dissolution of UO2 and AUC powders

  11. SNU-KAERI Degree and Research Center for Radiation Convergence Sciences

    In this study, we tried to establish and perform the demonstrative operation of the 'Degree and Research Center for Radiation Convergence Sciences' to raise the Korea's technology competitiveness. As results of this project we got the successful accomplishment as below: 1. Operation of Degree and Research Center for Radiation Convergence Sciences and establishment of expert researcher training system Ο Presentation of an efficient model for expert researcher training program through the operation of university-institute collaboration courses by combining of Graduate course and DRC system. Ο Radiation Convergence Sciences major is scheduled to be established in 2013 at SNU Graduate School of Convergence Science and Technology Ο A big project for research, education, and training of radiation convergence science is under planning 2. Establishment and conduction of joint research by organization of radiation convergence research consortium · Joint research was conducted in close connection with the research projects of researchers participating in this DRC project (44 articles published in journals, 6 patents applied, 88 papers presented in conferences) · The resources of the two organization (SNU and KAERI), such as research infrastructure (hightech equipment and etc), manpower (professor/researcher), and original technology and know how were utilized to conduct the joint research and to establish the collaboration system of the two organizations

  12. Rf System For The Industrial Linear Electron Accelerator At Kaeri (daejeon, Korea)

    Arbuzov, V S; Evtushenko, Yu A; Gorniker, E I; Kenjebulatov, E K; Kondakov, A A; Krutikhin, S A; Kurkin, G Ya; Motygin, S V; Osipov, V N; Petrov, V M; Pilan, Andrey M; Popov, A R; Shteinke, A M; Tribendis, A G

    2004-01-01

    Budker Institute of Nuclear Physics has developed and produced RF generators, feeder lines and a control system for an industrial linear electron accelerator at Korean Atomic Energy Research Institute (KAERI, Daejeon, Korea). The accelerator is based on two superconducting RF cavities produced by CERN. Design energy of the accelerator is 10 MeV and design beam current is 10 mA. A 2 MeV injector for the accelerator was made by BINP earlier. Two-channel RF system of the accelerator operates at the frequency of 352 MHz in CW mode. Each channel has two-stage tetrode amplifier with output power of 50 kW, 100 W transistor preamplifier and the control system. Both tetrode stages have identical design. TH571B tetrode tubes produced by THALES (France) are used. Output power of 45 kW per channel was reached in an equivalent resistive load. Now BINP continues development of the accelerator. The energy of 11 MeV and the beam current of 1.9 mA were achieved. The amplitude of accelerating voltage was 4.5 MV in each cavity,...

  13. Investigation for the Fossil Embryo using Neutron Tomography at HANARO, KAERI

    Neutron imaging technique is one of non-destructive method. It is similar to X-ray and g-ray methods in using the different attenuation characteristics depending on materials. However, there is great difference between them. The mass attenuation coefficients of X-ray and g- ray monotonically increase with the atomic number since they interact with electrons. Thus X-ray image method does not supply sufficient contrast between similar atomic numbers. On the other hand, that of thermal neutrons depends much on the nucleus not electrons. Especially thermal neutrons easily penetrate most of metals, while they are attenuated well by such materials as hydrogen, water, boron, gadolinium and cadmium. Because of these unique characteristics of neutron, neutron imaging technique has been utilized for NDT or researches for next power sources (fuel cell or Li-Ion battery). Recently, dinosaur egg was found at the Aptian. Albian Algui Ulaan Tsav site, Mongolia. In this study, we applied the neutron imaging technique to investigate dinosaur embryo at Neutron Radiography Facility of HANARO, KAERI

  14. Compton X-Ray Generation at the KAERI SC RF LINAC

    Park, S H; Jeong, Y U; Lee, B C; Lee, K

    2005-01-01

    The KAERI SC RF linac with one 352 MHz cryomodule is routinely operating at 10 MeV. The maximum accelerating gradient achieved so far is about 7.7 MV/m and is expected to increase up to 9 MV/m, if thermal loss and/or vibration instability is sufficiently suppressed. As a next step, we plan to generate Compton X-rays using external lasers at the straight section, just after the SC linac. This beamline will be relocated to downstream next to undulator beamline for a FEL, when the recirculating beamline is built. In this presentation, we estimate the parameters of Compton X-rays at a given system and suggest the new scheme to increase the flux, or to generate fs X-ray pulses using electron beams with a few tens ps pulse duration, using an intense ultra-short laser. We discussed a coherent condition for Relativistic Nonlinear Thomson Scattered (RNTS) radiation (or Nonlinear Compton Scattered radiation).

  15. Assessment of the KAERI 6*6 reflood experiment using the SPACE code

    Nuclear industries in Korea are developing the nuclear safety analysis code named SPACE (Safety and Performance Analysis Code) which is based on a multi-dimensional, two-fluid, three-field model for a licensing application of pressurized water reactors. A reflood heat transfer phenomena can be predicted with using a general wall heat transfer model or a separate reflood heat transfer model of the SPACE code based on a user option. The reflood heat transfer package takes into account the two-dimensional heat conduction effects near the quench fronts. This paper briefly introduces the heat transfer models of the SPACE code regarding the reflood heat transfer phenomena, and the preliminary assessment results against KAERI 6*6 reflood heat transfer experiments using the general film boiling and the reflood heat transfer models. The objectives of these assessments are to examine the preliminary prediction capabilities of the SPACE code against the reflood phenomena, and to suggest future directions for improvement. Both the general wall heat transfer and the reflood heat transfer models of the SPACE code predicts reasonably the wall temperature behavior and quenching time. However, for a high reflooding velocity, the SPACE code showed slightly earlier quenching than the experiment because of a faster water accumulation in the test section. Thus, physical models such as droplet entrainment, interfacial drag, and droplet diameter should be checked and improved for the high flooding rates. (authors)

  16. Intercomparison of analysis methods for seismically isolated nuclear structures (KAERI HLRB and CRIEPI Isolated Rigid Mass Mock-Up)

    The combined shear and compression behaviors of the KAERI HLRB made of MRPRA rubber and the shaking responses of the CRIEPI isolated rigid mass mock-up are analyzed. For FEM analyses of KAERI HLRB, three kinds of strain energy density functions of the ABAQUS program are used as constitutive law for rubber with hyperelastic characteristics. The analysis results are compared with test results, depending on the constitutive models. The simulation results for the shaking table tests of the CRIEPI rigid mass mock-up supported by scaled lead rubber bearings are obtained by ABAQUS time history analyses. In the analysis, the linear and bilinear hysterisis models simulating the behaviors of the rubber bearing are used. (author)

  17. Development of environmental sample analysis technique in KAERI. Bulk analysis and establishment of clean laboratory facility (CLASS)

    The development of analytical methods for environmental samples in Korea Atomic Energy Research Institute (KAERI) is discussed. An analysis scheme for environmental samples has been established with an MCICP-MS based bulk analysis with adopting UTEVA resin for chemical separation and a particle analysis using FTTIMS and SIMS. A clean laboratory facility called CLASS (class 100∼ class 1000) was also established in order to prevent any cross contamination of the samples. The amount of U and Pu in the process blank sample prepared in the CLASS facility was estimated as 20 pg and less than 0.005 pg, respectively. The control chart of the analytical performance for the uranium standard sample of 100 ppt (NBL U030) indicated that the analytical performance of KAERI in CLASS is within 5 % of the certified values. (author)

  18. Calculation of the Fast Flux Test Facility fuel pin tests with the WIMS-E and MCNP codes

    The Fuel Assembly Area (FAA) at the Fast Flux Test Facility site on the Hanford Site at Richland, Washington currently is being prepared to fabricate mixed oxide fuel (U, Pu) for the FFTF. Calculational tools are required to perform criticality safety analyses for various process locations and to establish safe limits for fissile material handling at the FAA. These codes require validation against experimental data appropriate for the compositions that will be handled. Critical array experiments performed by Bierman provide such data for mixed oxide fuel in the range Pu/(U+Pu) = 22 wt %, and with Pu-240 contents equal to 12 wt %. Both the Monte Carlo Neutron Photon (MCNP) and the Winfrith Improved Multigroup Scheme (WIMS-E) computer codes were used to calculate the neutron multiplication factor for explicit models of the various critical arrays. The W-CACTUS modules within the WIMS-E code system was used to calculate k∞ for the explicit array configuration, as well as few-group cross sections that were then used in a three-dimensional diffusion theory code for the calculation of keff for the finite array. 10 refs., 15 figs., 7 tabs

  19. Verification of the Bulk Analysis Procedure of Safeguards Environmental Samples by Thermal Ionization Mass Spectrometry in KAERI

    The Korea Atomic Energy Research Institute (KAERI) had been qualified as a member of the Network of Analytical Laboratories (NWAL) for bulk analysis on environmental samples (ES) in 2012. Recently a new clean facility had been constructed and opened in KAERI, which caused the validation issue as the analytical environment and the main analytical instrument had been changed after the qualification. This study is to verify the capability of KAERI to performed bulk analysis on environmental sample under the new analytical environment using thermal ionization mass spectrometry (TIMS). The verification of the quality assurance of the bulk analysis of environmental samples was performed by TIMS measurement of a simulated swipe sample. The analytical results for the determination of the isotopic ratios and the amount contents of nuclear materials in the simulated environmental samples were in good agreement with the certified values. Therefore, we believe that our laboratory can produce reliable results for the bulk analysis on environmental swipe samples performed in CLASS and contribute the analytical services as a member of NWAL

  20. Development of a web-based sorption database (KAERI-SDB) and application to the safety assessment of a radioactive waste disposal

    Sorption plays a key role in a retardation of radionuclide migration in various geological environments. Hence sorption of radionuclides onto geological media is one of the important factors for the safety assessment of radioactive waste disposal. A web-based radionuclide sorption database program named KAERI-SDB has been developed to provide a database for the sorption of radionuclides onto geological media at various geochemical conditions. The KAERI-SDB is designed to determine the distribution coefficient (Kd) of a radionuclide and evaluate sorption properties by easily accessing an internet web-site ( (http://sdb.kaeri.re.kr)). The KAERI-SDB provides a useful output and search result as a scatter plot chart or an index chart. The KAEI-SDB was designed to show the search results in a statistical way by representing the mean Kd value at 95% of confidence as a function of major geochemical indices. Several case studies were carried out to demonstrate the applicability of the KAERI-SDB and the result showed a successful applicability of the KAERI-SDB to various radionuclide sorption cases.

  1. Development of safeguards technology for lab-scale advanced fuel cycle facility at KAERI

    KAERI (Korea Atomic Energy Research Institute) has been developing the DUPIC (Direct Use of PWR spent fuel in CANDU) fuel cycle and ACP (Advanced Spent Fuel Conditioning Process) technology for the purpose of spent fuel management. A safeguards system has been applied to R and D process for fabricating DUPIC fuel directly with PWR spent fuel material. Safeguards issues to be resolved were identified in the areas such as international cooperation on handling foreign origin nuclear material, technology development of operator's measurement system of bulk handling process of spent fuel material, and built-in C/S system for independent verification of material flow. All those safeguards issues have been finally resolved. The lab-scale DUPIC facility (DFDF) safeguards system was successfully established under the international cooperation program. The ACP has been under development at KAERI since 1997 to tackle the problem of the accumulation of the spent fuel. The concept is to convert the spent oxide fuel into a metallic form in a high temperature molten salt in order to reduce the heat power, volume, and radioactivity of the spent fuel. The main objective of the ACP is to treat the PWR spent fuel for a long-term storage and eventual disposal in a proliferation resistant and cost effective way. Moreover, the electrolytic reduction method of the ACP can contribute to the innovative nuclear energy system as a key technology for the preparation of the metallic fuel. Since the inactive tests of the ACP have been successfully implemented to confirm the validity of the electrolytic reduction technology, a lab-scale hot test will be undertaken in the ACP facility (ACPF) to validate the concept. Based on the results of a safeguards implementation at DFDF hot cell, the reference safeguards design conditions are established for the ACPF. Basically, the nuclear material accounting will be performed by ASNC (ACP Safeguards Neutron Counter), which is the same concept as the

  2. R and D strategy on remote response technology for emergency situations of nuclear facilities in KAERI

    Jeong, Kyung Min; Cho, Jae Wan; Choi, Young Soo; Eom, Heung Seup; Seo, Yong Chil; Shin, Hoch Ul; Lee, Sung Uk; Kim, Chang Hoi; Jeong, Seung Ho; Kim, Seung Ho [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    equipped with various tools allowing it to scour surfaces, scoop samples and vacuum sludge. To perform cleanup tasks, they built Workhorse that featured system redundancy and had a boom extend able to reach high places, but it was never used because it had too many complexities and to clean and fix. While remote robotics technology has proven to remove the human from the radioactive environment, it is also difficult to make it useful because it may requires skill about remote control and obtaining remote situation awareness regardless of the actual task. The efficiency of the human robot interaction is very important to obtain the overall goal for the emergency response in timely manner. It would be a bottle neck to apply the robotic technology for carrying the emergency response in NPP. Simple remote operation schemes are not adequate, more intelligent autonomous operation schemes are required to enhance the effectiveness of robots for the emergency response. KAERI has been developing various robotic systems for nuclear power plants over than 20 years after the Chernobyl accident. But the majority of the developed robotic systems is for the inspection and maintenance of nuclear power plants during their outage periods. Based on the lessons learned from the Fukushima accident, KAERI has planned R and D projects for developing remote response technologies really applicable in emergency situations of nuclear facilities. This paper presents the R and D strategy to achieve real usability and the purpose and research activity plans of on going three projects derived from the strategy.

  3. R and D strategy on remote response technology for emergency situations of nuclear facilities in KAERI

    it to scour surfaces, scoop samples and vacuum sludge. To perform cleanup tasks, they built Workhorse that featured system redundancy and had a boom extend able to reach high places, but it was never used because it had too many complexities and to clean and fix. While remote robotics technology has proven to remove the human from the radioactive environment, it is also difficult to make it useful because it may requires skill about remote control and obtaining remote situation awareness regardless of the actual task. The efficiency of the human robot interaction is very important to obtain the overall goal for the emergency response in timely manner. It would be a bottle neck to apply the robotic technology for carrying the emergency response in NPP. Simple remote operation schemes are not adequate, more intelligent autonomous operation schemes are required to enhance the effectiveness of robots for the emergency response. KAERI has been developing various robotic systems for nuclear power plants over than 20 years after the Chernobyl accident. But the majority of the developed robotic systems is for the inspection and maintenance of nuclear power plants during their outage periods. Based on the lessons learned from the Fukushima accident, KAERI has planned R and D projects for developing remote response technologies really applicable in emergency situations of nuclear facilities. This paper presents the R and D strategy to achieve real usability and the purpose and research activity plans of on going three projects derived from the strategy

  4. Design of an engineering scale off-gas trapping system at KAERI

    KAERI has been developing a high temperature voloxidation process as a head-end process for pyroprocessing technology. This process may remove volatile fission products (Kr, H-3, C-14, I-129 etc.) and other semi-volatiles (Cs, Tc, Te, Mo etc.) that are problematic in the main pyroprocess. Engineering off-gas treatment system was designed to recover the primary semi-volatile products (Cs, Tc, Te, Mo, I, etc.) released from simulated reagents during the high temperature voloxidation process. The off-gas trapping system will trap selectively gaseous nuclides evolved from high temperature voloxidation process, this will also reduce the high level waste due to the separation of Cs. This paper describes design of the off-gas treatment systems for high temperature voloxidation process. Design of an engineering-scale semi-volatile trapping system of 50 kg-SF/batch was done. The gaseous waste arising for off-gas trapping system was estimated considering the release rate of each target fission product. Each unit process in the trapping system is arranged to effectively remove the species of interest by considering the chemical properties of the target fission products to be trapped. Cs and Rb are trapped on a fly ash filter at around 900degC. Tc(Re), Te, Se, and Mo on a calcium filter are trapped at about 700degC, and I on a AgX is trapped at about 250degC. Off-gas trapping system was designed based on the design requirements such as trapping media, fission products to be trapped, design temperatures of the trapping units, optimum operation temperatures and specifications of the filters. Off-gas trapping system was also designed based on the design requirements such as remote operability, accessibility, and flexibility of instrument, separability of trapping basket, material of instrument. (author)

  5. Thermal, hydraulic, and mechanical initial conditions around KAERI Underground Research Tunnel

    In KAERI underground research tunnel(KURT) various in situ experiments for the investigation of thermal, mechanical, hydraulic, and chemical behaviours related to the validation of high-level radioactive waste disposal system are carrying out. In this study, the geological characteristics, thermal, hydraulic, and mechanical(THM) properties of the rock mass, and groundwater level analyzed and derived relationship between the THM properties and depth. From this study, it was found that the THM properties varies with depth Z and many properties could be expressed well with an equation type, a+b/Zc. The calculated rock thermal properties were 3∼7% higher than the measurement and the difference was relatively higher in dry rock. With empirical equations and measured air and tunnel wall temperatures, it was also possible to estimate that the seasonal temperature variations at 5m and 10m distance from tunnel wall were 3 .deg. C and 0,75 .deg. C, respectively. The thermal-hydraulic-mechanical initial conditions around KURT derived from this study will be utilized for the selection of location and the design for various in situ experiments at KURT. Those will be also used as fundamental data for the analysis of the results from the in situ experiments. The understanding of the THM initial conditions will be useful for the investigation of low and intermediate level repository as well the site selection and system design for a temporary storage facility and a high-level radioactive waste repository. This will also be applied to the design of underground caverns for various purposes and the analysis of in situ measurements at underground excavations

  6. The Neutronic And Power Distribution Calculations For Triga 2 MW Reactor Using WIMS-D/4 And Citation Codes

    . The neutronic calculation has been carried out for TRIGA 2 MW reactor. These included criticality flux and power distributions. Computer code Citation which solves 7-groups, 3-dimensional hexagonal geometry has been used. The multi groups-cross-section is generated by the WIMS-D/4 code.This 7-group-39x39x38-mesh-points problem takes about 90 minutes on the Pentium-133 MHz PC. The calculation of the initial core of TRIGA 2 MW reactor shows that the excess reactivity of the core is 7,8% and the thermal fluxes in the irradiation positions are between 1.0-2.9*1013n cm-2s-1. The results are about 10% deviate from those calculated by General Atomics. In the initial core, the highest power is produced in the C-9 position. The fuel element in this position produces 30.7 k W thermal power

  7. Modeling TRIGA reactor pulses using the STAR 3D nodal kinetics and WIMS-D4 codes

    A detailed three-dimensional (3D) time-dependent STAR nodal kinetics model coupled to a one-dimensional (1D) thermal-hydraulics WIGL model has been developed to describe and benchmark the peak power and pulse behavior of the Penn State University (PSU) Breazeale TRIGA reactor. Different core loading patterns were used for several TRIGA pulse tests with different reactivity insertion worths (1.5 dollar, 2.0 dollar, 2.5 dollar). The STAR nodal kinetics code and TRIGA model adequately simulates TRIGA pulses when group constants are generated from physics codes (i.e., WIMS-D4) that can accurately model the TRIGA uranium-zirconium-hydride fuel

  8. Characteristics of HEPA filter waste compactor developed by KAERI and comparison with Japan's and U. S. A's compactors

    The HEPA filter waste (Hexahedron, 610 x 610 x 292 mm, 20 Kg), which is used for the ventilation system in the nuclear industries, has relatively large volume to compare with it's weight. Due to the large volume of HEPA filter waste, it needs the large storage and/or disposal space and managing cost during the long period of storage and management. Many countries use the compactor to reduce the volume of HEPA filter waste. On this paper we introduce the new type compactor developed by KAERI with the characteristic comparison to the Japan's and U. S. A's compactors

  9. KAERI Activities on the Cooling Performance of Ex-vessel Core Catcher

    Ha, Kwang Soon; Park, Rae Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Wi, Kyung Jin [Chungnam National University, Daejeon (Korea, Republic of); Thanh, Thuy Nguyen Thi [University of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    the integrity of the ex-vessel core catcher system. KAERI has performed various researches to validate the cooling performance of an ex-vessel core catcher. First, a scaling analysis was performed to design the scaled-down experimental facility and maintain the characteristics of the real natural circulation flow by solving the natural circulation flow loop equation for the cooling channel in the ex-vessel core catcher. Second, boiling-induced natural circulation flow experiments in the cooling channels of the ex-vessel core catcher were investigated. Finally, a new correlation was developed to estimate the natural circulation mass flow rate with the inclined downward facing heating surface. KAERI has performed various researches to validate the cooling performance of the ex-vessel core catcher. First, the scaling analysis was performed to design the scaled-down experimental facility and maintain the characteristics of the real natural circulation flow by solving the natural circulation flow loop equation for the cooling channel in the ex-vessel core catcher. Secondly, boiling-induced natural circulation flow experiments in the cooling channels of the ex-vessel core catcher were investigated. And finally, a new correlation has been developed to estimate the natural circulation mass flow rate with the inclined downward facing heating surface. The circulation mass flux, the quality, and void fraction at the exit of the cooling channel in the experimental facility with the selected orifice coincided exactly with the prototypic core catcher system even though the different void fraction models were applied. In conclusion, a scaling analysis methodology for the natural circulation flow loop was proposed and successfully verified. In the experiment, the effect of the water level, heat flux, heat flux distribution, core catcher vertical side-wall length, and coolant temperature were studied. A natural circulation test was carried out in two stages, one with freely increasing

  10. Graphic Design and the Edges of Common Sense. Thinking about design through the conflicting approaches of Wim Crouwel and Jan van Toorn

    Kortteinen, Tuomas

    2015-01-01

    This thesis is an attempt to define two contrasting approaches to graphic design practice and explain the reasoning behind their differing perspectives. The starting point and main subject matter is the public debate between Wim Crouwel and Jan van Toorn that occurred in 1972 and continued in different forms for the following decade. The point of departure for the thesis is that public debates in general, and this one in particular, offer an invaluable view of the implicit assumptions th...

  11. Recent Research Status on the Microbes in the Radioactive Waste Disposal and Identification of Aerobic Microbes in a Groundwater Sampled from the KAERI Underground Research Tunnel(KURT)

    In this report, a comprehensive review on the research results and status for the various effects of microbes in the radioactive waste disposal including definition and classification of microbes, and researches related with the waste containers, engineered barriers, natural barriers, natural analogue studies, and radionuclide migration and retardation. Cultivation, isolation, and classification of aerobic microbes found in a groundwater sampled from the KAERI Underground Research Tunnel (KURT) located in the KAERI site have carried out and over 20 microbes were found to be present in the groundwater. Microbial identification by a 16S rDNA genetic analysis of the selected major 10 aerobic microbes was performed and the identified microbes were characterized

  12. Energy response characteristics of several neutron measuring devices determined by using the scattered neutron calibration fields of KAERI

    The energy response characteristics of several neutron measuring devices used popularly for radiation protection purpose were determined under the simulated neutron calibration fields which was produced by using the radionuclide neutron sources and the shadow objects to scatter and to moderate the fast neutrons emitted from the source. The simulated neutron calibration fields for the calibration of personal dosemeters and survey meters were constructed in the Radiation Calibration Laboratory (RCL) of Korea Atomic Energy Research Institute (KAERI). The radionuclide sources of 252Cf and 241AmBe were used for producing the neutron calibration fields with little different from the method recommended by ISO. The calibration points of interest were behind the shadow objects and the concrete wall in the irradiation room. In order to characterize the neutron calibration fields at the point of test, the spectral neutron fluence rate was determined by means of the Bonner Multi-sphere Spectrometry System (BMSS) and the measured spectra unfolded using the BUNKI code. The dosimetric quantities were derived from the unfolded spectra and used as the reference value to determine the response of each detector. Five kinds of the active detector (three for detector with heavy moderator, one for detector having two spherical tubes with different size, and a TEPC, Tissue Equivalent Proportional Counter) and a TLD as the passive detector were used in this study. The spectral mean energy at the reference calibration points ranged from 0.1 MeV to 3.44 MeV and the dose rate from 0.12 mSv/hr to 4.62 mSv/hr. This paper shows that the big difference, more than four times in case of TLD, in the response of detector with the neutron field spectra should be corrected when the detector is used for monitoring and the dosimetric data of KAERI 's scattered neutron calibration fields. (author)

  13. Assessment of MARS 2.0 for direct DVI bypass during LBLOCA reflood using KAERI air-water DVI tests

    MARS code has been assessed for the direct ECC (Emergency Core Cooling) bypass that occurs during LBLOCA reflood of KNGR (Korean Next Generation Reactor) using the KAERI air-water DVI (Direct Vessel Injection) tests that are 1/50 scale-down tests simulating the LBLOCA reflood of KNGR. Assessment matrix is selected for the single and double DVI configurations with typical LBLOCA reflood conditions, that is, DVI injection velocity of 1.0 ∼ 1.6 m/sec and air injection velocity of 20 ∼ 35 m/sec. First, the MARS calculation is adjusted to match the DVI film distribution with the 1/50 scale test results, then the code assessments are carried out for the selected direct DVI bypass tests using the adjusted DVI film distribution. From the assessments, it has been found that the MARS is capable of predicting the direct DVI bypass phenomena as well as the multi-dimensional thermal hydraulics in the downcomer

  14. Investigation of the development and the effect of an excavation damaged zone at KAERI underground research tunnel

    Kwon, S.; Cho, W. J

    2008-01-15

    The understanding of the long term behavior of rock around an underground radioactive waste repository is essential for the safe design and operation of the repository and for assuring the safety and technical feasibility of geological disposal concept. The investigation of the influence of EDZ on rock mass behavior is important for the long term stability, economy, and safety points of view. In the case of underground repository, which requires high level safety criteria, the accurate prediction of the long and short term mechanical, hydraulic, and thermal behaviors is especially important. In this study, the size and characteristics of EDZ developed during the construction of the KAERI underground research tunnel, which was constructed by controlled blasting, were investigated using various methods. Goodman jack test for measuring deformation modulus, Georadar, rock core observation, MPBX, and stressmeter were carried out at KURT. The rock cores from the boreholes were tested in laboratory for estimating the EDZ size. Empirical and theoretical equations were also used for the prediction of EDZ. The results from laboratory and in situ tests were used in three-dimensional hydro-mechanical and thermo-mechanical analysis for the evaluation of the EDZ effect. The understanding of EDZ size and the property changes in EDZ from in situ and laboratory tests will be used for the planning, design, and analysis of in situ experiments in KURT. The results from the EDZ study will be helpful for the system design as well as safety analysis of a radioactive repository.

  15. Synthesis of the turbulent mixing in a rod bundle with vaned spacer grids based on the OECD-KAERI CFD benchmark exercise

    Highlights: • OECD/KAERI international CFD benchmark exercise was operated by KAERI. • The purpose is to validate relevant CFD codes based on the MATiS-H experiments. • Blind calculation results were synthesized in terms of mean velocity and RMS. • Quality of control volume rather than the number of it was emphasized. • Major findings were followed OECD/NEA CSNI report. - Abstract: The second international CFD benchmark exercise on turbulent mixing in a rod bundle has been launched by OECD/NEA, to validate relevant CFD (Computational Fluid Dynamics) codes and develop problem-specific Best Practice Guidelines (BPG) based on the KAERI (Korea Atomic Energy Research Institute) MATiS-H experiments on the turbulent mixing in a 5 × 5 rod array having two different types of vaned spacer grids: split and swirl types. For this 2nd international benchmark exercise (IBE-2), the MATiS-H testing provided a unique set of experimental data such as axial and lateral velocity components, turbulent intensity, and vorticity information. Blind CFD calculation results were submitted by twenty-five (25) participants to KAERI, who is the host organization of the IBE-2, and then analyzed and synthesized by comparing them with the MATiS-H data. Based on the synthesis of the results from both the experiments and blind CFD calculations for the IBE-2, and also by comparing with the IBE-1 benchmark exercise on the mixing in a T-junction, useful information for simulating this kind of complicated physical problem in a rod bundle was obtained. And some additional Best Practice Guidelines (BPG) are newly proposed. A summary of the synthesis results obtained in the IBE-2 is presented in this paper

  16. A study on the impacts of R and D expenditures of Korea Atomic Energy Research Institute on the national economy. A study on the contribution made by KAERI to the national economy

    This study analyzes the contribution of KAERI's R and D to the national economy. As a case study, the study also analyzes the economic impacts which KAERI's capacity of independent system design contributes to the national economy through localization of KSNP. The research method is Input-output methods which are frequently employed in evaluating economic impacts of R and D in both domestic and foreign academic areas

  17. The Potential and Beneficial Use of Weigh-In-Motion (WIM) Systems Integrated with Radio Frequency Identification (RFID) Systems for Characterizing Disposal of Waste Debris to Optimize the Waste Shipping Process

    Abercrombie, Robert K [ORNL; Buckner Jr, Dooley [ORNL; Newton, David D [ORNL

    2010-01-01

    The Oak Ridge National Laboratory (ORNL) Weigh-In-Motion (WIM) system provides a portable and/or semi-portable means of accurately weighing vehicles and its cargo as each vehicle crosses the scales (while in motion), and determining (1) axle weights and (2) axle spacing for vehicles (for determination of Bridge Formula compliance), (3) total vehicle/cargo weight and (4) longitudinal center of gravity (for safety considerations). The WIM system can also weigh the above statically. Because of the automated nature of the WIM system, it eliminates the introduction of human errors caused by manual computations and data entry, adverse weather conditions, and stress. Individual vehicles can be weighed continuously at low speeds (approximately 3-10 mph) and at intervals of less than one minute. The ORNL WIM system operates and is integrated into the Bethel Jacobs Company Transportation Management and Information System (TMIS, a Radio-Frequency Identification [RFID] enabled information system). The integrated process is as follows: Truck Identification Number and Tare Weight are programmed into a RFID Tag. Handheld RFID devices interact with the RFID Tag, and Electronic Shipping Document is written to the RFID Tag. The RFID tag read by an RFID tower identifies the vehicle and its associated cargo, the specific manifest of radioactive debris for the uniquely identified vehicle. The weight of the cargo (in this case waste debris) is calculated from total vehicle weight information supplied from WIM to TMIS and is further processed into the Information System and kept for historical and archival purposes. The assembled data is the further process in downstream information systems where waste coordination activities at the Y-12 Environmental Management Waste Management Facility (EMWMF) are written to RFID Tag. All cycle time information is monitored by Transportation Operations and Security personnel.

  18. The Potential and Beneficial Use of Weigh-In-Motion (WIM) Systems Integrated with Radio Frequency Identification (RFID) Systems for Characterizing Disposal of Waste Debris to Optimize the Waste Shipping Process

    The Oak Ridge National Laboratory (ORNL) Weigh-In-Motion (WIM) system provides a portable and/or semi-portable means of accurately weighing vehicles and its cargo as each vehicle crosses the scales (while in motion), and determining (1) axle weights and (2) axle spacing for vehicles (for determination of Bridge Formula compliance), (3) total vehicle/cargo weight and (4) longitudinal center of gravity (for safety considerations). The WIM system can also weigh the above statically. Because of the automated nature of the WIM system, it eliminates the introduction of human errors caused by manual computations and data entry, adverse weather conditions, and stress. Individual vehicles can be weighed continuously at low speeds (approximately 3-10 mph) and at intervals of less than one minute. The ORNL WIM system operates and is integrated into the Bethel Jacobs Company Transportation Management and Information System (TMIS, a Radio-Frequency Identification (RFID) enabled information system). The integrated process is as follows: Truck Identification Number and Tare Weight are programmed into a RFID Tag. Handheld RFID devices interact with the RFID Tag, and Electronic Shipping Document is written to the RFID Tag. The RFID tag read by an RFID tower identifies the vehicle and its associated cargo, the specific manifest of radioactive debris for the uniquely identified vehicle. The weight of the cargo (in this case waste debris) is calculated from total vehicle weight information supplied from WIM to TMIS and is further processed into the Information System and kept for historical and archival purposes. The assembled data is the further process in downstream information systems where waste coordination activities at the Y-12 Environmental Management Waste Management Facility (EMWMF) are written to RFID Tag. All cycle time information is monitored by Transportation Operations and Security personnel.

  19. PREP-PWR-1.0: a WIMS-D/4 pre-processor code for the generation of data for PWR fuel assemblies

    The PREP-PWR-1.0 computer code is a substantially modified version of the PREWIM code which formed part of the original MARIA System (Report J.E.N. 543). PREP-PWR-1.0 is a comprehensive pre-processor code which generates input data for the WIMS-D/4.1 code (Report PEL 294) for PWR fuel assemblies, with or without control and burnable poison rods. This data is generated at various base and off-base conditions. The overall cross section generation methodology is described, followed by a brief overview of the model. Aspects of the base/off-base calculational scheme are outlined. Additional features of the code are described while the input data format of PREP-PWR-1.0 is listed. The sample problems and suggestions for further improvements to the code are also described. 2 figs., 2 tabs., 12 refs

  20. Update of WIMS-D libraries using JENDL-3.2, ENDF/B-VI.5 and JEF-2.2

    The WIMS-D5 Libraries based on JENDL-3.2, ENDF/B-VI.5, and JEF-2.2 have been prepared and are being tested against the benchmark problems. Several sensitivity calculations for stabililty confirmation of the libraries were carried out such as the fission spectrum dependency, the self shielding effects of the elastic scattering cross sections, the self shielding effects of Pu -240 and Pu -242 capture cross sections below 4.0 eV, etc. The results of benchmark calculations with the libraries based on JENDL-3.2, ENDF/B-VI.5, JEF-2.2, and the '1986 library were intercompared. The predictions of criticalities and isotopic compositions with the updated libraries show good agreements with the measurements or the reference results. The multiplication factors with the library based on JENDL-3.2 are slightly higher than those of ENDF/B-VI.5 and JEF-2.2

  1. Turbulent mixing in a rod bundle with vaned spacer grids: OECD/NEA–KAERI CFD benchmark exercise test

    Highlights: • Detailed velocity profiles have been examined in a rod bundle with mixing spacer grids. • Mixing characteristics strongly depend on the type of the mixing vane on a spacer grid. • The swirl in subchannels is elliptic and the cross-flow in gaps is vigorous in the split-type. • Swirl-type vanes generate a circular swirl in a subchannel and a weak cross-flow in gaps. • Mixing performance is superior in the case of the split-type compared to the swirl-type. - Abstract: An experimental study titled the 2nd International Benchmark Exercise (IBE-2) has been conducted to provide high-precision data of detailed turbulent flow mixing in a rod bundle for validating the CFD codes being used widely in the nuclear power industry. A 5 × 5 rod bundle having mixing spacer grids was adopted as a test rig, and was contained in a square flow housing with a 170 mm side length and 4670 mm length. The 25 rods in a bundle have dimensions of 25.4 mm in outer diameter and a 3863 mm length. The benchmark experiments have been performed at the MATiS-H water loop facility in KAERI. The axial bulk velocity in a rod bundle was maintained at about 1.50 m/s (equivalent to Re ∼50,000) with loop conditions of 35 °C and 1.57 bar measured upstream of the spacer during the experiments. Detailed measurements of the turbulent flow in the subchannels were accomplished using 2-D LDA at four different distances (0.5, 1, 4 and 10 DH) from the downstream of the mixing spacer grid. The upstream flow profiles also have been measured at the inlet of the mixing spacer grid for the inlet boundary condition. Precise measurements of the lateral and axial velocities in the subchannels are presented at four downstream distances, as well as the inlet from the mixing spacer grid of two types. Turbulence intensities and vorticities in the subchannels are also evaluated from the velocity measurements

  2. Characteristics of HEPA filter waste compactor developed by KAERI and comparison with Japan's and U. S. A's compactors

    Lee, G. M.; Ann, S. J.; Bae, S. M.; Son, J. S.; Hong, K. P. [KAERI, Taejon (Korea, Republic of); Kim, H. T. [KINS, Taejon (Korea, Republic of)

    2003-07-01

    The HEPA filter waste (Hexahedron, 610 x 610 x 292 mm, 20 Kg), which is used for the ventilation system in the nuclear industries, has relatively large volume to compare with it's weight. Due to the large volume of HEPA filter waste, it needs the large storage and/or disposal space and managing cost during the long period of storage and management. Many countries use the compactor to reduce the volume of HEPA filter waste. On this paper we introduce the new type compactor developed by KAERI with the characteristic comparison to the Japan's and U. S. A's compactors.

  3. On the generation of resonance cross sections in the resonance region of the neutron energy spectrum for the Ghana Research Reactor-1 fuel lattice unit cell using the WIMS lattice code

    The different physical and mathematical models and methods for calculating the resonance group constants or cross sections from resonance integrals in the resonance region of the neutron energy spectrum has been reviewed. The methods as outlined in the WIMS lattice program were used to calculate effective resonance cross sections for unit cell calculations of the five-region fuel lattice of the Ghana Research Reactor-1, using WIMSPC, the PC version of the versatile WIMSD/4 lattice code. The Ludwig Boltzmann multigroup neutron transport equation was solved for this exercise using the discrete ordinate spatial model (DSN) which provides solution to the differential form of the transport equation by the Carlson-Sn approach for the 13 different resonance energy groups defined in the WIMS code. The resonance escape probability, flux depression factors corrected for resonance absorption and evaluated correction factors to the resonance cross sections due to the depression of the neutron flux for the 13 resonance energy groups were also calculated

  4. L’intertexte entre le réel et l’imaginaire dans le cinéma de Wim Wenders

    Gustavo Coura Guimarães Castro

    2013-04-01

    Full Text Available Cet article interroge la façon dont le cinéma représente à l’écran la frontière qui existe entre le monde imaginaire des personnages et celui qu’on appelle «monde réel». Afin de vérifier la figuration à l’écran de cette relation parfois contradictoire, voir complémentaire, on fera une brève étude du film l’Etat des Choses (1982, de Wim Wenders. Le cinéaste est né en Allemagne, en 1945. Initialement, Wenders a étudié médicine et philosophie pendant deux ans, mais, en 1967, il a changé d’avis afin de poursuivre ses études dans l’audiovisuel. A partir de là, le réalisateur s’est fait connaître comme un des représentants les plus célèbres du «nouveau cinéma allemand». En 1978, il a été invité par Francis Ford Coppola pour participer au tournage du film Hammett. Ainsi, son cinéma a été toujours marqué par cette intersection culturelle entre l’Amérique et l’Europe. Notre objectif, dans cet article, est celui de mette en évidence les croisements du cinéma avec l’art, la relation des sujets avec l’espace, ainsi que les influences d’un sur l’autre dans les représentations cinématographiques. Notre référentiel théorique sera fondé, surtout, sur les notions de «fiction» et de «documentaire» chez François Niney. Nous espérons, à partir de cette réflexion, pouvoir comprendre plus en profondeur la façon dont Wenders travaille dans son film la métaphorisation de la lumière dans les mises en scènes, en proposant, en quelque sorte, un dialogue entre les concepts de «clarté» et d’«ombre», en ayant comme arrière-plan l’idée de fiction.This article shows the way that the cinema represents on the screen the boundaries between the imaginary world of the characters and the one that we used to call “real world”. In order to verify the representation on the screen of this contradictory relationship, sometimes complementary, we will do a brief study of the movie The State

  5. Conceptual Thermal Hydraulic Design of a 20MW Multipurpose Research Reactor (KAERI/VAEC joint study on a new research reactor for Vietnam)

    Chae, Hee Taek; Seo, Chul Gyo; Park, Jong Hark; Park, Cheol [Kaeri, Daejeon (Korea, Republic of); Vinh, Le Vinh; Nghiem, Huynh Ton; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    The conceptual thermal hydraulics design analyses for the 20 MW reference AHR core have been jointly performed by the KAERI and DNRI(VAEC). The preliminary core thermal hydraulic characteristics and safety margins for the AHR core were studied for various core flow rates, fuel assembly powers and core inlet temperatures. Statistical method was applied to the thermal hydraulic design of the reactor core. The MATRA{sub h} subchannel code has been applied to evaluate the thermal hydraulic performances of the AHR and the resulting thermal margins of the core under the forced convection cooling mode during a nominal power operation and the natural circulation mode during a reactor shutdown condition. In addition, typical accident analyses were carried out for a loss of flow accident by a primary pump seizure and a reactivity induced accident by a CAR rod withdrawal during a normal full power operation. The normal full power operation of the AHR was ensured with a sufficient safety margin for the onset of nucleate boiling phenomena. The AHR also had a sufficient natural circulation cooling capability to cool the core without the onset of nucleate boiling in the channel after a normal reactor shutdown and the anticipated transients. It was confirmed by the typical accident analyses that the AHR core was sufficiently protected from the loss of flow by the primary cooling pump seizure and the overpower transients by the CAR withdrawal from the MCHFR and fuel temperature points of view.

  6. Joint research project to develop a training course or nuclear policy decision makers and planners in developing countries between KAERI and IAEA

    Lee, E. J.; Suh, I. S.; Lee, H. Y. and others

    2000-12-01

    KAERI developed training course curricula on nuclear power policy and planning for decision makers and planners in developing countries under the assistance of the IAEA. It was utilized two IAEA staff members and a Korean consultation group were utilized for the development of curricula. Curriculum consists of training objectives, training contents in modular basis, detailed contents of each training module, training setting, training duration, session hours, and entry requirements of audience. One is workshop on nuclear energy policy for high-level decision makers in developing countries. The other is training course on nuclear power planning and project management for middle level managers in developing countries. The textbook in English will be printed by the end of February in 2001. Developed curricula will be implemented for Vietnam high level nuclear decision makers, middle level managers in developing countries and north Korea nuclear high level decision makers in 2001. These training courses' curricula and textbook will be utilized as basic technical documents to promote the national nuclear bilateral technical cooperation programs with Morocco, Egypt, Bangladesh, Indonesia, Ukraine, etc.

  7. Modeling the hydraulic characteristics of a fractured rock mass with correlated fracture length and aperture: Application in the underground research tunnel at KAERI

    A three-dimensional discrete fracture network model was developed in order to simulate the hydraulic characteristics of a granitic rock mass at Korea Atomic Energy Research Institute (KAERI) Underground Research Tunnel (KURT). The model N used a three-dimensional discrete fracture network (DFN), assuming a correlation between the length and aperture of the fractures, and a trapezoid flow path in the fractures. These assumptions that previous studies have not considered could make the developed model more practical and reasonable. The geologic and hydraulic data of the fractures were obtained in the rock mass at the KURT. Then, these data were applied to the developed fracture discrete network model. The model was applied in estimating the representative elementary volume (REV), the equivalent hydraulic conductivity tensors, and the amount of groundwater inflow into the tunnel. The developed discrete fracture network model can determine the REV size for the rock mass with respect to the hydraulic behavior and estimate the groundwater flow into the tunnel at the KURT. Therefore, the assumptions that the fracture length is correlated to the fracture aperture and the flow in a fracture occurs in a trapezoid shape appear to be effective in the DFN analysis used to estimate the hydraulic behavior of the fractured rock mass.

  8. Joint research project to develop a training course or nuclear policy decision makers and planners in developing countries between KAERI and IAEA

    KAERI developed training course curricula on nuclear power policy and planning for decision makers and planners in developing countries under the assistance of the IAEA. It was utilized two IAEA staff members and a Korean consultation group were utilized for the development of curricula. Curriculum consists of training objectives, training contents in modular basis, detailed contents of each training module, training setting, training duration, session hours, and entry requirements of audience. One is workshop on nuclear energy policy for high-level decision makers in developing countries. The other is training course on nuclear power planning and project management for middle level managers in developing countries. The textbook in English will be printed by the end of February in 2001. Developed curricula will be implemented for Vietnam high level nuclear decision makers, middle level managers in developing countries and north Korea nuclear high level decision makers in 2001. These training courses' curricula and textbook will be utilized as basic technical documents to promote the national nuclear bilateral technical cooperation programs with Morocco, Egypt, Bangladesh, Indonesia, Ukraine, etc

  9. Advanced Electrorefining Process at KAERI

    In order to enhance the throughput of a pyro-processing in which electrochemical processes are mostly engaged, the design of a continuous concept is required. The graphite cathode in the electro-refiner enables the uranium deposit on the cathodes to be stripped off spontaneously, resulting in a continuous reaction. The collected uranium deposits at the bottom of the inner cone of the reactor are transferred by a conveyor. The residuals in the anode basket after the uranium is depleted are noble metals. These are also collected at the bottom of the outer shell of the reactor, and conveyed from the reactor for a further treatment. This work addresses the design of the electro-refiner for a continuous operation. The behavior of particles such as uranium dendrites or noble metals was analyzed to achieve the proper operating conditions. The operating conditions for the cathode processor in which molten salt is distilled were also investigated. (authors)

  10. 基于WIM数据的上海长江大桥钢箱梁的寿命评估%Life Evaluation of Box Girder of Shanghai Yangtze River Bridge Based on WIM Data

    廖霞霞; 胡明敏; 徐昊

    2014-01-01

    Evaluation of the box girder details of Shanghai Yangtze river can provide the basis for the bridge ’ s maintenance and repairment .The research result of “Fatigue stress spectrum study of the steel box girder of Shanghai Yangtze River Bridge ,based on the analysis of WIM data” was used to get all the details of the stress spectrum of steel box girder .Then according to BS 5400 the tenth part ,confirming S-N curve which can reflect the details quality of fatigue .Finally ,estimating the steel box girder’s life with the combination of the Miner fatigue cumulative damage rule .According to the result ,we can know that during the current oper-ating process ,the box girder details have small stress amplitude ,high cycles and small damage magnitude . These details will not bring about fatigue damage .During the current operation ,the structure will keep high safety performance and reliability .Determine the damage point and the size of the damage value is unique in this article .%通过对上海长江大桥细节寿命的估算为大桥的维护和修复提供依据。基于论文“基于WIM数据的上海长江大桥钢箱梁应力谱研究”的研究结果,得到钢箱梁各细节的应力谱,同时参考英国桥梁规范BS 5400第十部分,确定了能够体现细节疲劳性能的 S-N曲线,并结合M iner疲劳累积损伤准则,对大桥钢箱梁细节寿命进行评估。研究结果表明,在现行日常运营过程中,钢箱梁各细节应力幅值较小,循环次数多,损伤量级很小,故细节不会发生疲劳破坏,即结构在现行运营期间保持了很高的安全性和可靠度。

  11. The '1986' WIMS nuclear data library

    An updated version, known as the 1986 Library has been developed. It contains data in 69 energy groups for 129 nuclides including fuel, moderator and structural materials, fission products and miscellaneous detectors. This document gives a brief summary of the data, concentrating on 2200 m.sec-1 cross sections, Maxwellian averages and resonance integrals, but also including other important data and references where possible. (U.K.)

  12. Performance of a train WIM system

    DOLCEMASCOLO,V; DE GRAAF, HJ

    2006-01-01

    Dans le cadre du prorjet Eureka Footprint dont l'objectif est de développer des méthodes de mesure de l'interaction véhicule/Infrastructure en service, les sociétés bass R&D and NedTrain Consulting ont mis au point un capteur de pesage en marche basé sur la technologie des fibres optiques et une station de pesage capable d'estimer les poids statiques totaux, des groupes d'essieux (boggies) et des essieux de groupe. Le système est aussi capable de détecter les défauts de roue ou de rail impliq...

  13. The CACTUS transport method in WIMS

    CACTUS solves the differential transport equation in two dimensions by the so-called Characteristics Method. By means of a combination of carefully selected numerical 'tracks' and a general, if somewhat tedious method of geometry description, the method has been used to solve a wide range of very complicated geometries from simple pincells to supercells (colorsets) of Advanced Gas-cooled Reactor (AGR) channels. In all of these cases the 2D geometry is solved exactly without the need to smear fuel pins. This paper describes the rather simple equations solved by CACTUS and explains the tracking methods that automatically impose the required boundary conditions. The accuracy of the solution is dependent on the number of angles at which tracks are evaluated, and also on the spacing between them. In this respect, the method has similarities with numerical methods of integrating collision probabilities; it is significantly different in that the transport equation is solved along each track segment in turn. One disadvantage of the method is that both azimuthal and polar angles must be integrated numerically. The methods have been only worked out in detail for rectangular outer boundaries, i.e. for square or rectangular pitch lattices, but could be applied equally well to hexagonal systems. 5 refs, 6 figs

  14. Advanced SFR concept design studies at KAERI

    Full text: Advanced SFR design concepts have been proposed and evaluated against the design requirements to satisfy the Gen IV technology goals. Two types of conceptual core designs, Breakeven and TRU burner cores were developed. Breakeven core is 1,200 MWe and does not have blankets to enhance the proliferation resistance. According to the current study, TRU burning rate increases linearly with the rated core powers from 600 MWe to 1,200 MWe. Considering 1) the realistic size of an SFR demonstration reactor for the long-term R and D plan with the goal of a demonstration SFR construction by 2028, and 2) the availability of a KALIMER-600 reactor system design that was developed in the last R and D phase, a TRU burner of 600 MWe was selected. The heat transport system of Advanced SFR was designed to be a pool type to enhance system safety through slow system transients, where primary sodium is contained in a reactor vessel. The heat transport system is composed of Primary Heat Transport System (PHTS), Intermediate Heat Transport System (IHTS), Steam Generating System (SGS) and Residual Heat Removal System (RHRS). The heat transport system was established through trade studies in order to enhance the safety and to improve the economics and performance of the KALIMER-600 design. Trade studies were performed for the number of IHTS loops, the number of PHTS pumps, Steam Generator (SG) design concepts, energy conversion system concepts, cover gas operation methods, and an improved concept of safety-graded passive decay heat removal system. From the study, the heat transport system of Advanced SFR has design features such as two IHTS loops, a Rankine cycle energy conversion system, two double-wall straight tube type SGs, and a passive decay heat removal system. In order to secure the economic competitiveness of an SFR, several concepts were implemented in the mechanical structural design without losing the reactor safety level. The material of reactor vessel and internal structure is a Type 316 stainless steel. The outer diameter of the reactor vessel is 14.5m, which is a very compact size compared to the other designs. Various R and D activities have been performed in order to support the development of Advanced SFR design concepts and to update computational tools. These activities include validating neutronics analysis codes, PDRC experiment, the conceptual design of supercritical carbon dioxide (S-CO2) Brayton cycle system, Na-CO2 interaction test, under-sodium viewing technique, computerization of structural integrity evaluation, sodium technologies, and metal fuel technology. (author)

  15. Repair-welding technology of irradiated materials - WIM project

    A new project on the development of repair-welding technology for core internals and reactor (pressure) vessel, consigned by the Ministry of International Trade and Industry (MITI), has been started from October 1997. The objective of the project is classified into three points as follows: (1) to develop repair-welding techniques for neutron irradiated materials, (2) to prove the availability of the techniques for core internals and reactor (pressure) vessel, and (3) to recommend the updated repair-welding for the Technical Rules and Standards. Total planning, neutron irradiation, preparation of welding equipment are now in progress. The materials are austenitic stainless steels and a low alloy steel. Neutron irradiation is performed using test reactors. In order to suppress the helium aggregation along grain boundaries, low heat input welding techniques, such as laser, low heat input TIG and friction weldings, will be applied. (author)

  16. Benchmarking of the ZR-6 critical assemblies using WIMS

    During the 1970 and early 1980 a wide ranging series of experiments was performed in the ZR-6 facility in Budapest. The cores consisted of arrays of UO2 fuel rods on a hexagonal pitch with light water moderator. Criticality was achieved by varying the moderator height.(Authors)

  17. Neutron transport in WIMS by the characteristics method

    This report is the text of a Paper presented by the author at the American Nuclear Society meeting in San Diego, California in June 1993. It summarises the characteristics method known as CACTUS for solving the neutron transport equation, and describes its application to a benchmark problem with adjacent gadolinium pins. The new CACTUS options (a) to subdivide regions into computational meshes, and (b) to extend the method to allow for the spatial variation of source distributions are highlighted. (Author)

  18. Status of the atomized uranium silicide fuel development at KAERI

    Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-08-01

    While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.

  19. Decontamination of radioactive corrosion products by KAERI decontamination process

    Jung, Chong-Hun; Park, Sang-Yoon; Ahn, Byung-Gil; Lee, Byung-Jik; Oh, Won-Zin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-31

    A study was performed to develop the chemical decontamination process, which is effective in removing the radioactive corrosion products with large amounts of Ni and Cr. The dissolution characteristics of decontamination agents and the material integrity of disk arm holder with Type 304 stainless steel were examined in high temperature conditions and the results have been compared with low temperature decontamination process. Dissolution tests revealed that oxides on disk arm holder had a spinel-type structure in the form of Fe{sub 1.7}Ni{sub 0.5}Cr{sub 0.8}O{sub 4}. In the dissolution steps, component metals were dissolved fast from the oxide in the early stage, while were dissolved very slowly in the later stage. This might be caused by reduction in metal concentration in the near surface of the oxide and by precipitation of reaction by product, MnO{sub 2}, which prevents reactants in solution from diffusion to the oxide surface. The average DF(Decontamination Factor) after a chemical decontamination, consisting of 3 oxidation-reduction steps, was 75 and an improved DF, of 150, was observed when a ultrasonic treatment was applied after a chemical decontamination, since the corrosion oxide become soft by the dissolution of grain boundary and crack of the oxide during chemical decontamination process. High temperature decontamination process showed remarkable improvement in decontamination effectiveness compared with traditional low temperature process. An examination of corrosion rates monitored during the decontamination, using corrosion coupons, showed that all process reagents caused minimal corrosion(Type 304 stainless steel: 1.7 x 10{sup -3} mil, Inconel 600: 6.6 x 10{sup -3} mil, Stellite-6: 1.2 x 10{sup -2} mil). (author). 19 refs., 4 tabs., 9 figs.

  20. Borehole heater test at KAERI Underground Research Tunnel

    At HLW repository, the temperature change due to the decay heat in near field can affect the hydraulic, mechanical, and chemical behaviors and influence on the repository safety. Therefore, the understanding of the thermal behavior in near field is essential for the site selection, design, as well as operation of the repository. In this study, various studies for the in situ heater test, which is for the investigation of the thermo-mechanical behavior in rock mass, were carried out. At first, similar in situ tests at foreign URLs were reviewed and summarized the major conclusions from the tests. After then an adequate design of heater, observation sensors, and data logging system were developed and installed with a consideration of the site condition and test purposes. In order to minimize the effect of hydraulic phenomenon, a relatively day zone was chosen for the in situ test. Joint distribution and characteristics in the zone were surveyed and the rock mass properties were determined with various laboratory tests. In this study, an adequate location for an in situ borehole heater test was chosen. Also a heater for the test was designed and manufactured and the sensors for measuring the rock behavior were installed. It was possible to observe that stiff joints are developed overwhelmingly in the test area from the joint survey at the tunnel wall. The major rock and rock mass properties at the test site could be determined from the thermo-mechanical laboratory tests using the rock cores retrieved from the site. The measured data were implemented in the three-dimensional computer simulation. From the modeling using FLAC3D code, it was possible to find that the heat convection through the tunnel wall can influence on temperature distribution in rock. Because of that it was necessary to installed a blocking wall to minimize the effect of ventilation system on the heater test, which is carrying out nearby the tunnel wall. The in situ borehole heater test is the first thermo-mechanical test in Korea. In the future, the results from the test will be utilized for different projects such as spent fuel storage, geothermal energy, sequestration of carbon-dioxide, and underground petroleum storage, which require the clear understanding on the thermo-mechanical behavior of rock mass

  1. AMO research activities and data centre in KAERI

    Dr. Rhee presented data center activities, which support several experimental and theoretical atomic, molecular and optical (AMO) physical programmes. The activities are mainly for fusion science and high precision trace analysis for nuclear safety. There were also some improvements in the ALADDIN database interface at AMODS, which employ the original FORTRAN ALADDIN codes

  2. Melting decontamination technology development for metallic waste recycling at KAERI

    The radioactively contaminated metallic wastes have been produced from the decommissioning of both two research reactors in Seoul and an uranium conversion plant in Daejeon. The melting decontamination technology was studied to reduce the radioactive waste volume by self disposal as non radioactive waste. The partitioning of radioisotope cobalt 60 and cesium 137 among a molten ingot, slag and dust phases have been investigated in a plasma are melter. A direct current plasma arc furnace was used to melt contaminated stainless steel, mild steel and aluminum wastes with a acid, neutral and basic slag (SiO2. CaO, Al2O3, Fe2O3, MgO) containing radioactive 60Co and 137Cs, to measure the partitioning phenomena. Calcium oxide and ferric oxide were added to provide an increase in the slag fluidity and oxidative potential respectively. Most of the 60Co remained in the ingot phase and it was barely present in the slag during the steel melting. 60Co decontamination factor was not highly dependent on the slag composition. However, it was found that a highly fluid basic slag is a little effective. The distribution ratio of 60Co into the ingot and the slag phase showed that about 90% to 95% was recovered in the ingots. But in the melting of aluminum wastes, it was contaminated by up to 70% in the slag phase. 137Cs was completely eliminated from the melt of the stainless steel as well as the carbon steel and distributed to the dust phase. The partition remaining in the slag depended on whether the slag was basic or acidic and had a high oxidative flux (Fe2O3). A maximum of 65% of the 137Cs remained in the slag phase with a high slag concentration and basicity. Generally, the137Cs distribution in the slag was between 10% and 25% during a lab scale arc furnace test

  3. Overview of the neutronics calculation system for the HANARO

    KAERI established the HANAFMS (HANARO Nuclear Analyses and Fuel Management System) for the in-core fuel management. The major components of the HANAFMS are the WIMS-KAERI and VENTURE. And several auxiliary codes such as REGAV-K, WIMPAK, MAPHEX, HEXSHUF, are supporting the system. The HANARO have carried out various kinds of reactor physics tests and experiments for 11 years. To support those experimental activities and to ensure the safe operation of the HANARO, the core follow up calculation is always performed along with the cycle operation using HANAFMS. (author)

  4. On the structure of Lattice code WIMSD-5B

    The WIMS-D code is a freely available thermal reactor physics lattice code used widely for thermal research and power reactor calculation. Now the code WIMS-AECL, developed on the basis of WIMS-D, has been used as one of lattice codes for the cell calculation in Canada and also, in 1998, the latest version WIMSD-5B is released for OECD/NEA Data Bank. While WIMS-KAERI was developed and has been used, originated from WIMS-D, in Korea, it was adjusted for the cell calculation of research reactor HANARO and so it has no confirmaty to CANDU reactor. Therefore, the code development applicable to cell calculation of CANDU reactor is necessary not only for technological independence and but also for the establishment of CANDU safety analysis system. A lattice code WIMSD-5B was analyzed in order to set the system of reactor physics computer codes, to be used in the assessment of void reactivity effect. In order to improve and validate WIMSD-5B code, the analysis of the structure of WIMSD-5B lattice code was made and so its structure, algorithm and the subroutines of WIMSD-5B were presented for the cluster type and the pij method modelling the CANDU-6 fuel

  5. Nuclear Data Processing for Reactor Physics Calculation

    Nuclear data processing for reactor physics calculation has been done. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in neutronic calculation. The processing code system such as NJOY-PC code has been used from linearization of nuclear cross-sections data and background contribution of resonance parameter (MF2) using RECONR module (0K) with energy range from 10-5 to 107 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (i.e. 293K) by using BROADR module. The Grouper and Therma modules will be applied for multi-groups calculation which suitable for WIMS/D4 (69 groups) and thermalization of nuclear constants. The final stage of processing nuclear cross-sections is updating WIMS/D4 library. The WIMSR module in NJOY-PC and WILLIE code will be applied in this stage. The evaluated nuclear data file, especially for 1H1 isotope, was taken from JENDL-3.2 and ENDF/B-VI for preliminary study. The results of nuclear data processing 1H1 shows that the old-WIMS (WIMS-lama) library have much discrepancies comparing with JENDL-3.2 or ENDF/B-VI files, especially in energy around 5 keV

  6. Dominique Deneffe, Famke Peters and Wim Fremout, Pre-Eyckian Panel Painting in the Low Countries

    2009-04-01

    Full Text Available Ce très bel ouvrage en deux volumes, publié en anglais, présente le résultat de recherches menées à l’instigation de l’IRPA et du Centre pour l’étude de la peinture du XVe siècle dans les Pays-Bas du Sud et la principauté de Liège sur un corpus d’œuvres particulières. Les panneaux peints pré-Eyckiens datant des environs de 1400 sont relativement peu nombreux : on en recense aujourd’hui une trentaine au monde. Dix d’entre eux, constituant la majeure partie des panneaux conservés en Belgique, s...

  7. Dominique Deneffe, Famke Peters and Wim Fremout, Pre-Eyckian Panel Painting in the Low Countries

    Sophie Moreaux

    2009-10-01

    Full Text Available L’Institut Royal du Patrimoine Artistique et le Centre pour l’étude de la peinture du XVe siècle dans les Pays-Bas du Sud et la Principauté de Liège présentent les résultats des recherches menées sur dix oeuvres pré-eyckiennes, dans un très beau livre en deux volumes, publié en anglais. L’histoire de la peinture d’avant 1420 dans les Pays-Bas méridionaux a longtemps été teintée de légendes et de mythes persistants. Lorne Campbell, dans la préface, nous rappelle les propos de Gustave Friedrich...

  8. An update of the Wims-E data processing routines (WEDRO-1.1)

    The WEDRO-1.1 report is an addendum to the original WEDRO-1.0 technical report, RTE01-2/2-035 (March 1988), and describes some recent additions and changes incorporated into the code, which in now known as WEDRO-1.1. Attention is also drawn to certain errors present in the coding of WEDRO-1.0 which could have resulted in erroneous output cross-sections. A list of predefined material ID's used in WEDRO-1.1, as well as the keyword description and the format of the output WEDRO interface file, is given in this report. 1 tab., 9 refs

  9. A user's guide to the WIMS-E module W-CACTUS

    CACTUS is a code that solves the multigroup neutron transport equations using a characteristics method. This is a numerical approach in which the differential form of the Boltzmann equation is integrated along explicit tracks through the geometry and the neutron flux is obtained by a summation of the contributions made by a selected sample of such tracks. This report describes the theory very briefly, and concentrates on providing the information required to use the program. (author)

  10. Is Wim slim?: Samen een biogas netwerk gebruiken? : 8 december 2014

    Hengeveld, E.J.; Bekkering, J.; Gemert, W.J.T. van; Broekhuis, A.A.

    2014-01-01

    berekening van kosten voor biogas transport in een biogas verzamel netwerk; twee layouts worden vergeleken: de ster-layout en de visgraat-layout. Investeringen en operationele kosten, inclusief compressiekosten worden in een contante waarde berekening meegenomen. Flexigas symposium 8 december 2014

  11. Scientific Design of Large Scale Sodium Thermal-Hydraulic Test Facility in KAERI

    A full passive decay heat removal system is implemented as an advanced design feature for the SFR which is currently being developed in Korea. Its operation depends purely on the natural circulation in a primary heat transport system and a passive decay heat removal system, and no active component or operator action is required. For the demonstration of the design concept, a large scale sodium thermal-hydraulic test facility is being designed with the plan of installation in 2013. In the experiments, the cooling capability during the long- and short-term periods after reactor shutdown will be demonstrated and also the produced experimental data will be utilized for the assessment and verification of the safety and performance analysis codes. In this paper, the preliminary design features of the test facility are presented along with the design requirements and methodology. (author)

  12. In situ borehole heater test at the KAERI Underground Research Tunnel in granite

    Highlights: • An in situ heater test was carried for investigating TM behavior in granite. • When the heater temperature was 118 °C, the rock temperature at 0.3 m was 50 °C. • The heater was installed at disturbed rock zone, which is 0.5∼1.5m from the wall. • The influences of seasonal temperature variation and heat convection were observed. • The thermal stress increased almost linearly up to 5 MPa. - Abstract: An in situ borehole heater test was carried out in an underground research tunnel at a shallow depth in granite. During the test, the heater temperature was increased to 90 °C to simulate the thermo-mechanical behavior of crystalline rock under normal underground high-level radioactive waste repository conditions. The air, wall and rock temperatures were measured over a period of about four years. At the end of the test, the heater temperature was increased to 118 °C to simulate abnormal overheating conditions. The peak temperatures at the observation holes located at 0.3 m and 0.6 m from the heater hole were approximately 50 °C and 37 °C, respectively. The temperature measurements allowed observations of the effects of rock joints and heat convection through the tunnel wall on the rock temperature distribution. When the power was shut down, the rock temperatures and stress returned rapidly to the original rock temperature

  13. Research activities on radioecology for the past ten years: Experiments and modeling at KAERI

    Experiments in a greenhouse have been conducted to evaluate the effects of radionuclides on various plants. Transfer factors, translocation factors, and other parameters have been measured particularly for major foodstuffs, such as rice and vegetables. A computer code was established to assess the environment in case of acute radionuclide release by accident. Verification and sensitivity analysis have been carried out for the integrity of this code. (author)

  14. Present Status and Results from the KAERI Compact THz FEL Facility

    Jeong, Y U; Lee, B C; Park, S H

    2005-01-01

    We have developed a laboratory-scale users facility with a compact terahertz (THz) free electron laser (FEL). The FEL operates in the wavelength range of 100-1200 μm, which corresponds to 0.3-3 THz. The peak power of the FEL micropulse having 30 ps pulse duration is 1 kW and the pulse energy of the 3-μs-FEL-macropulse is approximately 0.3 mJ. The main application of the FEL is THz imaging for bio-medical researches. Transmitted THz imaging of various samples including bugs have been measured. The samples were scanned by a 2-dimensional stage at the focal point of the THz beam. The bugs were not dry because they were killed just before experiments. We could get the transmitted THz imaging of the bugs at 3 THz with the high power THz FEL. THz spectral characteristics of several materials have been studied by the FEL and a THz FTIR spectrometer. We will introduce recent results on the imaging and spectroscopy by the THz FEL.

  15. Temperature Measurement and Water Flow Calorimetry for the Neutral Beam Test Stand Operation at KAERI

    Temperature measurements during the beam line operation of the neutral beam test stand(NB-TS) is very important for the estimation of the absorbed energy by the beam line components such calorimeter and also for the temperature monitoring of the various components, and have been accomplished by the utilization of many of the thermocouples(TCs) installed onto the NB-TS and the data acquisition system(DAQ) based on the National Instruments' (NI) SCXI system. Preliminary estimations of the absorbed energy by the calorimeter during the beam extraction have been made. Greater efforts for the noise reduction in the TC signal acquisition has been made with partial success. We present the status of the temperature measurement and water flow calorimetry(WFC) related to the NB-TS operations

  16. A study on the management of intellectual property for the pending projects in KAERI

    This study targeted researching a main character of intellectual property and response strategy regarding a nuclear research reactor project in the ANSI region. The study shows that each member country of the ANSI has its own registering system of patent and other intellectual property. Moreover, we confirmed that there was no previously registered patent in Malaysia, Singapore, Thailand, Vietnam, and Indonesia that have an intent to import research reactor. As a result of this study we suggest that registering patent relating a nuclear research reactor not only in potential importing countries but also in major nuclear countries are preferable because this approach is a more basic strategy for technology and market protection. Although major nuclear country or company has own essential or unique patent regarding nuclear side, our registering that type of patent to potential importing countries is also valid for banning rival company's intrusion to the market and get a better position for negotiation with importing country as first register of intellectual property keeps a priority in the country

  17. Ultrasonic measurement of water layer thickness by horizontal flow pattern profile in a KAERI HAWL

    An ultrasonic measurement technique for determining water layer thickness is presented. The technique can obtain information of the water layer thickness in a tube in the form of a horizontal flow pattern profile through the used of a correct quantitative method. The main objective of the present work is to measure the water layer thickness of the flow using an ultrasonic measurement system. Ultrasonic measurement techniques of water layer thickness are produced to measure the variations in water layer thickness in the horizontal stratified flow and vertical annular flow regimes. (author)

  18. Statistical properties of background fractures observed at the deep borehole in KAERI Underground Research Tunnel

    From the analyses of borehole logging and hydraulic test results, the statistical properties of background fractures were characterized, and HRDs were defined. According to the geological model of the KURT site, the hydrogeological units in the site were categorized to the hydraulic soil domains(HSDs), the hydraulic rock domains and the hydraulic conductor domains. In this study, we analyzed the properties of background fractures observed at Db-1, which is a deep borehole located in KURT with the depth of 600m, and characterized the HRDs in the site

  19. Recent Progress on Atomic Data for Fusion Plasma in KAERI Nuclear Data Center

    Kwon, Duckhee; Hwang, In Hyuk; Rhee, Yongjoo; Lee, Youngouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Atomic structure and collision cross sections are essential data for spectroscopic diagnostics of fusion plasma. We have carried out state-of-the-art calculations on cross sections for electron-impact ionization (EII) cross sections of various atoms and ions. Here we report our recent progress on those calculations and discuss future research plan focusing on the actual need for fusion plasma diagnostics. We have calculated EII cross sections of P-like ions including Fe{sup 11+}, and W ions based on a DWA. Present calculations agree with experiments better than previous other calculations. However, for lowly charged ions, our DWA calculations which uses approximated, non unitary scattering matrix have sizable discrepancies with experiments. Hence unitary corrections would be required to improve EII calculations for lowly charged ions. As well more sophisticated R matrix calculations would be required for EII of those ions in order to test DWA calculations mutually.

  20. Trial Burns of low-level radioactive wastes the demonstration-scale incineration plant at KAERI

    Behavior of radionuclides such as Co, Mn and Cs in the incineration process was studied by trial burns of simulated wastes with radio-isotope tracers. Behavior of nonvolatiles, Co and Mn, was similar to that of particulate matters in the process. Decontamination factors(DFs) for Co and Mn were 4.7 x 10 and 6.2 x 10, respectively. Behavior of semivolatile radio-isotope, Cs was temperature dependent. DFs for Cs at two different incineration temperature of 850 deg C and 700 deg C were 2.8*10 and 2.6*10, respectively. Trial bums of dry active waste(DAW)transported from nuclear power station(NPS) Kori 3, 4 were also performed. DF for gross radioactivity in DAW was 1.1x10. This was a little higher than the estimated value, which was calculated from the tracer test results and nuclides distribution in the DAW. Average emission concentration was 0.019 B1/Nm, which could meet the maximal permissible concentration(MPC) in stack emission. 5 figs., 3 tabs., 10 refs. (Author)

  1. A field focused university education in NRI : A case of UST-KAERI

    The University of Science and Technology (UST) was founded in Oct. 2003 through the approval of the former Ministry of Education and Human Resources Development to nurture R and D professionals in convergence technology, who will lead us into the 21st century, the era of information technology. In the era of 'global talent war', every country competes to secure young scientific/technological leaders who will cope with future global and national agenda. In accordance with this need, advanced countries have diversified their higher level education channels utilizing the representative national research institutes or laboratories in addition to the traditional graduate school. Recently, almost all the advanced countries operate a unique graduate school or university to nurture higher talents based on the national research institutes (NRIs) which lead national strategic R and D fields. They include International Max Planck research school(IMPRS) and International Helmholtz graduate school in Germany, Watson school of the Cold Spring Harbor Lab. and Kellogg school of the Scripps Research Institute in US, Feinberg graduate school of the Weizmann Institute in Israel, SOKENDAI in Japan, and UST in Korea. UST has enormous research facilities and special high tech equipment, and has faculty members who have outstanding research records, which is not common in general universities. With high tech equipment, the excellent faculty members are participating in useful field focused R and D education. Instead of having a rigid department, UST allows flexible opening of a major for new convergence technology. By doing this, UST is responding actively to fast changes in science and technology. UST manages 29 campuses granted as government funded research institutes in the area of science and technology with educational functions. Each campus member and faculty are joining a network related to educating each other and cooperating with different research activities, which is expanding to enhance collaboration with institutions related to diverse areas of research. In addition, to share and reflect the newest trend of research on education, collaborated lectures are operating, which have grafted the know how each of its campus professors. The profiles of these professors are provided to students in a masters, integrative and doctorate program, who apply to enter the university. The students can determine the particular research area with the help of their expected academic advisor in advance, which allows a customized research education for the students

  2. Development of quality assurance for HLW disposal R and D in KAERI

    To assure the credibility of R and D results and to systematically and effectively perform experiments and computations for the performance assessment of high-level radioactive disposal in Korea, the total quality assurance(QA) program is under development. To effectively manage the R and D's and perform decision makings so called WEB based AQ system is proposed based on the U.S.N.R.C. 10CFR50. The current proto-type QA system shall be extended to accommodate functionalities such as QA procedures, forms, and decision-making pathways. In parallel with the QA system, the technical data management (TDM) system is also applied to get probabilistic density functions (PDF's) required for probabilistic safety assessment (PSA). So-called SNL-NRC protocol was applied to construct the PDF for solubility limits of two nuclides

  3. KAERI Underground Research Facility (KURF) for the Demonstration of HLW Disposal Technology

    In order to dispose of high-level radioactive waste(HLW) safely in geological formations, it is necessary to assess the feasibility, safety, appropriateness, and stability of the disposal concept at an underground research site, which is constructed in the same geological formation as the host rock. In this paper, the current status of the conceptual design and the construction of a small scale URL, which is named as KURF, were described. To confirm the validity of the conceptual design of the underground facility, a geological survey including a seismic refraction survey, an electronic resistivity survey, a borehole drilling, and in situ and laboratory tests had been carried out. Based on the site characterization results, it was possible to effectively design the KURF. The construction of the KURF was started in May 2005 and the access tunnel was successfully completed in March 2006. Now the construction of the research modules is under way

  4. A study on current status of KAERI's international cooperation programs and strategies for effective implementation

    This is a report on the status and analysis of standard agreement for technical cooperation, expert mission service and technical staff attachment. This report comprises a total of six chapters. Chapter II discusses the status of the technical cooperation agreement which took effect late 1998, and various other model agreements for technical cooperation. Chapter III provides information on the status, regulations, procedures for the expert mission service, and Chapter IV details the current status of the technical staff attachment and the related procedures. Chapter V deals with the utilization status and analysis of the English Counselor, and Chapter VI is the Conclusion. This report has tow objectives. First, we have never published reference books related to standard agreement for technical cooperation, expert mission service and technical staff attachment necessary for international joint research until now. As a result, the research divisions have often asked many questions to the office of international cooperation. Therefore, we expect that many difficulties will be removed and procedures simplified if the research divisions use this report as a reference book. Second, we plan to use this report reference book for policy decisions after establishing the database. (author). 5 tabs., 9 figs

  5. Overall Analysis of Meteorological Information in the KAERI Site (2008 Annual Report)

    Inspection and repair of tower structure and lift, instrument calibration have been done in the scope of 'Environmental Radiation Monitoring Around the Nuclear Facilities' project. Wind direction, wind speed, temperature, humidity at 67m, 27m, and 10m height and temperature, humidity, atmospheric pressure, solar radiation, precipitation, and visibility at surface have been measured and analyzed with statistical methods. At the site, the prevailing wind directions were N at 67m, NNW at 27m, and NNE at 10m height, but SW were also dominant with N at all heights. The calm distributed 20.9% at 67m, 39.3% at 27m, 40.7% at 10m height. Wireless data transmission to MIPS(Meteorological Information Processing System) has been done after collection in the DAS where environmental assessment can be done by the developed simulation programs in both cases of normal operation and emergency

  6. Overall Analysis of Meteorological Information in the KAERI Site (2007 Annual Report)

    Inspection and repair of tower structure and lift, instrument calibration have been done in the scope of 'Environmental Radiation Monitoring Around the Nuclear Facilities' project. Wind direction, wind speed, temperature, humidity at 67m, 27m, and 10m height and temperature, humidity, atmospheric pressure, solar radiation, precipitation, and visibility at surface have been measured and analyzed with statistical methods. At the site, the prevailing wind directions were N in winter, W in 2nd, SW in 3rd quaters. The calm distributed 24.1% at 67m, 43.4% at 27m, 54.7% at 10m height. Wireless data transmission to MIPS(Meteorological Information Processing System) has been done after collection in the DAS where environmental assessment can be done by the developed simulation programs in both cases of normal operation and emergency

  7. Overall Analysis of Meteorological Information in the KAERI Site (2006 Annual Report)

    Inspection and repair of tower structure and lift, instrument calibration have been done in the scope of 'Environmental Radiation Monitoring Around the Nuclear Facilities' project. Wind direction, wind speed, temperature, humidity at 67m, 27m, and 10m height and temperature, humidity, atmospheric pressure, solar radiation, precipitation, and visibility at surface have been measured and analyzed with statistical methods. At the site, the prevailing wind directions were N in winter, SW in 2nd, E in 3rd quaters. The calm distributed 14.7% at 67m, 33.2% at 27m, 57.3% at 10m height. Wireless data transmission to MIPS(Meteorological Information Processing System) has been done after collection in the DAS where environmental assessment can be done by the developed simulation programs in both cases of normal operation and emergency

  8. Overall Analysis of Meteorological Information in the KAERI Site (2009 Annual Report)

    Inspection and repair of tower structure and lift, instrument calibration have been done in the scope of 'Environmental Radiation Monitoring Around the Nuclear Facilities' project. Wind direction, wind speed, temperature, humidity at 67m, 27m, and 10m height and temperature, humidity, atmospheric pressure, solar radiation, precipitation, and visibility at surface have been measured and analyzed with statistical methods. At the site, the prevailing wind directions were N at 67m, NNW at 27m, and N at 10m height, but SW were also dominant with N at all heights. The calm distributed 27.2% at 67m, 27.9% at 27m, 53.2% at 10m height. Wireless data transmission to MIPS(Meteorological Information Processing System) has been done after collection in the DAS where environmental assessment can be done by the developed simulation programs in both cases of normal operation and emergency

  9. A Study on the Management of Intellectual Property for the Pending Projects in KAERI

    This study is to analysis legal status of intellectual property of the Jordan Researching and Training Reactor(JRTR). To get the goals, researching internal and international laws related with intellectual properties and reviewing the JRTR project are performed. Not only technology itself but also human resources joined the project are considered to find best solution for management. This study will be a good base for the JRTR project itself and other similar projects

  10. Establishment of KAERI Strategy and Organization for Fusion Power Technology Research

    International and domestic status of development activities of nuclear fusion energy technologies are analyzed and summarized. From these results a verifiable R and D strategy is derived which allows purposeful and successful participation in the ITER project and thus enables a domestic technological basis of the commercialization of nuclear fusion energy. A 45-year, three-stage plan is proposed with a detailed plan for the 10-year, 1st stage where a conceptual design of a Korean demonstration fusion power plant (KDEMO) will be developed as well as its key component designs such as breeder blanket

  11. A Novel Approach to Find Optimized Neutron Energy Group Structure in MOX Thermal Lattices Using Swarm Intelligence

    Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that UO2-PuO2 (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the UO2 fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of H2O moderated UO2-PuO2 (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure

  12. Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors

    The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.

  13. Progress of the Hard-wired Instrumentation and Control Works for the Neutral Beam Test Stand at KAERI

    Jung, Ki Sok

    2005-12-15

    Progress of the hard-wired instrumentation and control works for the neutral beam test stand(NB-TS) has been existed for the past one year period. Details of the installed arc detector circuit are explained. LN{sub 2} level and temperature control during the cryosorption pumping operation are explained with an emphasis on its control circuit. With an expectation of more accurate and sensitive measurement of temperatures than the thermocouple utilization during the calorimeter operation, PT-100 resistance temperature detector(RTD) utilization is initiated and the results are described. During the ion beam experiment, physical measurements are made with some delayed time than the beam extraction, and thus a delayed trigger pulse generator was fabricated and installed to the system. Underlying principles of the electronic circuits for the interlock implementation and optical signal transmission are introduced. These are basically the application of operational amplifier circuits. A cautious aspect of the SMPS(switch mode power supply) utilization is also give.

  14. KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

    Lee, Jang Soo; Kim, Jang Yeol; Eum, Heung Seop

    1997-07-01

    Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs.

  15. Progress of the Hard-wired Instrumentation and Control Works for the Neutral Beam Test Stand at KAERI

    Progress of the hard-wired instrumentation and control works for the neutral beam test stand(NB-TS) has been existed for the past one year period. Details of the installed arc detector circuit are explained. LN2 level and temperature control during the cryosorption pumping operation are explained with an emphasis on its control circuit. With an expectation of more accurate and sensitive measurement of temperatures than the thermocouple utilization during the calorimeter operation, PT-100 resistance temperature detector(RTD) utilization is initiated and the results are described. During the ion beam experiment, physical measurements are made with some delayed time than the beam extraction, and thus a delayed trigger pulse generator was fabricated and installed to the system. Underlying principles of the electronic circuits for the interlock implementation and optical signal transmission are introduced. These are basically the application of operational amplifier circuits. A cautious aspect of the SMPS(switch mode power supply) utilization is also give

  16. Finite element analysis of laminated rubber bearings - verification with KAERI HDRB, ALMR HDRB and CRIEPI LRB data

    In this paper, the results of numerical predictions (Finite element method) on single isolator data of High Damping Rubber Bearing supplied by Korea and USA and Lead Rubber Bearing supplied by Japan are presented. Due to symmetry in geometry and loading conditions, a 180 deg sector model has been analysed using the general purpose finite element computer code ABAQUS. The results show that for the combined compression and shear loading case the numerical predictions match well with the experimental results. The comparison is particularly very good in the shear strain range 50% to 150% which is of actual interest to design. The variation in the vertical compressive load from 0% to 200% of the design vertical load does not influence the horizontal displacement response due to shearing loads. (author)

  17. KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

    Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs

  18. KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

    This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs

  19. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  20. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

    The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, keff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal-hydraulics study of the TRIGA core.

  1. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

    Uddin, M.N. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh); Sarker, M.M., E-mail: sarker_md@yahoo.co [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, GPO Box 3787, Dhaka-1000 (Bangladesh); Khan, M.J.H. [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, GPO Box 3787, Dhaka-1000 (Bangladesh); Islam, S.M.A. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh)

    2010-03-15

    The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k{sub eff} and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal-hydraulics study of the TRIGA core.

  2. Indian advanced heavy water reactor for thorium utilisation and nuclear data requirements and status

    BARC is embarking on thorium utilisation program in a concerted and consistent manner to achieve all round capabilities in the entire Thorium cycle under the Advanced Heavy Water Reactor (AHWR) development program. Upgrading our nuclear data capability for thorium cycle is one of the main tasks of this program. This paper gives a brief overview of the physics design features of the AHWR. The basic starting point of the analysis has been the lattice simulation of the fuel cluster employing the WIMS-D4 code package with 1986 version of 69 group library. For the analysis of thorium cycle, the present multi group version contains the three major isotopes viz., 232Th, 233U and 233Pa. To correctly evaluate the fuel cycle we require many more isotopes of the Th burnup chain. With the help of NDS, IAEA, many other isotopes of interest in AHWR, actinides in the thorium burnup chain, burnable absorbers, etc., were generated. Some of them were added to the WIMS-D4 library and the results are discussed. The WIMS-D4 library is also being updated as part of the IAEA coordinated research project on Final Stage of WLUP with international cooperation. India is also taking part in CRP. The evaluation of AHWR lattice with this new library is presented. Some comments regarding the fission product data being used in WIMS libraries are given, which are tuned to U-Pu cycles. The measurements for 233U are rather old. Measurements in high energies are also very sparse. More attention by nuclear data community is required in this regard as well. India has also begun a modest program to assess the ADS concepts, with the aim of employing thermal reactor systems, such as AHWR. A one way coupled booster reactor concept is being analysed with available code systems and nuclear data. A brief summary of this concept is also being discussed in this paper. A general survey on the quality of the evaluated nuclear data of the major and minor isotopes of thorium cycle is also given. A major

  3. Immigrants' attitudes towards welfare redistribution. An exploration of role of government preferences among immigrants and natives across 18 European welfare states / Tim Reeskens, Wim van Oorschot

    Reeskens, Tim

    2015-01-01

    Artiklis käsitletakse pagulaste arvamusi heaoluriigi kohta 18 Euroopa riigis (sh Eesti), täpsemalt, kuidas nad suhtuvad saadavatesse sotsiaaltoetustesse, võrreldakse ka pagulaste hoiakuid põlisrahvastiku omadega. Aluseks 2008. aasta Euroopa sotsiaaluuringu heaolu hoiakute moodul

  4. Studying the fuel burnup of MNSR reactor and estimating the concentrations of main fission products using the codes WIMS-D4 and CITATION

    The codes WIMSD-4 and BORGLES - part of the MTR-PC code package- have been applied to prepare the microscopic cross section library for the main elements of MNSR core for 6 neutron energy groups. The generated library was utilized from the 3D code CITATION to perform the calculation of fuel burn up and depletion including the identification of main fission products and its effects on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products. The burn up results indicated that the core life time of MNSR is being mainly estimated by Sm-149 following by Gd-157 and Cd-113. The accumulation of these actinides during 100 continuous operation days caused a reduction of ca. 2 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 1.8 mk which relates to the whole discontinuous operation period of the reactor since its start and up to now. The calculation procedure simulates the sporadic operation with an adequate continuous operation period. This approximation is valid for the long lived actinides that mainly dictate the core life time. However, it is an overestimation for the concentration of short lived radioactive products like Xe-135. In the framework of this analysis the possibility of replacement of current MNSR fuel through low enriched fuels has been explored for two the fuel types U02-Mg and U3Si-Al. The results indicate that the first type (UO2-Mg) realize the criticality conditions with low enrichment of 20%, whereas the second type (U3Si-Al) required increasing the uranium enrichment up to 33%. For both fuel types the contribution of plutonium isotopes on the criticality has been also evaluated. Additionally, the influence of mixing burnable absorbers (Gd-113, Cd- 113) with the fresh fuels was investigated to identify their long-term control effect on the excess reactivity. The calculation showed that the decreasing of excess reactivity in case of Cd-113 is more homogenous and over the time due to the lower burn up rate of Cd-113. Finally, the study was devoted also to calculate the contribution of adding a 3 mm Beryllium shim to the top reflector. The comparison of calculation results with the available experimental data indicates good agreement with a relative deviation of about 5%. (author)

  5. Income inequality and depression: the role of social comparisons and coping resources / Ioana van Deurzen, Erik van Ingen; Wim J. H. van Oorschot

    Deurzen, Ioana van

    2015-01-01

    Artiklis käsitletakse sissetulekute ebavõrdsuse ja depressiooni seoseid, pöörates tähelepanu riikidevahelistele erinevustele ja seda mõjutavatele faktoritele Euroopa riikides, sh Eesti. Aluseks on Euroopa sotsiaaluuringu (2006/2007) andmed

  6. Assessment of MARS for downcomer multi-dimensional thermal hydraulics during LBLOCA reflood using KAERI air-water direct vessel injection tests

    The MARS code has been assessed for the downcomer multi-dimensional thermal hydraulics during a large break loss-of-coolant accident (LBLOCA) reflood of Korean Next Generation Reactor (KNGR) that adopted an upper direct vessel injection (DVI) design. Direct DVI bypass and downcomer level sweep-out tests carried out at 1/50-scale air-water DVI test facility are simulated to examine the capability of MARS. Test conditions are selected such that they represent typical reflood conditions of KNGR, that is, DVI injection velocities of 1.0 ∼ 1.6 m/sec and air injection velocities of 18.0 ∼ 35.0 m/sec, for single and double DVI configurations. MARS calculation is first adjusted to the experimental DVI film distribution that largely affects air-water interaction in a scaled-down downcomer, then, the code is assessed for the selected test matrix. With some improvements of MARS thermal-hydraulic (T/H) models, it has been demonstrated that the MARS code is capable of simulating the direct DVI bypass and downcomer level sweep-out as well as the multi-dimensional thermal hydraulics in downcomer, where condensation effect is excluded. (authors)

  7. Development of operation scenarios with high bootstrap, negative shear configuration for large-aspect-ratio (LAR) bootstrap tokamak KT-2 at KAERI

    Through time dependent transport simulation, the authors have developed an operation scenario with high bootstrap, negative shear configuration for KT-2 tokamak, so that investigations of the so-called advanced tokamak operations are possible. In this study, they have concentrated on calculation of power requirements of heating and current drive system, MHD (Magnetohydrodynamic) stability and compatibility with PF (Poloidal Field) coil system. The study shows that high bootstrap (>70%), MHD stable (up to βN ≤ 3.9) operation scenario with negative shear configuration is possible within KT-2 tokamak design specification

  8. Establishment of Safety Analysis System and Technology for CANDU Reactors

    To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

  9. Analysis of LWHCR-PROTEUS Phase 2 experiments performed using the AARE system with JEF-1 based data libraries, and comparison with other codes

    In this report the capability of the AARE modular code system and JEF-1 based nuclear data libraries to analyse LWHCR lattices is investigated by calculating the wet and dry cells of the PROTEUS-LWHCR Phase 2 experiment. The results are compared to those obtained using several cell codes, including WIMS-D, BOXER, MICROX-2, KARBUS, GRUCAH, and SPEKTRA. In particular, the main features of AARE, such as the self-shielding of resonance cross sections in the whole energy range of importance for structural materials and actinides (including the low energy resonances of heavy actinides), the shielding of oxygen resonances in the MeV range, the generation of adequate fission source spectra, the accurate calculation of migration areas, and the efficiency of the removal correction are investigated. It is shown that AARE can predict the k∞ void coefficient well with a 1 % deviation from experiment, even if a coarse 70 netron group library is used. KARBUS and the related 69 group KEDAK-4 library give as well a reliable estimate, but lead to less accurate prediction of reaction rates. The other codes give larger deviations. The JEF-1 evaluation for 242Pu gives systematically about 25 % too high capture rates in the fast energy range (above 1 keV). (author) 39 refs., 24 figs., 13 tabs

  10. Development calculational procedures for the neutron physics design of advanced pressurized water reactors (APWR) with tight triangular lattices in hexagonal fuel assemblies

    The new procedures for the calculation of infinite reactor zones build a synthesis of wellknown fast breeder (FBR) and light water reactor (LWR) methods. The data libraries are based on the 69 energy group structure of the WIMS code for thermal reactors and use the flexible storage mode of the FBR libraries. For the calculation of effective cross sections in the energy of neutron resonances, being very important in the APWR with its strongly epithermal neutron spectrum, several options are available. In most applications improved selfshielding tabulation formalisms (f-factor concept) are used. For more accurate investigations the fine flux programs ULFISP (own development) or RESABK (IKE, Stuttgart) may be selected. All cross section calculations use a modified version of the FBR code GRUCAL. Particularly the calculation of voided lattices may be improved at 69 groups with the program REMOCO or with a new 334 group library. The new calculational procedures could be qualified for a large number of LWR, APWR and FBR applications. The fuel assembly and whole core calculations are performed with available FBR methods. A new reactor core simulation program ARCOSI has been developed for the investigation of an APWR equilibrium core. The required cross sections come from fast interpolations of fuel assembly data on code-own libraries. The whole core calculations are performed with the fast nodal code HEXNODK, adopted from KWU. All calculational procedures are part of the powerful FBR code system KAPROS. (orig.)

  11. The history for thirty years of Korea Atomic Energy Research Institute

    This book gives description of history for thirty years of KAERI. It contains five chapters, which reports the process and development of KAERI, embryonic stage with nuclear energy for peace, the process of establishment of KAERI and building of the KAERI in 1960s, period of growth with change of international situation and measurement of KAERI and launching for KAERI in 1970s, period of technical independence for safe regulation and establishment nuclear safe center in 1980s and prospect on technical development of nuclear energy research like basic R and D.

  12. Validation of Decommissioning Engineering System Application against KRR-2

    KAERI is the only expert group which has decommissioning experiences and KAERI is trying to develop computer code to converge all the data which has been accumulated during KRR (Korea Research Reactor)-1 and 2 and UCP (Uranium Conversion Plant) decommission. This paper contains validation results of the KAERI DES by using KRR-2 decommissioning data. As a responsible leading group of Korean decommissioning research field, KAERI has been developing DES application program. One of decommissioning experience data, KRR-2 was used for KAERI DES validation and it successfully is reflected in KAERI DES

  13. Validation of Decommissioning Engineering System Application against KRR-2

    Jin, Hyung Gon; Park, Seungkook; Park, Heeseong; Song, Chanho; Ha, Jaehyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    KAERI is the only expert group which has decommissioning experiences and KAERI is trying to develop computer code to converge all the data which has been accumulated during KRR (Korea Research Reactor)-1 and 2 and UCP (Uranium Conversion Plant) decommission. This paper contains validation results of the KAERI DES by using KRR-2 decommissioning data. As a responsible leading group of Korean decommissioning research field, KAERI has been developing DES application program. One of decommissioning experience data, KRR-2 was used for KAERI DES validation and it successfully is reflected in KAERI DES.

  14. Assessment of the Accuracy and Classification of Weigh-in-Motion Systems: Part 2 European Specification

    Jacob, Bernard; O'Brien, Eugene J.; Newton, William

    2000-01-01

    This is the second part of a two-part paper which addresses the issue of accuracy in weigh-in-motion (WIM) systems. The first part develops the statistical background necessary for any system of accuracy classification applied to a WIM system. This second part describes a draft European specification for the weigh-in-motion (WIM) of road vehicles, prepared by the COST 323 management committee. The philosophy behind the specification is outlined and the basic structure detailed. The specificat...

  15. Adapting and implementing the Wine in Moderation-Art de Vivre programme in Argentina

    Alvarez Natalia

    2014-01-01

    Full Text Available The Wine in Moderation-Art de Vivre (WIM programme was officially launched in 2008 as the wine sector's contribution to the EU Alcohol & Health Forum, within the framework of the EU strategy to support Member States in reducing alcohol-related harm. Building on the values of the “wine culture” and founded on information backed by science, broad education & self-regulation, WIM aims at promoting moderate and responsible behavior in the consumption of wine as a social and cultural norm. Considering the global and national trends in the wine market, the drinking patterns and the alcohol & health policy, Bodegas de Argentina (BAAC decided to mobilize concrete actions to contribute to the reduction of alcohol related harm, by adapting and implementing the WIM programme in Argentina. BAAC decided to engage the whole national wine value chain in the WIM programme and empower them with necessary knowledge and tools to properly implement it and disseminate the WIM message. The first step was to adapt the WIM programme and message to the cultural and linguistic context respecting the programme's common approach and creating an action plan with 2 main phases. With the Argentinean wine value chain participating in WIM and having the proper skills to do, the challenge now lies in reaching consumer. The successful implementation in Argentina has set a milestone in WIM's international development.

  16. Hybrid bayesian networks for traffic load models from weigh-in-motion data

    Morales-Nápoles, O.; Steenbergen, R.D.J.M.

    2012-01-01

    The Weigh-in-Motion (WIM) systems are used, among other applications, in pavement and bridge engineering, in infrastructure monitoring and assessment and inspection and reinforcement strategies. In the Netherlands and some other countries, the video-WIM system was implemented for pre-selection, and

  17. Key Qualifications in Work and Education.

    Nijhof, Wim J., Ed.; Streumer, Jan N., Ed.

    This book contains the following chapters: "The Demarcation Issue: Introduction" (Wim J. Nijhof, Jan N. Streumer); "Qualifying for the Future" (Wim J. Nijhof); "The Many Meanings of Occupational Competence and Qualification" (Per-Erik Ellstroem); "Qualification and Labour Markets: Institutionalisation and Individualisation" (Ben Hoevels); "The…

  18. Description and user manual of the WIMSLIC, FIXER and COMPA programs

    In this work the WIMS library and those WIMSLIC, FIXER and COMPA codes that are used to give him maintenance or to create a new one, the way to use them and its scopes are described. The objective of WIMSLIC, is the one of generating data nuclear with the WIMS code format, uses those results obtained with the NJOY system and the one POWR module, the FIXER function is to generate a new WIMS library or to already modify an existent, using the results of the one WIMSLIC code or those obtained with NJOY and WIMSR, while the COMPA function is to compare data groups with WIMS format. In the Appendix 1 the files of data are had that would use to generate a new library of WIMS that contain nuclear data for U-235, U-238, H-1, 0-16 and Al-27. In the Appendix 2 one has the listing of the programs before mentioned. (Author)

  19. The dependence of the global neutronic parameters on the fuel burnup for CANDU SEU43 core

    Balaceanu, V. [Institute for Nuclear Research, Pitesti (Romania); Pavelescu, M. [Academy of Romanian Scientists, Bucharest (Romania)

    2010-05-15

    In order to reduce the total fuel costs for the CANDU reactors, mainly by extending the fuel burnup limits, some fuel bundle concepts have been developed in different CANDU owner countries. Therefore, in our Institute the SEU43 (Slightly Enriched Uranium with 43 fuel elements) project was started in early '90s. The neutronic behavior analysis of the CANDU core with SEU43 fuel was an important step in our project design. The objective of this paper is to highline an analysis of the neutronic behavior of the CANDU SEU43 core with the fuel burnup. More exactly, the study refers to the dependence of some global neutronic parameters, mainly the reactivity, on the fuel burnup. Two types of CANDU core were taken into consideration: reference core (without any reactivity devices) and perturbed core (with a strong reactivity system inserted). The considered reactivity system is the Mechanical Control Absorber (MCA) one. The performed parameters are: k{sub eff.} values, the MCA reactivity worth and flux distributions. The fuel bundles in the core are SEU43, with the fuel enrichment in U{sup 235} of 0.96% and at nominal power. For the fuel burnup the values are: 0.00 GWd/tU (fresh fuel); 8.00 GWd/tU and 25.00 GWd/tU. For reaching this objective, a global neutronic calculation system named WIMSPIJXYZ LEGENTR is used. Starting from a 69-groups ENDF/B-V based library, this system uses three transport codes: (1) the standard lattice-cell code WIMS, for generating macroscopic cross sections in supercell option and also for burnup calculations; (2) the PIJXYZ code for 3D simulation of the MCA reactivity devices and the 3D correction of the macroscopic cross sections; (3) the LEGENTR 3D transport code for estimating global neutronic parameters (CANDU core). The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON and DIREN codes. This comparison shows a good agreement between these results. (orig.)

  20. 基于WIMS和MCNP耦合程序的医院中子照射器Ⅰ型堆燃耗计算%The burnup calculation for in-hospital neutron irradiator mark 1 reactor based on coupled code of WIMS and MCNP

    郭和伟; 江新标; 赵柱民; 陈立新; 张信一; 周永茂

    2012-01-01

    建立了基于WIMS和MCNP的临界-燃耗耦合计算方法,并对此方法进行了验算.通过栅元和组件问题的分析计算以及西安脉冲堆燃耗实验对比,验证了此耦合程序的可靠性和正确性.最后应用此耦合程序对医院中子照射器Ⅰ型堆的燃耗进行了计算和分析.%Numerical calculation for the equivalent surface source of the thermal neutron duct of in-hospital neutron irradiator mark 1 (IHNI-1) reactor is carried out using MCNP Monte Carlo code. Cold clean criticality of B-core is searched. Neutron beam parameters at the exit of thermal neutron duct are calculated. Equivalent neutron and -y surface sources for BNCT are built using equivalent surface source model. And these sources are reliable to calculate absorbed dose distribution in equivalent model of head quickly.

  1. Development of chemical flocculant for wastewater treatment

    Park, Jang Jin; Shin, J. M.; Lee, H. H.; Kim, M. J.; Yang, M. S.; Park, H. S

    2000-12-01

    Reagents 'KAERI-I and KAERI-II' which were developed as coagulants for industrial wastewater treatment in the study showed far superior performance to the existing inorganic coagulants such as Alum and Iron salt(FeSO4) when compared to their wastewater treatment performance in color and COD removal. Besides, it was not frozen at -25 deg C {approx} -30 deg C. When reagents 'KAERI-I and KAERI-II' were used as coagulant for wastewater treatment, the proper dosage was ranged from 0.1% to 0.5%(v/v) and proper pH range was 10.5 {approx} 11.5 in the area of alkaline pH.Reagents 'KAERI-I and KAERI-II' showed good performance with 95% or more removal of color-causing material and 60% or more removal of COD.

  2. Development of chemical flocculant for wastewater treatment

    Reagents 'KAERI-I and KAERI-II' which were developed as coagulants for industrial wastewater treatment in the study showed far superior performance to the existing inorganic coagulants such as Alum and Iron salt(FeSO4) when compared to their wastewater treatment performance in color and COD removal. Besides, it was not frozen at -25 deg C ∼ -30 deg C. When reagents 'KAERI-I and KAERI-II' were used as coagulant for wastewater treatment, the proper dosage was ranged from 0.1% to 0.5%(v/v) and proper pH range was 10.5 ∼ 11.5 in the area of alkaline pH.Reagents 'KAERI-I and KAERI-II' showed good performance with 95% or more removal of color-causing material and 60% or more removal of COD

  3. Film Premiere: Arrows of time at exploratorium

    2007-01-01

    "Join award-wimming artist and filmmaker Ken McMullen for a uniqude presentation of films engaging ideas at the forefront of science and culture on Sunday, Apri 29 at 2pm at the Exploratorium." (1/3 page)

  4. SPS tunnels fill up

    1974-01-01

    Here the installation of a 25 tons beam dump. Wim Middelkoop stands near the wall (2nd on the left); Peter Sievers stands on the left behind the magnet; Eric Geiger is 3rd along the wall on the right.

  5. Paris, Texas. Saksamaal ja sinu peas / Kairi Prints

    Prints, Kairi

    2010-01-01

    6.- 12. oktoobrini Tallinnas ja Tartus toimuval filmifestivalil "Uus Saksa Kino" saab vaadata Wim Wendersi muusikadokumentaale. "Film ja filosoofia" rubriigis koha- ja rahvusespetsiifika eksistentsist tänapäeva filmikunstis ja W. Wendersi filmist "Paris, Texas" (USA 1984)

  6. Use of new cross section data for uranium and boron isotopes obtained from recent ENDF/B-VI files

    Recently nuclear data for a few selected isotopes, as part of preliminary results of the IAEA WIMS Library Update Project derived from ENDF/B-VI, were made informally available. The data in WIMS library format was generated by them using the latest version of the NJOY code system by processing the ENDF/B-VI library. These data were incorporated in our WIMS69 data library. The eta values for U-235 are compared in Fig.1 in the energy range from 0.005 eV to 0.3 eV. The U-238 resonance integrals are now available up to 1100 deg. K compared to earlier tabulations of up to 900 deg. K. The new (WIMSKAL-88) B-10 cross sections in thermal energy range are seen to be higher by about 2% as compared to old WIMS data. 1 ref., 1 fig

  7. Teel tagasi laste juurde / Jaak Lõhmus

    Lõhmus, Jaak

    2005-01-01

    Cannes'i 58. filmifestivalilt : Jean-Pierre ja Luc Dardenne'i "Laps" ("L'Enfant"), Jim Jarmuschi "Murtud lilled" ("Broken Flowers"), Wim Wendersi "Ära tule koputama" ("Don't Come Knocking"), Michael Haneke "Varjatud" ("Cache")

  8. Water Well Locations - MO 2012 Certified Wells (SHP)

    NSGIC GIS Inventory (aka Ramona) — This data set provides information about wells that are certified by the State of Missouri. The parent data set is the Wellhead Information Management System (WIMS)...

  9. Automatic Traffic Recorder (ATR) Stations

    Department of Homeland Security — The data included in the GIS Traffic Stations Version database have been assimilated from station description files provided by FHWA for Weigh-in-Motion (WIM), and...

  10. Weigh-in-Motion Stations

    Department of Homeland Security — The data included in the GIS Traffic Stations Version database have been assimilated from station description files provided by FHWA for Weigh-in-Motion (WIM), and...

  11. Heating of the Warm Ionized Medium by Low-Energy Cosmic Rays

    Walker, Mark A

    2015-01-01

    In light of evidence for a high ionization rate due to Low-Energy Cosmic Rays (LECR), in diffuse molecular gas in the solar neighbourhood, we evaluate their heat input to the Warm Ionized Medium (WIM). LECR are much more effective at heating plasma than they are at heating neutrals. We show that the upper end of the measured ionization rates corresponds to a local LECR heating rate sufficient to maintain the WIM against radiative cooling, independent of the nature of the ionizing particles or the detailed shape of their spectrum. Elsewhere in the Galaxy the LECR heating rates may be higher than measured locally. In particular, higher fluxes of LECR have been suggested for the inner Galactic disk, based on the observed hard X-ray emission, with correspondingly larger heating rates implied for the WIM. We conclude that LECR play an important, perhaps dominant role in the thermal balance of the WIM.

  12. Heating of the Warm Ionized Medium by Low-energy Cosmic Rays

    Walker, Mark A.

    2016-02-01

    In light of evidence for a high ionization rate due to low-energy cosmic rays (LECR) in diffuse molecular gas in the solar neighborhood, we evaluate their heat input to the warm ionized medium (WIM). LECR are much more effective at heating plasma than they are at heating neutrals. We show that the upper end of the measured ionization rates corresponds to a local LECR heating rate sufficient to maintain the WIM against radiative cooling, independent of the nature of the ionizing particles or the detailed shape of their spectrum. Elsewhere in the Galaxy the LECR heating rates may be higher than those measured locally. In particular, higher fluxes of LECR have been suggested for the inner Galactic disk, based on the observed hard X-ray emission, with correspondingly larger heating rates implied for the WIM. We conclude that LECR play an important and perhaps dominant role in the thermal balance of the WIM.

  13. The FACADE version of JOSHUA

    JOSHUA is a computer program, which simulates the neutronics, hydraulics, burn-up and refuelling of water cooled reactors. Its neutronics data in the form of macroscopic cross-sections are normally prepared by a lattice cell program such as WIMS. The method of transfer of data from WIMS to JOSHUA has been simplified by the use of the FACADE system. This report describes the input to the version of JOSHUA that uses FACADE. (author)

  14. A new approach to the calculation of axial and radial neutron diffusion coefficients

    A new approach to the calculation of axial and radial diffusion coefficients is developed that introduces a new angle dependent 'transport cross-section'. Calculations are carried out using the programme WDSN(ST) for both AGR-type and SGHWR lattices. The results are compared with the standard method due to Benoist embodied in WIMS and, in the AGR case, with Monte Carlo results. The comparisons suggest major errors in the WIMS results for SGHWR clusters in the axial direction. (author)

  15. Current Status of Research Activities Related to THM-Coupled Processes in Buffer

    Choi, Heui Joo; Lee, Changsoo; Choi, Young Chul; Lee, Minsoo; Kim, Jin Seop [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For the purpose of enhancing the understanding of THM-coupled behavior in and around buffer, a computer code, KAERI-SIMULATOR, is being developed and verified by participating in Decovalex-2015 project. The THM data collected from this facility will be used to validate the KAERI-SIMULATOR.

  16. Current Status of Research Activities Related to THM-Coupled Processes in Buffer

    For the purpose of enhancing the understanding of THM-coupled behavior in and around buffer, a computer code, KAERI-SIMULATOR, is being developed and verified by participating in Decovalex-2015 project. The THM data collected from this facility will be used to validate the KAERI-SIMULATOR

  17. Feedback Experience from Decommissioning of Uranium Conversion Plant

    KAERI has been conducting decommissioning activities of Uranium Conversion Plant (UCP) for the last decade. As a result of all this work KAERI has accumulated significant experience in the field of decommissioning of nuclear facilities. On the basis of the experience gained from decommissioning activities, this paper describes several lessons learned

  18. Status of ADS research in Korea

    KAERI has been working on ADS since 1997. The KAERI ADS system is called HYPER (HYbrid Power Extraction Reactor). HYPER research started as a 10 year nuclear research program funded by the government. The ADS research of KAERI consists of 3 stages. A basic concept of HYPER was established in the first stage (1997 - 2000) of the development. The basic technology related to HYPER was investigated in the second stage (2001 - 2003) while upgrading the design. The third stage of research started in March, 2004. The conceptual design of HYPER core will be completed in the third stage (2004 - 2006). The investigation of key technologies will be continued in the third stage. The conceptual design of HYPER core was almost finished in the second stage. The upgrade of core design and transient study will be done in the third stage. Regarding experimental research, fuel and Pb-Bi study were performed during the second stage. U surrogate fuel was fabricated and tested. KAERI joined MEGAPIE project in 2001 for Pb-Bi research. KAERI also installed the static Pb-Bi corrosion test device in 2003 and started experiments. KAERI will complete the construction of Pb-Bi corrosion loop in 2004. I-NERI program related to lead-alloy experiment will be launched in 2004. For the fuel/FP target study, KAERI plans to perform fission product irradiation test using KAERI's research reactor HANARO

  19. Use of Finite Elements Analysis for a Weigh-in-Motion Sensor Design

    Viorel Goanta

    2012-05-01

    Full Text Available High speed weigh-in-motion (WIM sensors are utilized as components of complex traffic monitoring and measurement systems. They should be able to determine the weights on wheels, axles and vehicle gross weights, and to help the classification of vehicles (depending on the number of axles. WIM sensors must meet the following main requirements: good accuracy, high endurance, low price and easy installation in the road structure. It is not advisable to use cheap materials in constructing these devices for lower prices, since the sensors are normally working in harsh environmental conditions such as temperatures between –40 °C and +70 °C, dust, temporary water immersion, shocks and vibrations. Consequently, less expensive manufacturing technologies are recommended. Because the installation cost in the road structure is high and proportional to the WIM sensor cross section (especially with its thickness, the device needs to be made as flat as possible. The WIM sensor model presented and analyzed in this paper uses a spring element equipped with strain gages. Using Finite Element Analysis (FEA, the authors have attempted to obtain a more sensitive, reliable, lower profile and overall cheaper elastic element for a new WIM sensor.

  20. Numerical simulations of a propeller wake impacting

    Hi, M.; Veitch, B.; Bose, N. [Memorial Univ. of Newfoundland, Faculty of Engineering and Applied Science, St. John' s, Newfoundland (Canada)]. E-mail: moqinhe@engr.mun.ca; Bruce, C.; Liu, P. [National Research Council Canada, Inst. for Ocean Technology, St. John' s, Newfoundland (Canada)

    2005-07-01

    This paper introduces a newly developed Wake Impingement Model (WIM) that aims to simulate of the dynamic loads induced by a three dimensional, unsteady, and strong vortical propeller wake. Simulations of loads on an ice class, tractor type podded propeller in straight ahead motion are presented consisting of mean loads on the propeller and side force on the pod and strut. The side force fluctuations for three different advance coefficients have also been predicted. These simulations were carried out by using a panel code, PROPELLA, with or without WIM. Simulated results were compared with and without WIM and with experimental data. The comparison of the propeller open water characteristics of two simulated results shown there is almost no difference between predictions with and without WIM. It was found by comparing with experimental data that the simulations of the side force on the pod and strut with WIM successfully captured the fluctuation which was dominated by the component at the blade passing frequency, although this was at a reduced level compared with the measurements. (author)

  1. Participation in the 1999 IAEA interlaboratory comparison on chemical analysis of groundwater

    Joe, Kih Soo; Choi, Kwang Soon; Han, Sun Ho; Suh, Moo Yul; Park, Kyung Kyun; Choi, Ke Chun; Kim, Won Ho

    2000-08-01

    KAERI analytical laboratory participated in the 1999 IAEA interlaboratory comparison on chemical analysis of groundwater organized by IAEA Hydrology Laboratory(RAS/8/084). 13 items such as pH, electroconductivity, HCO{sub 3}, Cl, SO{sub 4}, NO{sub 3}, SiO{sub 2}, B, Li, Na, K, Ca, Mg were analyzed. The result of this program showed that KAERI laboratory was ranked within 10% range from top level. An analytical expert in KAERI attended the 'Consultants' Meeting' at IAEA headquater and prepared the guideline for chemical analysis of groundwater.

  2. Development of total quality management technology for the establishment of better-quality R and D system

    The existing total quality management technology(TGMT) has been surveyed to establish R and D quality system by upgrading such TGMT and deploying the concept of quality management. However, there has been a variety of differences on the quality management environments between R and D institutes in advanced countries and KAERI. So the case studies of quality management in advanced countries and those in domestic corporations have been analyzed. Finally, based on this analysis, the questionnaire for all KAERI staff-members has been drawn up for applying TQM to KAERI. 4 tabs., 13 figs., 21 refs. (Author)

  3. Quality assurance for the research and development of nuclear technology

    To establish the KAERI-wide quality system, the KAERI QA program was developed. Internal evaluation was performed to determine the current status of quality activities. Study of the advanced quality technology such as QA in R and D , design QA, reliability, software QA was performed for the future application. The KAERI-wide quality system should be implemented as proposed in this report and will be improved through evaluation of the quality and application of the advanced quality technology to have the optimum quality system. (Author)

  4. Participation in the 1999 IAEA interlaboratory comparison on chemical analysis of groundwater

    KAERI analytical laboratory participated in the 1999 IAEA interlaboratory comparison on chemical analysis of groundwater organized by IAEA Hydrology Laboratory(RAS/8/084). 13 items such as pH, electroconductivity, HCO3, Cl, SO4, NO3, SiO2, B, Li, Na, K, Ca, Mg were analyzed. The result of this program showed that KAERI laboratory was ranked within 10% range from top level. An analytical expert in KAERI attended the 'Consultants' Meeting' at IAEA headquater and prepared the guideline for chemical analysis of groundwater

  5. Preclinical Study for Application of Fabricated High Activity Ir-192

    This study was performed to evaluate the feasibility and safety of high activity Ir-192 sources manufactured by KAERI(Korea Atomic Energy Research Institute) for application to present equipment such as various applicators inserted to patients and PLATO(Nucletron, Netherland) of treatment planning system and to evaluate safety and accuracy of Ir-192 as practical clinic use through in vitro dosimetry of Ir-192. We confirmed the physical and radiobiological safety of KAERI sources to use practical. KAERI sources are applicable to commercial high dose rate brachytherapy machine safely. Then those can be substituted for the imported sources such as sources made by Nucletron, Gammamed and exported to the foreign country

  6. Evidence for dust emission in the Warm Ionised Medium using WHAM data

    Lagache, G; Reynolds, R J; Tufte, S L

    1999-01-01

    We have used the WHAM H_alpha survey and Leiden/Dwingeloo HI data to decompose the Far-infrared emission (from 100 to 1000 micron) at high Galactic latitude into components associated with the Warm Ionised Medium (WIM) and the Warm Neutral Medium (WNM). This decomposition is possible for the first time thanks to preliminary WHAM data that cover a significant fraction of the sky (about 10%). We confirm the first detection of dust emission from the WIM (Lagache et al. 1999) and show that the WIM dust temperature and emissivity are very similar to those in the WNM. The analysis suggests moreover that about 25% of the far-IR dust emission at high galactic latitude is uncorrelated with the HI gas. The decomposition again reveals a Cosmic Far-Infrared Background (CFIRB) which is determined for the first time from 100 to 1000 micron using two independant gas tracers.

  7. Code package of the physics calculation and fuel management of uranium hydride zirconium reactor

    The code package of uranium hydride zirconium reactor physics calculation is established. considering the thermalization of H in ZrH, the nuclear data of H in ZrH in WIMS library pattern are provided and WIMS-N2 library is obtained. The cell parameters are calculated using WIMS-D/4 code and SIMS-N2 library. The diffusion calculation is performed using CITATION code and SIXTUS-2 code. The in-core fuel management code XPR-ICFM is obtained on the basis of SIXTUS-2 code. To check the accuracy and reliability of the code package, the critical keff and the value of the control rod of abroad TRIGA, the pulsed reactor in china are calculated. The results are satisfied

  8. Analysis of heavy water lattice experiments on research reactors for testing nuclear data

    There is a need for updated multigroup libraries for lattices codes of WIMS type for PHWR reactors calculations. Different multigroup libraries are used with WIMS and other codes, but these libraries are not normally updated to the level of last revision of ENDF/B-VI and other evaluated nuclear data files. Then, a special attention to the application of new WIMS libraries on PHWR calculations is justified. Some research and development activities associated to PHWR type of reactors, that need updated nuclear libraries of WIMS type, are: use of slightly enriched uranium (SEU cycle), use of UO2-ThO2 fuels, use of burnable poisons mixed in fuel pellets (UO2-Gd2O3) and absorber rods, new types of fuel elements (in Argentina: CARA Project-Advanced Fuel for Argentine Reactors) Taking into account the need of new WIMS libraries associated to these activities, a set of benchmarks have been identified and coded for PHWR lattice calculations.. The experimental benchmarks are identified with the name of the facility or research reactor where the measurements were carried out. The main references for this type of benchmarks is the ZED-2 Canadian reactor and DCA Japanese reactor. This work cover benchmark results of the following cases: ZED-2 analysis: experiments with 37 and 28 CANDU-type rod Fuel Clusters and lattice experiments with 19-rod Clusters with ThO2-UO2 Fuel; DCA analysis: Evaluation of Neutronic Parameters in Heavy Water and Slightly Enriched Uranium UO2 Fuel (28-rod Cluster) and critical experiments on Gadolinium poisoned cluster-type fuel assemblies of 54 rods in heavy water lattices of DCA facility. For several cases, results are included for different pitches and coolants. The parameters analysed are: k-effective with experimental bucklings, fast fission ratio [U-238 fissions/U-235 fissions], relative conversion ratio [U-238 captures/U-235 fissions], U-235 fission rate distribution, Cu-63 absorption rate distribution, Lutetium-Manganese activity ratio, ratio of

  9. FLYSAFE, nowcasting of in flight icing supporting aircrew decision making process

    Drouin, A.; Le Bot, C.

    2009-09-01

    FLYSAFE is an Integrated Project of the 6th framework of the European Commission with the aim to improve flight safety through the development of a Next Generation Integrated Surveillance System (NGISS). The NGISS provides information to the flight crew on the three major external hazards for aviation: weather, air traffic and terrain. The NGISS has the capability of displaying data about all three hazards on a single display screen, facilitating rapid pilot appreciation of the situation by the flight crew. Weather Information Management Systems (WIMS) were developed to provide the NGISS and the flight crew with weather related information on in-flight icing, thunderstorms, wake-vortex and clear-air turbulence. These products are generated on the ground from observations and model forecasts. WIMS supply relevant information on three different scales: global, regional and local (over airport Terminal Manoeuvring Area). Within the flysafe program, around 120 hours of flight trials were performed during February 2008 and August 2008. Two aircraft were involved each with separate objectives : - to assess FLYSAFE's innovative solutions for the data-link, on-board data fusion, data-display, and data-updates during flight; - to evaluate the new weather information management systems (in flight icing and thunderstorms) using in-situ measurements recorded on board the test aircraft. In this presentation we will focus on the in-flight icing nowcasting system developed at Météo France in the framework of FLYSAFE: the local ICE WIMS. The local ICE WIMS is based on data fusion. The most relevant information for icing detection is extracted from the numerical weather prediction model, the infra-red and visible satellite imagery and the ground weather radar reflectivities. After a presentation of the local ICE WIMS, we detail the evaluation of the local ICE WIMS performed using the winter and summer flight trial data.

  10. Korea signs for 2nd CANDU at Wolsong

    The sale of a second CANDU 6 reactor to Korea for the Wolsong site is discussed in relation to nuclear power in Korea, the Korean economy generally, Canadian trade with Korea, and cooperation between AECL and KAERI

  11. Study on upgrade on nuclear control related open source information website

    The open source information relevant to the nuclear control is regularly collected, analyzed, and published to the three web sites by the Technology Center for Nuclear Control (TCNC) of the Korea Atomic Energy Research Institute (KAERI). These web sites are world-wide, KAERI-wide, and TCNC-wide, respectively. We are to upgrade the KAERI-wide website to the access-controlled world-wide web site with some additional functionality. In this research, the current status of the three nuclear control related open source information websites managed by the TCNC was introduced and methods for upgrading the KAERI-wide open source information website and associated information security technology were reviewed

  12. Process Faults Analysis and Design Considerations for Pyroprocess Hot Cell Safety

    You, G. S.; Choung, W. M.; Ku, J. H.; Moon, S. I.; Kim, H. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    KAERI (Korea Atomic Energy Research Institute) has been studied the pyroprocess since 1997. For demonstration of pyroprocess, KAERI developed two facilities, the ACPF (Advanced spent fuel Conditioning Process Facility) and the PRIDE (PyRoprocess-Integrated inactive DEmonstration facility). From 2013 KAERI performs a pre-conceptual design of the ESPF (Engineering-Scale Pyroprocess Facility). In this paper, the process faults analysis and design considerations for pyroprocess hot cell safety are described. KAERI has been developing a pyroprocess for conditioning and reutilization of PWR spent nuclear fuels. The safety evaluations of the pyroprocess facilities were performed to confirm the safe design. The process safety as one of the safety evaluations was analyzed by the faults tree method. The corresponding safe design considerations for each fault type were also considered.

  13. Development of technical information processing systems

    The major goal of this project is to develop a more efficient information management system by connecting the KAREI serials database which enable the users to access from their own laboratory facilities through KAREI-NET. The importance of this project is to make the serials information of KAERI easily accessible to users as valuable resources for R and D activities. The results of the project are as follows. 1) Development of the serials database and retrieval system enabled us to access to the serials holding information through KAERI-NET. 2) The database construction establishes a foundation for the management of 1,600 serials held in KAERI. 3) The system can be applied not only to KAERI but also to similar medium-level libraries. (Author)

  14. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  15. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  16. Englemøtet over Berlin

    Ørjasæter, Kristin

    2006-01-01

    This reading of Dag Solstads novel 16.07.41 (2002) is focussed on the presentation of Berlin in the text. The article also a discusses the angel-theme that can be find in this novel as well as in Wim Wenders Der himmel über Berlin (1987).......This reading of Dag Solstads novel 16.07.41 (2002) is focussed on the presentation of Berlin in the text. The article also a discusses the angel-theme that can be find in this novel as well as in Wim Wenders Der himmel über Berlin (1987)....

  17. Utilities programs for the WIMSD4 code

    The WIMSD4 code is widely known around the world. For its better use, it is convenient to count with auxiliary programs. Two of these programs, developed in FORTRAN 77, in the VAX computer of the Bariloche Atomic Center, are herein presented. WINTER (Wims INTERactive) to generate input data of WIMSD4 in an interactive way, and AMICO (Anisn MIx and COndense) to deal with cross sections data of a multigroup data library and of WIMS output to be used in other programs, such as: ANISN, DOT, CITATION, DIPOBAR, etc. (Author)

  18. An outstanding guest

    2004-01-01

    The person celebrating CERN's 25th anniversary whose photograph we published last week was not just any guest. Readers have pointed out that it was Wim Klein, whose remarkable abilities are part of CERN's history. Well known as one of the best "human calculators", Wim was recruited by CERN in 1957 to verify computer programs, which at the time were still stumbling. Moreover, he regularly beat their speed in calculating and gave breath-taking demonstrations. During one such demonstration in September 1973, he calculated the 19th root of a 133-digit number in less than 2 minutes !

  19. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Muhammad Atta; Iqbal Masood; Mahmood Tayyab

    2011-01-01

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determin...

  20. Knowledge Preservation and Data Collection on FR in Republic of Korea

    Nuclear technology archives project at KAERI: Object & Expected Effects - • Establish the nuclear knowledge archives system: → Archive nuclear knowledge from 1959 to now; → Provide the base of progress and development of nuclear technology; • Promote to transfer and share the nuclear knowledge of KAERI for 50 years: → Transfer nuclear knowledge, lessons and experiences to the younger generation; → Facilitate knowledge sharing across projects and across departments

  1. Establishment of ANSI N13.11 X-ray radiation fields for personal dosimetry performance test by computation and experiment.

    Kim, J L; Kim, B. H.(Seoul National University, 151-742, Seoul, South Korea); Chang, S Y; J. K. Lee

    1997-01-01

    This paper describes establishment by computational and experimental methods of the American National Standard Institute (ANSI) N13.11 X-ray radiation fields by the Korea Atomic Energy Research Institute (KAERI). These fields were used in the standard irradiations of various personal dosimeters for the personal dosimetry performance test program performed by the Ministry of Science and Technology of Korea in the autumn of 1995. Theoretical X-ray spectra produced from two KAERI X-ray generator...

  2. Country presentation

    The development of the Web-Portal for the ANENT is one of the action plans under activity 1 led by KAERI, ROK. The web-based networking is intended to establish an effective and sustainable focal point. KAERI had developed the Web-Portal for the ANENT and ANENT members were requested to evaluate the Web-Portal. Malaysia has applied SWOT analysis to evaluate the Web-Portal, analyzing the strengths, weaknesses, opportunities, threats, challenges and the way forward

  3. A survey on the status of ATM based LAN

    This report presents the technical status of the ATM(Asynchronous Transfer Mode) as a new high speed data communication method. Since the FDDI(Fiber optic Distributed Data Interchange) backbone has been installed in september 1995, it has been used as a main network structure of KAERI. However, recently high speed and multimedia data communication environment is being required to accommodate the recent trend of the network usage in KAERI. For example, the rapid growth of Internet usage and increased activities of information retrieval systems on KAERI-Net demand more effective network system. Chapter 1 discusses the status of KAERI-Net and the selection criteria of a network model according to the national plan of super high speed network structure. In Chapter 2, the basic concept of ATM such as communication method and communication structure is studied, and Chapter 3 presents the overall concepts of standard model of ATM. In Chapter 4, we survey the recent trend of technical development of ATM and analyze the status of ATM technology. As a concluding remark, Chapter 5 discusses the criteria and check points for optimal design of KAERI-Net backbone. This report will be used as a technical reference for the installation of ATM in KAERI-Net. 10 tabs., 32 figs., 11 refs. (Author)

  4. Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

    The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% 235U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% 235U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO2 powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report

  5. Neutronics and thermal hydraulics feedback models of the Harwell materials testing reactors DIDO and PLUTO: I Neutronics analysis

    Neutronics modelling of the Harwell MTRs DIDO and PLUTO has been achieved in the WIMS-E framework using (r,z) and (x,y) two dimensional diffusion theory. The modelling takes into account fuel burnup and the presence of the coarse control arms and experimental rigs. The modelling is validated by comparisons with measurements of thermal and fast flux distributions. (author)

  6. Neutronics and thermal hydraulics modelling of the Harwell Materials Testing Reactors DIDO and PLUTO

    A detailed 2-D cylindrical diffusion theory neutronics model is presented for the Harwell reactors DIDO and PLUTO, based on the WIMS-E program. The model for these highly asymmetric reactors allows for the presence of the various control systems, experimental rigs and fuel burnup. Comparisons made with measurements of burnup and of radial and axial flux distributions validate the approach. (author)

  7. Analysis of axle and vehicle load properties through Bayesian networks based on weigh-in-motion data

    Morales Napoles, O.; Steenbergen, R.D.J.M.

    2014-01-01

    Weigh-in-Motion (WIM) systems are used, among other applications, in pavement and bridge reliability. The system measures quantities such as individual axle load, vehicular loads, vehicle speed, vehicle length and number ofaxles. Because ofthe nature ofúamc configuration, the quantities measured are

  8. Carbon storage and flow: Their relationship to water, soil and land management

    This article is prompted by the paper by Wim G. Sombroek, Freddy O. Nachtergaele and Axel Hebel on 'Amounts, dynamics and sequestering of carbon in tropical and subtropical soils', and the paper by Malin Falkenmark and John Rockstroem on 'Curbing rural exodus from tropical drylands', p.417-437 in this issue of Ambio. 12 refs

  9. Eestlased kaotasid Telekomi aktsiatega kaks ja pool miljardit / Erkki Erilaid

    Erilaid, Erkki

    2001-01-01

    Eesti Telekomi aktsia püstitas uue odavusrekordi, kukkudes päevaga 6,35% 44,25 kroonini. Diagramm: Eesti Telekom taas rekordmadal. Vt. samas: Kangus, Aivo. Rasked valikud. Arvamust avaldavad: Wim Duisenberg, Siim Kallas ja Horst Koehler

  10. Excitation of the diffuse ionized gas in galaxies

    Crystal L. Martin

    2000-01-01

    Full Text Available Se revisan las propiedades del gas ionizado difuso (DIG en galaxias cercanas con formaci on estelar, enfatizando la evidencia para un mecanismo de excitaci on adicional a la fotoionizaci on estelar. Discuto la pregunta frecuentemente hecha sobre si el DIG en galaxias externas es similar al medio ionizado tibio (WIM descrito en esta reuni on por Reynolds y Ha ner.

  11. Excitation of the diffuse ionized gas in galaxies

    Martin, Crystal L.

    2000-01-01

    Se revisan las propiedades del gas ionizado difuso (DIG) en galaxias cercanas con formaci on estelar, enfatizando la evidencia para un mecanismo de excitaci on adicional a la fotoionizaci on estelar. Discuto la pregunta frecuentemente hecha sobre si el DIG en galaxias externas es similar al medio ionizado tibio (WIM) descrito en esta reuni on por Reynolds y Ha ner.

  12. Notitie over het uitplanten van korstmossen

    Geraerdts, W.H.J.M.; Ketner-Oostra, H.G.M.

    2006-01-01

    In de voormalige tuin van Wim Geraedts (Wijchen, Gelderland) blijkt het uitplanten van korstmossen een sukses te zijn. Het gaat om enkele Cladonia-soorten, een Cladina- en twee Peltigera-soorten. Het biotoop waarin ze uitgezet zijn, ligt op een wal van heideplaggen om een kunstmatig hoogveentje met

  13. Cross sections for fuel depletion and radioisotope production calculations in TRIGA reactors

    For TRIGA Reactors, the fuel depletion and isotopic inventory calculations, depends on the computer code and in the cross sections of some important actinides used. Among these we have U-235, U-238, Pu-239, Pu-240 and Pu-241. We choose ORIGEN2, a code with a good reputation in this kind of calculations, we observed the cross sections for these actinides in the libraries that we have (PWR's and BWR), the fission cross section for U-235 was about 50 barns. We used a PWR library and our results were not satisfactory, specially for standard elements. We decided to calculate cross sections more suitable for our reactor, for that purpose we simulate the standard and FLIP TRIGA cells with the transport code WIMS. We used the fuel average flux and COLAPS (a home made program), to generate suitable cross sections for ORIGEN2, by collapsing the WIMS library cross sections of these nuclides. For the radioisotope production studies using the Central Thimble, we simulate the A and B rings and used the A average flux to collapse cross sections. For these studies, the required nuclides sometimes are not present in WIMS library, for them we are planning to process the ENDF/B data, with NJOY system, and include the cross sections to WIMS library or to collapse them using the appropriate average-flux and the program COLAPS. (author)

  14. The numerical benchmark CB2-S, final evaluation

    In this paper are final results of numerical benchmark CB2-S compared (activity, gamma and neutron sources, concentration of important nuclides and decay heat). The participants are: Vladimir Chrapciak (SCALE), Ludmila Markova (SCALE), Svetlana Zabrodskaja (SCALA), Pavel Mikolas (WIMS). Eva Tinkova (HELIOS) and Maria Manolova (SCALE) (Authors)

  15. Cover Image, Volume 170A, Number 6, June 2016.

    Bayat, Allan; Fijalkowski, Igor; Andersen, Tobias; Abdulmunem, Sura Azhar; van den Ende, Jenneke; Van Hul, Wim

    2016-06-01

    The cover image, by Wim Van Hul et al., is based on the Original Article Further delineation of facioaudiosymphalangism syndrome: Description of a family with a novel NOG mutation and without hearing loss, DOI: 10.1002/ajmg.a.37626. PMID:27191530

  16. MINOSSE R301/02-03. Nuclear analysis

    This report contains the results of the WIMS6 nuclear calculations performed for the experiments MINOSSE R301-02 and 03, loaded in a standard TRIO capsule together with two aluminium dummies in HFR positions C3 and E7 respectively. (orig.)

  17. Interface Screenings

    Thomsen, Bodil Marie Stavning

    2015-01-01

    In Wim Wenders' film Until the End of the World (1991), three different diagrams for the visual integration of bodies are presented: 1) GPS tracking and mapping in a landscape, 2) video recordings layered with the memory perception of these recordings, and 3) data-created images from dreams and m...

  18. Knowledge Management.

    1999

    The first of the four papers in this symposium, "Knowledge Management and Knowledge Dissemination" (Wim J. Nijhof), presents two case studies exploring the strategies companies use in sharing and disseminating knowledge and expertise among employees. "A Theory of Knowledge Management" (Richard J. Torraco), develops a conceptual framework for…

  19. Ion-Neutral Collisions in the Interstellar Medium: Wave Damping and Elimination of Collisionless Processes

    Spangler, Steven R.; Savage, Allison H.; Redfield, Seth

    2011-09-01

    Most phases of the interstellar medium contain neutral atoms in addition to ions and electrons. This introduces differences in plasma physics processes in those media relative to the solar corona and the solar wind at a heliocentric distance of 1 astronomical unit. In this paper, we consider two well-diagnosed, partially-ionized interstellar plasmas. The first is the Warm Ionized Medium (WIM) which is probably the most extensive phase in terms of volume. The second is the gas of the Local Clouds of the Very Local Interstellar Medium (VLISM). Ion-neutral interactions seem to be important in both media. In the WIM, ion-neutral collisions are relatively rare, but sufficiently frequent to damp magnetohydrodynamic (MHD) waves (as well as propagating MHD eddies) within less than a parsec of the site of generation. This result raises interesting questions about the sources of turbulence in the WIM. In the case of the VLISM, the ion-neutral collision frequency is higher than that in the WIM, because the hydrogen is partially neutral rather than fully ionized. We present results showing that prominent features of coronal and solar wind turbulence seem to be absent in VLISM turbulence. For example, ion temperature does not depend on ion mass. This difference may be due to ion-neutral collisions, which distribute power from more effectively heated massive ions such as iron to other ion species and neutral atoms.

  20. Study of power distribution in the CZP, HFP and normal operation states of VVER-1000 (Bushehr) nuclear reactor core by coupling nuclear codes

    Highlights: • Simulation of one-sixth of VVER-1000 reactor core by WIMS-D4 based on core symmetry. • Obtaining the cross sections of some nuclides by WIMS-D4 from BOC to EOC. • Transferring the obtained cross sections into CITATION as inputs for codes coupling. • Obtaining neutron fluxes and power by CITATION and program cycle in the CZP and HFP. • Distribution depiction of neutron fluxes and power in CZP, HFP and normal operation. - Abstract: In this research, the simulation of one-sixth of VVER-1000 (Bushehr) reactor core is carried out by WIMS-D4 nuclear code, based on symmetry of core and also by information obtained from FSAR. The cross sections of some nuclides are obtained by WIMS-D4 from the beginning of cycle (BOC) to the end of cycle (EOC), and they are transferred into the CITATION code as inputs. In the next stage, the amounts of neutron fluxes and power of reactor core are obtained by CITATION code in the CZP and HFP states. Then, the received products are returned again into the extended program cycle, thereby distributions of neutron fluxes and power are finally depicted. In the meantime, the space distribution of neutron fluxes and power throughout the core are presented during the normal operation by this simulation. It can be inferred that if the reactor operation continues, a flat power distribution will be made in the reactor core that might cause maximum power

  1. Näitused / Mari Sobolev

    Sobolev, Mari, 1968-

    1998-01-01

    Tallinna XI graafikatriennaal. Rotermanni soolaladu: "Puudutus"; Kunstihoone: "Puutujad"; Rahvusraamatukogu: Laste ja noorte graafikanäitus; Kastellaanimaja galerii: Wim Lamboo (s. 1949) "Eesti kunstnike portreed 1997" (fotograafia); Tallinna Botaanikaaed: "Maised rõõmud"; Rahvusraamatukogu: Antanas Sutkus (s. 1939) "Pro memoria" (fotograafia); Tammsaare majamuuseum: Eve Pärnaste "Ülelend" (maalid); Haapsalu Linnagalerii: Paul Allik "Seisundid" (aktimaalid).

  2. Rock-hard coatings

    Muller, M.

    2007-01-01

    Aircraft jet engines have to be able to withstand infernal conditions. Extreme heat and bitter cold tax coatings to the limit. Materials expert Dr Ir. Wim Sloof fits atoms together to develop rock-hard coatings. The latest invention in this field is known as ceramic matrix composites. Sloof has sign

  3. Sorteerwater gereinigd met ozon

    PPO Fruit,

    2009-01-01

    Energie uit de lucht wordt het tegenwoordig ook wel genoemd. Futuroloog Wim de Ridder weet het zeker : uit zonne-energie gehaalde elektriciteit wordt steeds goedkoper. Over zo'n tien jaar komt er een omslag en wordt het mogelijk stroom voor een habbekrats te produceren

  4. Project of the supporting information technology for the nuclear research (year 2009{approx}2011)

    Song, Tai Gil; Kang, Sin Book; Sohn, Jae Min; Kim, Jin Hee; Hwang, Hye Seon; Mun, Dong Seop; Ko, Young Cheol

    2011-12-15

    These benefits can be obtained by effectively and efficiently maximizing the information conveyed by your data and ensuring that the data are of the highest quality. Therefore, continuous update of MIS(Management Information System) is needed. The KAERI's representative web site(http://www.kaeri.re.kr) which is supporting the web services for a nation is doing the role of an on line external contact point. But the contents of the web site was required the overall editing because it's building was a long time. The KAERI Net was separated into an intranet and an extranet at the end of 2009. For this reason, we need a convenient data transfer system between intranet and extranet. The major project scope is same as follows. o The development and management of the documents for the electronic approval system o The re construction and management of the KAERI's representative web site o The development of electronic data transfer system o A study on an emergency recovery training and the efficient utilizations of the KAERI electronic mail system Improve the web accessibility of web site with a focus on business plans, and to improve the overall content of the site was performed. To increase the satisfaction of users outside and to provide services that can improve the phase of the KAERI, the established new web site is need for the continued interest and management. Emergency recovery training has been successfully carried out. We hope that the results of this training will be used for the stable operation of the KAERI's e mail systems, and that the improvement plans will also be used for efficient utilization.

  5. Project of the supporting information technology for the nuclear research (year 2009∼2011)

    These benefits can be obtained by effectively and efficiently maximizing the information conveyed by your data and ensuring that the data are of the highest quality. Therefore, continuous update of MIS(Management Information System) is needed. The KAERI's representative web site(http://www.kaeri.re.kr) which is supporting the web services for a nation is doing the role of an on line external contact point. But the contents of the web site was required the overall editing because it's building was a long time. The KAERI Net was separated into an intranet and an extranet at the end of 2009. For this reason, we need a convenient data transfer system between intranet and extranet. The major project scope is same as follows. o The development and management of the documents for the electronic approval system o The re construction and management of the KAERI's representative web site o The development of electronic data transfer system o A study on an emergency recovery training and the efficient utilizations of the KAERI electronic mail system Improve the web accessibility of web site with a focus on business plans, and to improve the overall content of the site was performed. To increase the satisfaction of users outside and to provide services that can improve the phase of the KAERI, the established new web site is need for the continued interest and management. Emergency recovery training has been successfully carried out. We hope that the results of this training will be used for the stable operation of the KAERI's e mail systems, and that the improvement plans will also be used for efficient utilization

  6. The relationship between fatty acid profiles in milk identified by Fourier transform infrared spectroscopy and onset of luteal activity in Norwegian dairy cattle.

    Martin, A D; Afseth, N K; Kohler, A; Randby, Å; Eknæs, M; Waldmann, A; Dørum, G; Måge, I; Reksen, O

    2015-08-01

    To investigate the feasibility of milk fatty acids as predictors of onset of luteal activity (OLA), 87 lactations taken from 73 healthy Norwegian Red cattle were surveyed over 2 winter housing seasons. The feasibility of using frozen milk samples for dry-film Fourier transform infrared (FTIR) determination of milk samples was also tested. Morning milk samples were collected thrice weekly (Monday, Wednesday, Friday) for the first 10 wk in milk (WIM). These samples had bronopol (2-bromo-2-nitropropane-1,3-diol) added to them before being frozen at -20°C, thawed, and analyzed by ELISA to determine progesterone concentration and the concentrations of the milk fatty acids C4:0, C14:0, C16:0, C18:0, and cis-9 C18:1 as a proportion of total milk fatty acid content using dry-film FTIR, and averaged by WIM. Onset of luteal activity was defined as the first day that milk progesterone concentrations were >3 ng/mL for 2 successive measurements; the study population was categorized as early (n=47) or late (n=40) OLA, using the median value of 21 DIM as the cutoff. Further milk samples were collected 6 times weekly, from morning and afternoon milkings, these were pooled by WIM, and one proportional sample was analyzed fresh for fat, protein, and lactose content by the dairy company Tine SA, using traditional FTIR spectrography in the wet phase of milk. Daily energy-balance calculations were performed in 42 lactations and averaged by WIM. Animals experiencing late OLA had a more negative energy balance in WIM 1, 3, 4, and 5, with the greatest differences been seen in WIM 3 and 4. A higher proportion of the fatty acids were medium chained, C14:0 and C16:0, in the early than in the late OLA group from WIM 1. In WIM 4, the proportion of total fatty acid content that was C16:0 predicted late OLA, with 74% sensitivity and 80% specificity. The long-chain proportion of the fatty acids C18:0 and cis-9 C18:1 were lower in the early than in the late OLA group. Differences were greatest in

  7. Warm ionized gas in CALIFA early-type galaxies. 2D emission-line patterns and kinematics for 32 galaxies

    Gomes, J. M.; Papaderos, P.; Kehrig, C.; Vílchez, J. M.; Lehnert, M. D.; Sánchez, S. F.; Ziegler, B.; Breda, I.; Dos Reis, S. N.; Iglesias-Páramo, J.; Bland-Hawthorn, J.; Galbany, L.; Bomans, D. J.; Rosales-Ortega, F. F.; Cid Fernandes, R.; Walcher, C. J.; Falcón-Barroso, J.; García-Benito, R.; Márquez, I.; Del Olmo, A.; Masegosa, J.; Mollá, M.; Marino, R. A.; González Delgado, R. M.; López-Sánchez, Á. R.; Califa Collaboration

    2016-04-01

    Context. The morphological, spectroscopic, and kinematical properties of the warm interstellar medium (wim) in early-type galaxies (ETGs) hold key observational constraints to nuclear activity and the buildup history of these massive, quiescent systems. High-quality integral field spectroscopy (IFS) data with a wide spectral and spatial coverage, such as those from the CALIFA survey, offer an unprecedented opportunity for advancing our understanding of the wim in ETGs. Aims: This article centers on a 2D investigation of the wim component in 32 nearby (≲150 Mpc) ETGs from CALIFA, complementing a previous 1D analysis of the same sample. Methods: The analysis presented here includes Hα intensity and equivalent width (EW) maps and radial profiles, diagnostic emission-line ratios, and ionized-gas and stellar kinematics. It is supplemented by τ-ratio maps, which are a more efficient means to quantify the role of photoionization by the post-AGB stellar component than alternative mechanisms (e.g., AGN, low-level star formation). Results: Confirming and strengthening our previous conclusions, we find that ETGs span a broad continuous sequence in the properties of their wim, exemplified by two characteristic classes. The first (type i) comprises systems with a nearly constant EW(Hα) in their extranuclear component, which quantitatively agrees with (but is no proof of) the hypothesis that photoionization by the post-AGB stellar component is the main driver of extended wim emission. The second class (type ii) stands for virtually wim-evacuated ETGs with a very low (≤0.5 Å), outwardly increasing EW(Hα). These two classes appear indistinguishable from one another by their LINER-specific emission-line ratios in their extranuclear component. Here we extend the tentative classification we proposed previously by the type i+, which is assigned to a subset of type i ETGs exhibiting ongoing low-level star-forming activity in their periphery. This finding along with faint

  8. New generation polyphase resonant converter-modulators for the Korean atomic energy research institute

    This paper will present operational data and performance parameters of the newest generation polyphase resonant high voltage converter modulator (HVCM) as developed and delivered to the KAERI 100 MeV ''PEFP'' accelerator (1). The KAERI design realizes improvements from the SNS and SLAC designs (2). To improve the IGBT switching performance at 20 kHz for the KAERI system, the HVCM utilizes the typical zero-voltage-switching (ZVS) at turn on and as well as artificial zero-current-switching (ZCS) at turn-off. The new technique of artificial ZCS technique should result in a 6 fold reduction of IGBT switching losses (3). This improves the HCVM conversion efficiency to better than 95% at full average power, which is 500 kW for the KAERI two klystron 105 kV, 50 A application. The artificial ZCS is accomplished by placing a resonant RLC circuit across the input busswork to the resonant boost transformer. This secondary resonant circuit provides a damped ''kick-back'' to assist in IGBT commutation. As the transformer input busswork is extremely low inductance (< 10 nH), the single RLC network acts like it is across each of the four IGBT collector-emitter terminals of the H-bridge switching network. We will review these topological improvements and the overall system as delivered to the KAERI accelerator and provide details of the operational results.

  9. Quality assurance for the research and development of nuclear technology

    KAERI is carrying out several large nuclear R and D projects to achieve the indigenization of nuclear technology in Korea. In order to accomplish nuclear projects effectively, the KAERI-wide quality assurance system as well as project quality systems has been prepared for the coordination and effective implementation of various quality activities. The revision of KAERI QA Program Plan will help to establish and upgrade the effective and efficient KAERI-wide QA system. Technical support activities to the project QA program were performed in more systematic way. KAERI QA Committee was organized, and the meeting was held periodically to discuss and find out the optimum solution for the critical quality problems. Quality evaluation including internal audits was carried out to analyze the QA activities in the various projects and evlauation results was condensed to quality trend analysis. QA record preserving facility was built and was being used to maintain the QA records. The basic studies on the computer S/W QA, QA in R and D, quality costs analysis were also performed to upgrade the safety and reliability. (Author)

  10. Technical cooperation for the wider uses of Ho-166 therapeutic agents in European countries

    Czech has put their priority in developing the radiopharmaceuticals based on reactor produced Ho-166 and a related fabrication will be extended to other EU conturies including Germany, France, etc after a development of project. The collaboration will be based on the mutual agreement for developing the between research institutes, industries and academic institutes and further researches should be followed by the issue of developing radiopharmaceuticals using Ho-166. To strengthen the collaboration, detailed discussions for the practical collaboration have been made through the visitation to the research institution of each counter part. For implementing the collaboration between NPI and KAERI, an institutional basis technical cooperation agreement(TCA) will be concluded. Furthermore, agreement for the substantial collaboration on Ho-166 related researches will be made after the conclusion of the TCA. It will accelerate the commercialization of KAERI developed Ho-166 therapeutic agents into other European countries once authorization is acquired in Czech because the regulatory authorization for the approval as a commercialization of radiopharmaceuticals in Czech is very similar to other EU countries. Thereby, its commercialization of Ho-166 products in Czech will lead to potential commercialization on other EU countries and this would be returning to KAERI in the from of investment profits. It is certain that the technical exportation of Ho-166 based therapeutic agents that have been developed KAERI into European countries such as France and Germany, etc. will be ensured through the successful technical collaboration program between KAERI and NPI

  11. ADS National Programmes: Republic of Korea

    ADS research of Korea was carried out at KAERI (Korea Atomic Energy Research Institute) for the transmutation of long lived nuclides in spent fuels. The KAERI ADS system was called HYPER (HYbrid Power Extraction Reactor). The HYPER research was carried out as a 10 year nuclear research programme funded by the government. The ADS research of KAERI consisted of 3 phases. A basic concept of HYPER was established in the first phase (1997– 2000) of the development. The key technologies related to HYPER was studied in the second phase (2001–2003) while upgrading the design. The conceptual design of HYPER core was completed in the third phase (2004–2006). The research of key technologies was continued in the third phase. The research results can be summarized in three categories: Design and analysis; Fuel experiment; Pb-Bi experiment

  12. Development of Voloxidation Process for Treatment of LWR Spent Fuel

    Park, J. J.; Jung, I. H.; Shin, J. M. (and others)

    2007-08-15

    The objective of the project is to develop a process which provides a means to recover fuel from the cladding, and to simplify downstream processes by recovering volatile fission products. This work focuses on the process development in three areas ; the measurement and assessment of the release behavior for the volatile and semi-volatile fission products from the voloxidation process, the assessment of techniques to trap and recover gaseous fission products, and the development of process cycles to optimize fuel cladding separation and fuel particle size. High temperature adsorption method of KAERI was adopted in the co-design of OTS for hot experiment in INL. KAERI supplied 6 sets of filter for hot experiment. Three hot experiment in INL hot cell from the 25th of November for two weeks with attaching 4 KAERI staffs had been carried out. The results were promising. For example, trapping efficiency of Cs was 95% and that of I was 99%, etc.

  13. Current status of 700 MWe class PHWR NSSS design and engineering technology

    The capability of NSSS design and engineering technology of KAERI for 700 MWe class PHWR (CANDU 6) as of 1996 March 30 is comprehensively summarized in this report. The design and engineering capability of KAERI which have been gained during the implementation of Wolsung 2, 3 and 4 project are assessed, and showed with tangible scale. The status of Technology Transfer Materials received from Atomic Energy of Canada Limited under the Technology Transfer Agreement (TTA) which is effective simultaneously to Wolsung 3 and 4 contract, is also given in this report. The division of responsibility (DOR) of KAERI for Wolsung 2 and Wolsung 3 and 4 contract is also given, and expansion of DOR from Wolsung 2 contract to Wolsung 3 and 4 is presented. 3 refs. (Author)

  14. The operational procedure of an electron beam accelerator

    Lee, Byung Cheol; Choi, Hwa Lim; Yang, Ki Ho; Han, Young Hwan; Kim, Sung Chan

    2008-12-15

    The KAERI(Korea Atomic Energy of Research Institute) high-power electron beam irradiation facility, operating at the energies between 0.3 MeV and 10 MeV, has provided irradiation services to users in industries, universities, and institute in various fields. This manual is for the operation of an electron beam which is established in KAERI, and describes elementary operation procedures of electron beam between 0.3 Mev and 10 MeV. KAERI Electron Accelerator facility(Daejeon, Korea) consists of two irradiators: one is a low-energy electron beam irradiator operated by normal conducting RF accelerator, the other is medium-energy irradiator operated by superconducting RF accelerator. We explain the check points of prior to operation, operation procedure of this facility and the essential parts of electron beam accelerator.

  15. Development of Nuclear Control and Management Information Treatment System

    To implement obligations under the Non-Proliferation Treaty (NPT) and the bilateral agreements more effectively, we proposed a computerized system named the Nuclear Control and Management Information Treatment System (NCAMITS) as a part of the Nuclear Transparency Enhancement Project at the Korea Atomic Energy Research Institute (KAERI). The database system is designed not only to undertake the facility-level accounting for and control of nuclear material at KAERI, but also to meet the requirements of the State (National) System of Accounting and Control (SSAC). Since the NCAMITS will provide services for the facility operators as well as the safeguard information managers at KAERI, the development of the system is supposed to accommodate the end-user's convenience and the manager's sophisticated specifications as well

  16. Study on promotion of venture business

    This study reviewed the concepts of venture business and surveyed venture business support system nationwide. The venture business support system is summarized in depth to help the pre-entrepreneurs under establishing venture business. This study also reviewed the technology management system of KAERI and surveyed its historical accomplishment of technology transfer. Then, this study suggested its future direction by surveying the system of advanced countries and also suggested the measures to meet the future direction. The main finding of this study is that the direct investment to venture business by KAERI could greatly contribute to promoting venture business. Therefore, the government and KAERI should make efforts to change the technology management system toward the direct investment. Finally, this study concluded by offering policy suggestions to the government on improvement of technology management system

  17. Numerical investigation of the radiation characteristics of a variable-period helical undulator

    A helical undulator with a variable-period capability has been developed at the Korea Atomic Energy Research Institute (KAERI) to generate high power radiation in the terahertz range. A simulation code for the spontaneous emission from an electron beam inside an undulator has been developed to characterize the performance of the undulator. In the case of the KAERI undulator, there is a non-negligible high-order harmonics in the longitudinal field distribution compared with a bifilar one.The axial velocity modulation by the high-order harmonics in the field distribution has been found to lead to small deviation of the spectrum of spontaneous emission from the KAERI undulator with respect to the bifilars one. The gain functions obtained from the spontaneous emission spectra according to the Madey theory, show similar shapes for both undulators

  18. International Collaboration for Enhancement of Nuclear Transparency in Laser Technologies

    KAERI had executed the laser isotope separation for nuclear material in the past, but now wants to have international collaboration in laser technologies in order to inform our peaceful purpose and to enhance nuclear transparency in the field. Also KAERI wants to keep developing our original technologies and some advanced laser technologies to be utilized in practice by international collaboration in a short period. The materials science department in INL and the quantum optics division in KAERI has cooperated in both laser ultrasonics and nondestructive examination fields. Ultrashort laser ultrasonics and resonance spectroscopy for laser ultrasonics were collaborated. And we also develop our own experimental facility from the experience obtained from collaboration with INL

  19. Localization of NSLS Design Technology

    Korea is entering into the third generation nuclear power program with the commencement of KNU 11 and 12 project. The NSLS system design technology is to be established domestically as joint effort between the local recipient, KAERI, and the foreign supplier, Combustion Engineering through active participation in the Joint System Design of KNU 11 and 12 as well as the Technology Transfer Agreement and Standardization Project for the future units. The design center is expected to move gradually from C-E Windsor site to KAERI Diked site in order to maximize the localization potential. All necessary infrastructure such as manpower training, project management system, computer codes and technical documents is expected to be fully operational at KAERI site through the Technology Transfer Agreement

  20. Technical review of the LB-LOCA analysis for China Qinshan-2 nuclear power plant

    This report is prepared as a first product of LB-LOCA analysis review program for 600 MWe Qinshan nuclear power plant, which was agreed between NPIC and KAERI. Characteristics of Qinshan plant were identified by comparing relevant parameters with those of Kori 2 and Kori 3/4 which are Westinghouse 2- and 3-loop plants in Korea. Technical review on the LB-LOCA analysis which was performed by NPIC for Qinshan plant using RELAP5/MOD2 code is carried out to examine and ascertain the soundness of evaluation methodology and the properness of calculated LOCA phenomena. This review is expected to provide a guidance to future LOCA analysis which will be jointly accomplished by NPIC and KAERI at KAERI site. (Author) 1 fig., 3 tabs

  1. Inter-comparison study of the ENSEMBLE project

    During the days of the Chernobyl accident, the European national long-range dispersion forecasts would differ because of differences in national models, and differences in weather prediction methods. ENSEMBLE project was launched for a reconciliation and harmonization of the disparate long-range dispersion forecasts. Responsible European emergency organizations in addition to Canadian, Japanese, Korean and US agencies have participated in ENSEMBLE. KAERI joined the inter-comparison study for the exercise on the 901-001 scenario in ENSEMBLE. KAERI was assigned KR1 as a national code and 53 as a model number. The model of KAERI was compared with the other models of the participants in ENSEMBLE. The comparative results are presented with the scatter plots and statistical methods in this paper. (author)

  2. Development of Nuclear Control and Management Information Treatment System

    Yoo, J. G.; Lee, B. D.; So, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    To implement obligations under the Non-Proliferation Treaty (NPT) and the bilateral agreements more effectively, we proposed a computerized system named the Nuclear Control and Management Information Treatment System (NCAMITS) as a part of the Nuclear Transparency Enhancement Project at the Korea Atomic Energy Research Institute (KAERI). The database system is designed not only to undertake the facility-level accounting for and control of nuclear material at KAERI, but also to meet the requirements of the State (National) System of Accounting and Control (SSAC). Since the NCAMITS will provide services for the facility operators as well as the safeguard information managers at KAERI, the development of the system is supposed to accommodate the end-user's convenience and the manager's sophisticated specifications as well.

  3. A Procedure for Determination of Degradation Acceptance Criteria for Structures and Passive Components in Nuclear Power Plants

    Nie, J.; Braverman, J.; Hofmayer, C.; Choun, Y-S.; Hahm, D.; Choi, I-K.

    2012-01-30

    The Korea Atomic Energy Research Institute (KAERI) has been collaborating with Brookhaven National Laboratory since 2007 to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). This collaboration program aims at providing technical support to a five-year KAERI research project, which includes three specific areas that are essential to seismic probabilistic risk assessment: (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. The understanding and assessment of age-related degradations of structures, systems, and components and their impact on plant safety is the major goal of this KAERI-BNL collaboration. Four annual reports have been published before this report as a result of the collaboration research.

  4. The study for the high qualification of international nuclear training

    Noh, Byong Chull; Kim, Hyun Jin

    2012-12-15

    It is suggested how to reach high qualification of KAERI international nuclear training and how to play a leading role for new paradigm on the international training on the world. 1. The formulation of the core nuclear training framework- The systematic formulation of nuclear training framework based on the existing turning course design 2. Planning and operation of KAERI- Excellent Technology Series training course- The advertisement for KAERI Excellent Technology through the continuous international training and the future market development on the world for the nuclear technology 3. e-Learning training contents development- e-Learning training contents development to play a leading role for new training paradigm on the world and to overcome the limit of time/spacy.

  5. First detection of [N II] 205 μm absorption in interstellar gas. Herschel-HIFI observations towards W 31C, W 49N, W 51, and G34.3+0.1

    Persson, C. M.; Gerin, M.; Mookerjea, B.; Black, J. H.; Olberg, M.; Goicoechea, J. R.; Hassel, G. E.; Falgarone, E.; Levrier, F.; Menten, K. M.; Pety, J.

    2014-08-01

    We present high resolution [N ii] 205 μm (3P1 - 3P0) spectra obtained with Herschel-HIFI towards a small sample of far-infrared bright star forming regions in the Galactic plane: W 31C (G10.6-0.4), W 49N (G43.2-0.1), W 51 (G49.5-0.4), and G34.3+0.1. All sources display an emission line profile associated directly with the H ii regions themselves. For the first time we also detect absorption of the [N ii] 205 μm line by extended low-density foreground material towards W 31C and W 49N over a wide range of velocities. We attribute this absorption to the warm ionised medium (WIM) and find N(N+) ≈ 1.5 × 1017 cm-2 towards both sources. This is in agreement with recent Herschel-HIFI observations of [C ii] 158 μm, also observed in absorption in the same sight-lines, if ≈7-10% of all C+ ions exist in the WIM on average. Using an abundance ratio of [N] / [H] = 6.76 × 10-5 in the gas phase we find that the mean electron and proton volume densities are ~0.1-0.3 cm-3 assuming a WIM volume filling fraction of 0.1-0.4 with a corresponding line-of-sight filling fraction of 0.46-0.74. A low density and a high WIM filling fraction are also supported by RADEX modelling of the [N ii] 205 μm absorption and emission together with visible emission lines attributed mainly to the WIM. The detection of the 205 μm line in absorption emphasises the importance of a high spectral resolution, and also offers a new tool for investigation of the WIM. Appendix A is available in electronic form at http://www.aanda.orgHerschel is an ESA space observatory with science instruments provided by European-led Principal Investigator consortia and with important participation from NASA.

  6. Assessment of neutron transport codes for application to CANDU fuel lattices analysis

    Roh, Gyu Hong; Choi, Hang Bok

    1999-08-01

    In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS

  7. An International Peer Review of the Programme for the Deep Geological Disposal of High Level Radioactive Waste from Pyro-Processing in the Republic of Korea. Report of an IAEA International Review Team

    The development of a radioactive waste disposal system is indispensable in maintaining the sustainability of nuclear energy. The Korea Atomic Energy Research Institute (KAERI) has studied the direct geological disposal of spent nuclear fuel since 1997. KAERI has also focused on the development of processes suitable for reducing the volume of spent nuclear fuel and the recycling of valuable fissile material. One of the most promising technologies investigated by KAERI is the pyro-processing of spent nuclear fuel followed by the geological disposal of the generated high level waste (HLW). Since 2007, KAERI has been running a research programme focusing on the recycling of spent nuclear fuel, as well as studies aimed at the development of a relevant geological disposal system able to accept the resulting HLW. The core aims of the KAERI study were to characterize the geological media, design a repository system and assess the overall safety of the disposal system. The development of pyro-processing technology is ongoing and has not yet been demonstrated at the commercial level. Thus, the government of the Republic of Korea requested an assessment of the technical feasibility of this technology. The assessment also included the appraisal of a disposal solution for waste generated by pyro-processing. With regard to the latter, KAERI requested that the IAEA review the status of the disposal project within the Waste Management Assessment and Technical Review Programme (WATRP). Peer reviews are increasingly being acknowledged as an important element in building broader stakeholder confidence in the safety and viability of related facilities. This report presents the consensus view of the international group of experts convened by the IAEA to perform the review

  8. The Effect of External Vessel Cooling for a 2 inch LOCA Severe Accident Scenario at SMART with MIDAS/SMR

    KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. This feature may prevent the large size of LOCA. However it is necessary to estimate the hypothetical severe accidents progression for improving the degree of safety and identifying the unknown weakness of the system against an accident. To simulate a hypothetical severe accident for the SMART, we adopt the MIDAS/SMR code which was developed by KAERI

  9. The Effect of External Vessel Cooling for a 2 inch LOCA Severe Accident Scenario at SMART with MIDAS/SMR

    Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seong Won [KORTIC, Daejeon (Korea, Republic of)

    2010-05-15

    KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. This feature may prevent the large size of LOCA. However it is necessary to estimate the hypothetical severe accidents progression for improving the degree of safety and identifying the unknown weakness of the system against an accident. To simulate a hypothetical severe accident for the SMART, we adopt the MIDAS/SMR code which was developed by KAERI

  10. Karma1.1 benchmark calculations for the numerical benchmark problems and the critical experiments

    The transport lattice code KARMA 1.1 has been developed at KAERI for the reactor physics analysis of the pressurized water reactor. This program includes the multi-group library processed from ENDF/B-VI R8 and also utilizes the macroscopic cross sections for the benchmark problems. Benchmark calculations were performed for the C5G7 and the KAERI benchmark problems given with seven group cross sections, for various fuels loaded in the operating pressurized water reactors in South Korea, and for the critical experiments including CE, B and W and KRITZ. Benchmark results show that KARMA 1.1 is working reasonably. (author)

  11. Korea's CANDU fuel R and D program

    As the first R and D activity led to the nuclear fuel industrialization in Korea, KAERI had successfully developed the CANDU-6 fuel bundle in the period of 1981 to 1986 and has commercially produced more than 35,000 fuel bundles for the use in Wolsong Unit 1 since 1987. The commercial production of the CANDU-6 fuel in KAERI will be terminated on the end of 1997 and KNFC will take over the mission of CANDU-6 fuel production with a capacity of 400 tons of uranium per year form 1998. (author)

  12. Development of Wall-to-Fluid Heat Transfer Package for the SPACE Code

    A SPACE code which has a multi-dimensional analysis capability by incorporating a dispersed liquid phase into the thermo-hydraulic field equations is under development for a safety analysis of PWRs. Several research and industrial organizations are participating in the collaboration for the development program, including KAERI, KOPEC, KNF, and KEPRI. The main task of KAERI is to develop physical models and correlation packages for constitutive equations; a wall heat transfer package, a wall and interfacial friction package, an interfacial heat and mass transfer package and a flow regime selection package. This paper describes the development program for a wall-to-fluid heat transfer package for the SPACE code

  13. The korea multi-purpose research reactor

    This paper presents and discusses background and status of the design of the 30MW Korea Multi-purpose Research Reactor(KMRR) which is planed to achieve its first criticality in December, 19992, at Daeduk site of the Korea Advanced Energy Research Institute (KAERI). KAERI playing the leading role in Korea's nuclear technology development takes the total responsibility for its design, construction and operation. Number of Korean nuclear industries are, also, actively participating in the project while making the most of their expertise in relevant areas. (Author)

  14. A Study to Promote a Collaboration of R and D for Nuclear Energy Technology Development between Korea and Kazakhstan

    The ultimate goal of this investigation to promote a collaboration of R and D for Nuclear Energy Technology Development between Korea and Kazakhstan. To understand the research power of the Kazakhstan, we visited the INP(Institute of Nuclear Physics) which is one of the branch of Nation Cuclear Center-Repunlic Kazakhstan. We presented the present status of the nuclear energy related research in KAERI. The director of international cooperation in the ministry of Mineral resources, the director of INP and vice director of IAE had visited KAERI, KIRAM and discussed about potential cooperation in nuclear research related field

  15. Application of TSH bioindicator for studying the biological efficiency of radiation

    Dose response relationships for various endpoints (gene and lethal mutations, cell cycle alterations) in somatic cells of Tradescantia clone 4430 were established for X-rays and for mixed fast and thermal neutrons from Cf-252 source of KAERI and from U-120 cyclotron of INP. This was a pilot experiment to check if it is possible to establish the relative biological effectiveness values for Cf-252 irradiated TSH cells, with and without boron ion pretreatment, in conditions of mutual KAERI-INP experiment. When T-4430 was pretreated with boron ion, there was and enhancement in biological efficacy of neutron form Cf-252 source. 2 tabs., 7 figs., 7 refs. (Author)

  16. Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code

    In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)

  17. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  18. Development of an accident consequence analysis program based on the object oriented programming technique

    The KAERI accident consequence analysis program KAPAC is being developed on the basis of reusable objects in PPAM (platform for the development of plant analysis and management codes). Development of PPAM is being conducted at the Korea Atomic Energy Research Institute (KAERI) in order to be able to provide portability and reusability of computer codes, and consistent user interface in developing software with the use of object oriented programming (OOP) under a Microsoft Windows environment. By constructing the platform, software development can benefit from a shorter development cycle and an easier validation and verification process. 1 ref., 2 figs

  19. Geo-synthetic study on the geological and hydro-geological model around KURT area

    Kurt (KAERI Underground Research Tunnel) is a small-scale research tunnel which was located at Korea Atomic Energy Research Institute (KAERI). This research tunnel was constructed for the demonstration study including engineering and natural barrier system for a radioactive waste disposal. Before beginning of demonstration study, the site characterization works should be preceded all of the research activities. It is caused that the site specific conditions may affect the environments of demonstration studies such as the heat transfer test, solute migration test and groundwater flow test

  20. In-place leak testing of multiple HEPA filter

    Lee, H. K.; Hong, K. P.; Jun, Y. B.; Min, D. K.; Park, K. J.; Yang, S. Y.; Lee, E. P.; Hwang, Y. H.; Su, H. S.; Kim, K. S.; Kwon, H. M. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    In-place leak test for the filter banks with multiple high efficiency particulate air(HEPA) filters was carried in KAERI. As a result of the test, in the air flow capacity of 31,500 CMH and 22,800 CMH, penetration rate appeared to be 0.015 {approx} 0.036 %. These values satisfy the Regulatory Guide 1.14 reference, 0.05%. Although test has a complicated procedure, the testing technique was established by KAERI, and will be contributed to the safety inspection of ventilation system in nuclear facilities in the future.

  1. A Study to Promote a Collaboration of R and D for Nuclear Energy Technology Development between Korea and Kazakhstan

    Ryh, Sipyo; Kim, Cheoljung; Yoo, Bungduk; Lee, Yongjoo; Kim, Hansoo; Yoon, Sungwon; Jeong, Hwansam; Jeong, Gijung

    2005-01-15

    The ultimate goal of this investigation to promote a collaboration of R and D for Nuclear Energy Technology Development between Korea and Kazakhstan. To understand the research power of the Kazakhstan, we visited the INP(Institute of Nuclear Physics) which is one of the branch of Nation Cuclear Center-Repunlic Kazakhstan. We presented the present status of the nuclear energy related research in KAERI. The director of international cooperation in the ministry of Mineral resources, the director of INP and vice director of IAE had visited KAERI, KIRAM and discussed about potential cooperation in nuclear research related field.

  2. Internal structure of spiral arms traced with [C II]: Unraveling the warm ionized medium, H I, and molecular emission lanes

    Velusamy, T.; Langer, W. D.; Goldsmith, P. F.; Pineda, J. L.

    2015-06-01

    Context. The spiral arm tangencies are ideal lines of sight in which to determine the distribution of interstellar gas components in the spiral arms and study the influence of spiral density waves on the interarm gas in the Milky Way. [C II] emission in the tangencies delineates the warm ionized component and the photon-dominated regions and is thus an important probe of spiral arm structure and dynamics. Aims: We aim to use [C II], H I, and 12CO spectral line maps of the Crux, Norma, and Perseus tangencies to analyze the internal structure of the spiral arms in different gas layers. Methods: We used [C II] l-V maps along with those for H I and 12CO to derive the average spectral line intensity profiles over the longitudinal range of each tangency. Using the VLSR of the emission features, we located the [C II], H I, and 12CO emissions along a cross cut of the spiral arm. We used the [C II] velocity profile to identify the compressed warm ionized medium (WIM) in the spiral arm. Results: We present a large-scale (~15°) position-velocity map of the Galactic plane in [C II] from l = 326.6° to 341.4° observed with Herschel HIFI. In the spectral line profiles at the tangencies, [C II] has two emission peaks, one associated with the compressed WIM and the other the molecular gas photon-dominated regions. When represented as a cut across the inner to outer edge of the spiral arm, the [C II]-WIM peak appears closest to the inner edge while 12CO and [C II] associated with molecular gas are at the outermost edge. H I has broader emission with an intermediate peak located nearer to that of 12CO. Conclusions: The velocity-resolved spectral line data of the spiral arm tangencies unravel the internal structure in the arms locating the emission lanes within them. We interpret the excess [C II] near the tangent velocities as shock compression of the WIM induced by the spiral density waves and as the innermost edge of spiral arms. For the Norma and Perseus arms, we estimate

  3. Time-Average Calculation using FEM in a CANDU Reactor

    Ryu, Eun Hyun; Park, Joo Hwan; Song, Yong Man; Lee Chung Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

    2012-05-15

    To get a much accurate result and to be sure about the calculated reactor physics value, new code system which is appropriate to the CANDU reactor and has high fidelity is required. This study here is to understand and analyze the existing code system, WIMS-RFSP. Because the FEM codes used here can calculate multiplication factor, group flux, channel power easily with cross section data from WIMS and geometrical data from GMSH, the results of FEM are good examples to compare with RFSP results. With the comparison process itself and numerical experiments, it is expected that the basis of new code system become abundant. Time-average module is mainly discussed with regular process in RFSP

  4. Time-Average Calculation using FEM in a CANDU Reactor

    To get a much accurate result and to be sure about the calculated reactor physics value, new code system which is appropriate to the CANDU reactor and has high fidelity is required. This study here is to understand and analyze the existing code system, WIMS-RFSP. Because the FEM codes used here can calculate multiplication factor, group flux, channel power easily with cross section data from WIMS and geometrical data from GMSH, the results of FEM are good examples to compare with RFSP results. With the comparison process itself and numerical experiments, it is expected that the basis of new code system become abundant. Time-average module is mainly discussed with regular process in RFSP

  5. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  6. Modification to ORIGEN2 for generating N Reactor source terms. Volume 1

    This report discusses work that has been done to upgrade the ORIGEN2 code cross sections to be compatible with the WIMS computer code data. Because of the changes in the ORIGEN2 calculations. Details on changes made to the ORIGEN2 computer code and the Radnuc code will be discussed along with additional work that should be done in the future to upgrade both ORIGEN2 and Radnuc. A detailed historical description of how source terms have been generated for N Reactor fuel stored in the K Basins has been generated. The neutron source discussed in this description was generated by the WIMS computer code (Gubbins et al. 1982) because of known shortcomings in the ORIGEN2 (Croff 1980) cross sections. Another document includes a discussion of the ORIGEN2 cross sections

  7. The square boundary version of the WIMSE module W-PIJ

    W-PIJ is one component of the modular reactor physics code scheme known as WIMS-E. Its purpose is to read cross section data from a WIMS-E interface, a two dimensional cluster geometry from user input, and to calculate the associated region to region first flight neutron collision probabilities. These it writes back to the interface as data for a solution module which can calculate fluxes. Recent modifications to the code have provided better data checking facilities including a lineprinter geometry plotting option, and the option of an explicit square outer boundary. This report briefly describes the modifications that have been made to the code and includes a complete user guide to all options, old and new. (author)

  8. Application of WIMSD-4 for ''MARIA'' reactor lattice calculations

    A general description of the WIMSD-4 lattice code is given with the emphasis on available geometrical models. The difficulties encountered while modelling reactor lattices with the tubular type fuel elements are explained. Then the analysis of code options allowing to overcome these difficulties is carried out. Eventually, recommendations of options and input parameters for calculations of MARIA reactor lattice with satisfactory accuracy are given. During the work a set of modifications had to be introduced leading to a new code version called WIMS-S. Another version, under the name WIMS-T has been developed to allow for burnup calculations of the MARIA reactor lattice with improved resonance approach. (author). 14 refs, 6 figs, 10 tabs

  9. Evaluation of the DRAGON code for VHTR design analysis.

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  10. Fuel management at the Petten high flux reactor

    Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high 235U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing facilities can be used for many different types of application. (author)

  11. Calculational study on irradiation of americium fuel samples in the Petten High Flux Reactor

    A calculational study on the irradiation of americium samples in the Petten High Flux Reactor (HFR) has been performed. This has been done in the framework of the international EFTTRA cooperation. For several reasons the americium in the samples is supposed to be diluted with a neutron inert matrix, but the main reason is to limit the power density in the sample. The low americium nuclide density in the sample (10 weight % americium oxide) leads to a low radial dependence of the burnup. Three different calculational methods have been used to calculate the burnup in the americium sample: Two-dimensional calculations with WIMS-6, one-dimensional calculations with WIMS-6, and one-dimensional calculations with SCALE. The results of the different methods agree fairly well. It is concluded that the radiotoxicity of the americium sample can be reduced upon irradiation in our scenario. This is especially the case for the radiotoxicity between 100 and 1000 years after storage. (orig.)

  12. Modification to ORIGEN2 for generating N Reactor source terms. Volume 1

    Schwarz, R.A.

    1997-04-01

    This report discusses work that has been done to upgrade the ORIGEN2 code cross sections to be compatible with the WIMS computer code data. Because of the changes in the ORIGEN2 calculations. Details on changes made to the ORIGEN2 computer code and the Radnuc code will be discussed along with additional work that should be done in the future to upgrade both ORIGEN2 and Radnuc. A detailed historical description of how source terms have been generated for N Reactor fuel stored in the K Basins has been generated. The neutron source discussed in this description was generated by the WIMS computer code (Gubbins et al. 1982) because of known shortcomings in the ORIGEN2 (Croff 1980) cross sections. Another document includes a discussion of the ORIGEN2 cross sections.

  13. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease. (author)

  14. Depleted Reactor Analysis With MCNP-4B

    Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

  15. The use of WIMSD4 and LWRWIMS, READWT and FILSIX, to generate two-group data for reactor calculations

    This report brings together information describing the transfer of data between WIMS lattice calculations and JOSHUA reactor calculations. It describes the interface in considerable detail and then provides the information that a user of WIMSD4 or LWRWIMS would need to generate an interface. Two auxiliary codes, READWT and FILSIX, are used to process the interface and maintain a library of reactor code data. Input and use of these codes are specified. (author)

  16. Persistence and adherence in multiple sclerosis patients starting glatiramer acetate treatment: assessment of relationship with care received from multiple disciplines

    Jongen PJ; Lemmens WA; Hupperts R; Hoogervorst ELJ; Schrijver HM; Slettenaar A; de Schryver EL; Boringa J; van Noort E; Donders R

    2016-01-01

    Peter Joseph Jongen,1,2 Wim A Lemmens,3 Raymond Hupperts,4 Erwin LJ Hoogervorst,5 Hans M Schrijver,6 Astrid Slettenaar,7 Els L de Schryver,8 Jan Boringa,9 Esther van Noort,10 Rogier Donders3 1Department of Community and Occupational Medicine, University Medical Centre Groningen, University Groningen, Groningen, 2MS4 Research Institute, 3Department for Health Evidence, Radboud university medical center, Nijmegen, 4Department of Neurology, Zuijderland Medisch Centrum Sittard, Sittard, 5Departm...

  17. Europeaisches Kern Forschungszentrum

    US Information Service in Germany

    1959-01-01

    Propaganda film (heroic) which emphasises international collaboration. No practical application of the work which goes on here. Interesting facts which come up: CERN employed 1000 people, average age of 33 years. Ford Foundation paid many people. Just after PS comes into operation, no talk of experiments. Very little explanation of what goes on at CERN. Images of: CERN cafetaria, CERN main building, CERN personalities (Lew Kowarski, Wim Klein, Roger Anthoine ....)

  18. Study for 228Th reduction in thermal reactor with Th-U fuel cycls

    1999-01-01

    By using computercode WIMS/CENDL, the effects of some parameters, core configuration such as fuel element structure, neutron flux and burn-up, are discussed in thispaper.It is shown that high neutron flux, small fuel rod diameter,large volume ratio of coolant to fuel, seed-blank heterogeneous corearrangement and 231Pa chemical separation are necessary for reducing 228Th production in reactor.

  19. Conversion and evaluation of the THOR reactor core to TRIGA fuel elements

    The THOR reactor is a pool type 1 MW research reactor and has been operated since 1961. The original MTR fuel elements have been gradually replaced by TRIGA fuel elements since 1977 and the conversion completed in 1987. The calculations were performed for various core configurations by using computer codes, WIMS/CITATION. The computing results have been evaluated and compared with the core measurements after the fuel conversion. The analysis results are in good correspondence with the measurements. (author)

  20. Use of the 'DRAGON' program for the calculation of reactivity devices

    DRAGON is a computer program developed at the Ecole Polytechnique of the University of Montreal and adopted by AECL for the transport calculations associated to reactivity devices. This report presents aspects of the implementation in NASA of the DRAGON program. Some cases of interest were evaluated. Comparisons with results of known programs as WIMS D5, and with experiments were done. a) Embalse (CANDU 6) cell without burnup and leakage. Calculations of macroscopic cross sections with WIMS and DRAGON show very good agreement with smaller differences in the thermal constants. b) Embalse fresh cell with different leakage options. c) Embalse cell with leakage and burnup. A comparison of k-infinity and k-effective with WIMS and DRAGON as a function of burnup shows that the differences ((D-W)/D) for fresh fuel are -0.17% roughly constant up to about 2500 MWd/tU, and then decrease to -0.06 % for 8500 MWd/tU. Experiments made in 1977 in ZED-2 critical facility, reported in [3], were used as a benchmark for the cell and supercell DRAGON calculations. Calculated fluxes were compared with experimental values and the agreement is so good. d) ZED-2 cell calculation. The measured buckling was used as geometric buckling. This case can be considered an experimental verification. The calculated reactivity with DRAGON is about 2 mk, and can be considered satisfactory. WIMS k-effective value is about one mk higher. e) Supercell calculations for ZED-2 vertical and horizontal tube and rod adjuster using 2D and 3D models were done. Comparisons between measured and calculated fluxes in the vicinity of the adjuster rods. Incremental cross sections for these adjusters were calculated using different options. f) ZED-2 reactor calculations with PUMA reveal a good concordance with critical heights measured in experiments. The report describes also particular features of the code and recommendations regarding its use that may be useful for new users. (author)

  1. Accelerating Into the Future: From 0 to GeV in a Few Centimeters (LBNL Summer Lecture Series)

    Summer Lecture Series 2008: By exciting electric fields in plasma-based waveguides, lasers accelerate electrons in a fraction of the distance conventional accelerators require. The Accelerator and Fusion Research Division's LOASIS program, headed by Wim Leemans, has used 40-trillion-watt laser pulses to deliver billion-electron-volt (1 GeV) electron beams within centimeters. Leemans looks ahead to BELLA, 10-GeV accelerating modules that could power a future linear collider.

  2. Russki "Tjulpan" pokoril lazurnõi bereg / Maria Davtjan, Tatjana Pinskaja

    Davtjan, Maria

    2008-01-01

    61. Cannes'i filmifestival on lõppenud. Vaatenurga preemia sai Kasahstani režissöör Sergei Dvortsevoi filmi "Tulp" ("Tjulpan") eest. Lisatud ka lühiintervjuu režissööriga. Ka mõnest filmist, nii auhinnatuist (Steven Soderbergh'i "Che", Clint Eastwoodi "Changeling") kui ilmajäänuist (Atom Egoyani "Adoration", Wim Wendersi "Palermo Shooting"), Sharon Stone'i juhitud heategevast AIDSi ravi toetavast oksjonist

  3. PHYSICAL CONDITIONS IN BARNARD'S LOOP, COMPONENTS OF THE ORION-ERIDANUS BUBBLE, AND IMPLICATIONS FOR THE WARM IONIZED MEDIUM COMPONENT OF THE INTERSTELLAR MEDIUM

    We have supplemented existing spectra of Barnard's Loop with high accuracy spectrophotometry of one new position. Cloudy photoionization models were calculated for a variety of ionization parameters and stellar temperatures and compared with the observations. After testing the procedure with recent observations of M43, we establish that Barnard's Loop is photoionized by four candidate ionizing stars, but agreement between the models and observations is only possible if Barnard's Loop is enhanced in heavy elements by about a factor of 1.4. Barnard's Loop is very similar in properties to the brightest components of the Orion-Eridanus Bubble and the warm ionized medium (WIM). We are able to establish models that bound the range populated in low-ionization color-color diagrams (I([S II])/I(Hα) versus I([N II])/I(Hα)) using only a limited range of ionization parameters and stellar temperatures. Previously established variations in the relative abundance of heavy elements render uncertain the most common method of determining electron temperatures for components of the Orion-Eridanus Bubble and the WIM based only on the I([N II])/I(Hα) ratio, although we confirm that the lowest surface brightness components of the WIM are on average of higher electron temperature. The electron temperatures for a few high surface brightness WIM components determined by direct methods are comparable to those of classical bright H II regions. In contrast, the low surface brightness H II regions studied by the Wisconsin Hα Mapper are of lower temperatures than the classical bright H II regions.

  4. Reticular basement membrane in asthma and COPD: Similar thickness, yet different composition

    Jeroen JW Liesker; Ten Hacken, Nick H.; Mieke Zeinstra-Smith; Rutgers, Steven R; Dirkje S Postma; et al.

    2009-01-01

    Jeroen JW Liesker1, Nick H Ten Hacken1, Mieke Zeinstra-Smith2, Steven R Rutgers1, Dirkje S Postma1, Wim Timens21Department of Pulmonology; 2Department of Pathology, University Medical Center Groningen, University of Groningen, Groningen, The Netherlands Background: Reticular basement membrane (RBM) thickening has been variably associated with asthma and chronic obstructive pulmonary disease (COPD). Even if RBM thickness is similar in both diseases, its composition might still differ. Objectiv...

  5. Criticality calculation and control rods for the Westinghouse Reactor Evaluation Center facility

    This work evaluates two clean critical cores by WIMS-TRACA/CITATION codes calculation, 4 energy groups and bi dimension geometry. The first core is composed of U O2 with a clad of stainless steel and 20 absorbers Ag-In-Cd absorbers rods, the second is composed of U O2 with a clad of Zircaloy and 12 B4 C absorbers rods. (author)

  6. Confirmation of delayed menarche based on regression evaluation of age at menarche for age at MPV of height in female ball game players

    Fujii, Katsunori; Demura, Schinichi

    2005-01-01

    A general delay in menarche in female athletes has been confirmed based on comparisons of mean ages between athletes and non-athletes; however, it has not been possible to judge such delays individually. If delayed menarche could be evaluated for an individual, the athlete could be advised as to necessary precautions. In this study, the age at maximum peak velocity (MPV) of height, adopted as an index of physical maturation, was identified by the wavelet interpolation method (WIM). The relati...

  7. Site Verification of Weigh-in-Motion Traffic and TIRTL Classification Data

    Li, Shuo; Du, Yingzi (Eliza); Jiang, Yi

    2010-01-01

    Quality weigh-in-motion (WIM) traffic data is essential not only in general transportation application, but also in pavement design. The new AASHTO Mechanistic-Empirical Pavement Design Guide (MEPDG) for New and Rehabilitated Pavement Structures requires information on the detailed truck traffic, such as truck traffic volume, truck traffic monthly and hourly variations, vehicle class distribution, axle load, and axle load distributions, instead of the traditional ESALs. In addition, the India...

  8. Uus poliitiline kriitika Ameerika filmis / Mathura

    Mathura, pseud., 1973-

    2006-01-01

    Uued USA mängufilmid, mis on kriitilised oma riigi agressiivse poliitika suhtes : "Süriaana" ("Syriana") : režissöör Stephen Gaghan, "München" ("Munich") : režissöör Steven Spielberg, "Kahuriliha" ("Jarhead") : režissöör Sam Mendes, "Küllusemaa" ("Land of Plenty") : režissöör Wim Wenders

  9. TRIGLAV - a computer programme for research reactor calculation

    Persic, A.; Ravnik, M.; Slavic, S.; Zagar, T. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    TRIGLAV is a new computer programme for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport programme WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. (orig.)

  10. Integrated transportation monitoring system for both pavement and traffic

    Xue, Wenjing

    2013-01-01

    In the passing decades, the monitoring of pavements and passing vehicles was developed vigorously with the growth of information and sensing technology. Pavement monitoring is an essential part of pavement research and plays an important role in transportation system. At the same time, the monitoring system about the traffic, such as Weigh-in-Motion (WIM) system and traffic classification system, also attracted lots of attention because of their importance in traffic statistics and management...

  11. Static and dynamic weighing of vehicles

    Meluš, Ladislav

    2014-01-01

    The work discusses the basic principles used in systems of static and dynamic weighing vehicles and is formally divided into a theoretical part and a practical part. The theoretical part describes the basic principles of strain weighing, types and functions of truck scales, related standards and evaluate the properties currently available types of scales. In the practical part contains a conceptual design of automated weighing station based WIM (Weighing in motion).

  12. Exotic-7. Nuclear analysis

    This report contains the results of the WIMS-E 5A calculations performed for the EXOTIC-7 212-25/28 experiment, namely: - The nuclear constants for the experiment; - the reactivity effect of the experiment; - the activation of the materials during an irradiation time of 11 cycles in the reactor position H6 and the decay of the active nuclides during a 'cooling period' of one year after the end of the irradiation. (orig.)

  13. Interprofessional team management in pediatric critical care: some challenges and possible solutions

    Stocker M.; Pilgrim SB; Burmester M.; Allen ML; Gijselaers WH

    2016-01-01

    Martin Stocker,1 Sina B Pilgrim,2 Margarita Burmester,3 Meredith L Allen,4 Wim H Gijselaers5 1Neonatal and Pediatric Intensive Care Unit, Children's Hospital Lucerne, Lucerne, 2Pediatric Intensive Care, University Children's Hospital Berne, Berne, Switzerland; 3Pediatric Intensive Care Unit, Royal Brompton Hospital, London, UK; 4Department of Pediatrics, The Royal Children's Hospital, Victoria, Australia; 5Educational Research and Development, School of Business and Ec...

  14. Kinetic study of the Tehran research reactor core with low enriched fuel

    Pazirandeh, A.; Afshar Bakeshloo, A. [Tehran Univ. (Iran, Islamic Republic of). Physics Dept.; Bartsch, G. [Technische Univ. Berlin (Germany). Inst. fuer Energietechnik

    1997-11-01

    For evaluating the performance of the newly refuelled Tehran Research Reactor core with low enriched uranium fuel (LEU) in transient states a two group time dependent diffusion equation code (COSTANZA) was used. This paper presents results of calculations of the fast transients, revealing the steady performance of the core and fuel integrity during transient for a probable reactivity insertion of less than or equal dollar 1.5/0.5 s. The temperature dependant reactivity coefficients of the Doppler resonance broadening effect and of the moderator absorption cross section change and density dilution were calculated using cell-averaged 69 energy group WIMS-D/4 for two main libraries, old library and WIMKAL88, to 13 groups. The two group parameters for the COSTANZA code were also obtained by WIMS-D/4. (orig.) [Deutsch] Zur Bewertung der Leistungsfaehigkeit des neu beladenen Teheraner Forschungsreaktors mit niedrig angereichertem Uranbrennstoff bei Reaktivitaetstransienten wurde ein 2-Gruppen zeitabhaengiges Diffusionsprogramm COSTANZA verwendet. In der vorliegenden Arbeit werden Ergebnisse der Berechnung schneller Transienten vorgestellt, die das Verhalten des Reaktorkerns bzw. die Integritaet der Brennstaebe waehrend der Transienten fuer eine Reaktivitaetsaenderung von kleiner oder gleich Dollar 1.5/0.5 s zeigen. Die temperaturabhaengigen Reaktivitaetskoeffizienten der Doppler-Verbreitung im Brennstoff sowie der Dichteaenderung und der Neutronenabsorption im Moderator wurden mit Hilfe zellengemittelter 69 Energie-Gruppen der Datenbank WIMS-D/4 und fuer 13 Energiegruppen mit der Datenbank WIMKAL 88 ermittelt. Die Zweigruppendaten fuer das COSTANZA-Programm wurden ebenfalls mit Hilfe von WIMS-D/4 bestimmt. (orig.)

  15. Future Developments in Non-Repudiation in GSM WAP Applications

    Cristian Toma

    2009-01-01

    The paper presents issues and architectures for mobile applications and GSM infrastructure. The paper shows the redesign of the solution for avoiding denial of service from WAP applications using WIM features. The first section contains the structure of GSM network from voice and data point of view. The security in GSM network is presented in second chapter. The third chapter presents a solution for realizing mobile subscriber non-repudiation. The solution is based on the HTTP protocol over WAP.

  16. Future Developments in Non-Repudiation in GSM WAP Applications

    Cristian Toma

    2009-06-01

    Full Text Available The paper presents issues and architectures for mobile applications and GSM infrastructure. The paper shows the redesign of the solution for avoiding denial of service from WAP applications using WIM features. The first section contains the structure of GSM network from voice and data point of view. The security in GSM network is presented in second chapter. The third chapter presents a solution for realizing mobile subscriber non-repudiation. The solution is based on the HTTP protocol over WAP.

  17. Delta's zoeken steun bij elkaar

    Driel, van, L.

    2012-01-01

    Wereldwijd staan delta’s onder druk door bevolkingsgroei, industrialisatie en klimaatverandering. In de Delta Alliance zijn nu tien delta’s uit Noord- en Zuid-Amerika, Afrika, Azië en Nederland verenigd, en een tiental andere Afrikaanse en Aziatische delta’s zoeken nog aansluiting. ‘Men wil weten hoe andere delta’s met problemen omgaan’, verklaart programmamanager Wim van Driel de belangstelling. In juni 2011 is de stichting Delta Alliance International opgericht om het geheel meer slagkracht...

  18. Book Reviews

    Redactie KITLV

    2008-01-01

    Michael Williams; Deforesting the earth; From prehistory to global crisis (Greg Bankoff) Alexander Adelaar, Nikolaus P. Himmelmann (eds); The Austronesian languages of Asia and Madagascar (René van den Berg) Wim Ravesteijn, Jan Kop (eds); Bouwen in de archipel; Burgerlijke Openbare Werken in Nederlands-Indië en Indonesië 1800-2000 (Freek Colombijn) Susan Rodgers; Print, poetics, and politics; A Sumatran epic in the colonial Indies and New Order Indonesia (Bern...

  19. Formation and evolution of magma-poor margins, an example of the West Iberia margin

    Perez-Gussinye, Marta; Andres-Martinez, Miguel; Morgan, Jason P.; Ranero, Cesar R.; Reston, Tim

    2016-04-01

    The West Iberia-Newfoundland (WIM-NF) conjugate margins have been geophysically and geologically surveyed for the last 30 years and have arguably become a paradigm for magma-poor extensional margins. Here we present a coherent picture of the WIM-NF rift to drift evolution that emerges from these observations and numerical modeling, and point out important differences that may exist with other magma-poor margins world-wide. The WIM-NF is characterized by a continental crust that thins asymmetrically and a wide and symmetric continent-ocean transition (COT) interpreted to consist of exhumed and serpentinised mantle with magmatic products increasing oceanward. The architectural evolution of these margins is mainly dominated by cooling under very slow extension velocities (uplift and weakening of the hanginwall of the active fault, where a new fault forms. This continued process leads to the formation of an array of sequential faults that dip and become younger oceanward. Here we show that these processes acting in concert: 1) reproduce the margin asymmetry observed at the WIM-NF, 2) explain the fault geometry evolution from planar, to listric to detachment like by having one common Andersonian framework, 3) lead to the symmetric exhumation of mantle with little magmatism, and 4) explain the younging of the syn-rift towards the basin centre and imply that unconformities separating syn- and post-rift may be diachronous and younger towards the ocean. Finally, we show that different lower crustal rheologies lead to different patterns of extension and to an abrupt transition to oceanic crust, even at magma-poor margins.

  20. A Special Fiber Optic Sensor for Measuring Wheel Loads of Vehicles on Highways

    Garrick, Norman W.; Amlan Sen; Malla, Ramesh B.

    2008-01-01

    This paper presents results from an investigation on a special optical fiber as a load sensor for application in Weigh-in-Motion (WIM) systems to measure wheel loads of vehicles traveling at normal speed on highways. The fiber used has a unique design with two concentric light guiding regions of different effective optical path lengths, which has the potential to enable direct measurement of magnitudes as well as locations of forces acting at multiple points along a single fiber. The optical ...

  1. On the thermal scattering law data for reactor lattice calculations

    Thermal scattering law data for hydrogen bound in water, hydrogen bound in zirconium hydride and deuterium bound in heavy water have been re-evaluated. The influence of the thermal scattering law data on critical lattices has been studied with detailed Monte Carlo calculations and a summary of results is presented for a numerical benchmark and for the TRIGA reactor benchmark. Systematics for a large sequence of benchmarks analysed with the WIMS-D lattice code are also presented. (author)

  2. WIMSCORE

    A code was developed for producing group constants and other data needed for reactor core burnup calculations. The code named WIMSCORE gets its input from WIMS output files and evaluates it to be served as input to the diffusion-burnup codes TDB, TRITON and CITATION. The purpose of this code is to facilitate the automation of data transfer between codes which is otherwise a very time-consuming and bound-to-error process. (author)

  3. Criticality safety of fresh and irradiated fuel of the 6.5 MW heavy water research reactor

    Criticality safety problems, arising in away-from-reactor handling of the 6.5 MW heavy water research reactor fuel, are considered. The present status of fresh and spent fuel storage is described. To calculate criticality parameters of different non-reactor configurations of fresh and irradiated fuel, the well known WIMS code is combined with a few group three dimensional diffusion theory code TRITON. Typical results are presented and discussed. (author)

  4. Once-daily dose regimen of ribavirin is interchangeable with a twice-daily dose regimen: randomized open clinical trial

    Balk JM; Haenen GRMM; Koc ÖM; Peters R; Bast A; van der Vijgh WJF; Koek GH

    2015-01-01

    Jiska M Balk,1 Guido RMM Haenen,1 Özgür M Koc,2 Ron Peters,3 Aalt Bast,1 Wim JF van der Vijgh,1 Ger H Koek,4 1Department of Toxicology, NUTRIM School for Nutrition, Toxicology and Metabolism, Maastricht University Medical Centre, 2Faculty of Health, Medicine and Life Sciences, Maastricht University, Maastricht, 3DSM Resolve, Geleen, 4Department of Internal Medicine, Division of Gastroenterology and Hepatology, Maastricht University Medical Centre, Maastricht, the Netherlands Backgr...

  5. Book Reviews

    Hans Antlöv; Harry A. Poeze; Holk H. Dengel; H.A.J. Klooster; Gill, Ronald G.; George Hotze; Chr.G.F. de Jong; Alle G. Hoekema; Liaw Yock Fang; Jan van der Putten; Brigitte Müller; Helmut Buchholt; M. Hekker; James J. Fox; Janet Carsten

    1996-01-01

    - R. Anderson Sutton, Wim van Zanten, Ethnomusicology in the Netherlands: present situation and traces of the past. Leiden: Centre of Non-Western Studies, Leiden University, 1995, ix + 330 pp. [Oideion; The performing arts worldwide 2. Special Issue]., Marjolijn van Roon (eds.) - T.E. Behrend, Willem Remmelink, The Chinese War and the collapse of the Javanese state, 1725-1743. Leiden: KITLV Press, 1994, 297 pp. [Verhandelingen 162]. - Erik Brandt, Eric Venbrux, A death in the Tiwi Islands; Co...

  6. Preparation of processed nuclear data libraries for thermal, fast and fusion research and power reactor applications. Texts of papers presented at the IAEA consultants' meeting

    The report contains 12 papers on nuclear data processing activities in Algeria, India, Indonesia, Italy, Japan, Republic of Korea, the Netherlands, Russia, Slovenia, United Kingdom, U.S.A., including ENDF formatted nuclear data libraries and computer code systems such as NJOY, AMPX, NSLINK, MCNP, multigroup data schemes such as WIMS, ABBN, and others. The role of the IAEA Nuclear Data Section in the establishment of nuclear data centers in developing countries is reviewed. (author). Refs, figs, tabs

  7. Validation Report for FY 1997--Final Report

    Pavlovichev, A.M.

    2001-09-28

    The report issued according to ''Work Release 02. P. 99-8'' presents a comparison of results on VVER Calculational Benchmarks computed with various codes: design code TVS-M and precision code MCU-REA elaborated in RRC KI, IPPE codes WIMS-ABBN, TRIANG-PWR and CONKEMO and 2-D fuel assembly analysis code HELIOS developed by Studsvik Scandpower.

  8. Book Reviews

    Jeremy Kemp; Nico G. Schulte Nordholt; Maureen A. MacKenzie; Richard Michael Bourke; Anton Ploeg; J.N. Sneddon; Bernhard Dahm; J. Noorduyn; Christin Kocher Schmid; Marie Alexandrine Martin; Mary F. Somers Heidhues; Elsbeth Locher-Scholten; Jonathon Falla; Ward Keeler; Ole Bruun

    1993-01-01

    - Anne Booth, W.L. Korthals Altes, Changing economy in Indonesia, Amsterdam: Royal Tropical Institute (General trade statistics, 1822-1949; volume 12a). - Wim van den Doel, Robert Cribb, Historical dictionary of Indonesia. Metuchen, N.J., & London: The Scarecrow Press, 1992. - C.D. Grijns, Kingsley Bolton, Sociolinguistics today; International perspectives. London and New York: Routledge, 1992, 383 pp., Helen Kwok (eds.) - David Henley, Ole Bruun, Asian perceptions of nature: Papers prese...

  9. Specifications for reactor physics experiments on CANFLEX-RU fuel

    This is to describe reactor physics experiments to be performed in the ZED-2 reactor to study CANFLEX-RU fuel bundles in CANDU-type fuel channels. The experiments are to provide benchmark quality validation data for the computer codes and associated nuclear databases used for physics calculations, in particular WIMS-AECL. Such validation data is likely to be a requirement by the regulator as condition for licensing a CANDU reactor based on an enriched fuel cycle

  10. Accelerating Into the Future: From 0 to GeV in a Few Centimeters (LBNL Summer Lecture Series)

    Leemans, Wim [LOASIS Program, AFRD

    2008-07-08

    Summer Lecture Series 2008: By exciting electric fields in plasma-based waveguides, lasers accelerate electrons in a fraction of the distance conventional accelerators require. The Accelerator and Fusion Research Division's LOASIS program, headed by Wim Leemans, has used 40-trillion-watt laser pulses to deliver billion-electron-volt (1 GeV) electron beams within centimeters. Leemans looks ahead to BELLA, 10-GeV accelerating modules that could power a future linear collider.

  11. Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

    Zare, Nafiseh [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of); Fadaei, Amir Hosein, E-mail: Fadaei_amir@aut.ac.i [Faculty of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnique), Hafez Street, Tehran (Iran, Islamic Republic of); Rahgoshay, Mohammad [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of); Fadaei, Mohammad Mehdi [Department of Electrical Engineering, Faculty of Engineering, Central Tehran Branch, Islamic Azad University, Punak Square, Tehran (Iran, Islamic Republic of); Kia, Shabnam [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of)

    2010-11-15

    Research highlights: {yields} Reactor behavior affects from reciprocal effects between neutronic and thermo-hydraulic. {yields} Reliable reactor analysis requires coupling of neutronic and thermal-hydraulic calculation. {yields} Iterative process can be used to perform neutronic and thermal-hydraulic calculation. - Abstract: In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.

  12. Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

    Research highlights: → Reactor behavior affects from reciprocal effects between neutronic and thermo-hydraulic. → Reliable reactor analysis requires coupling of neutronic and thermal-hydraulic calculation. → Iterative process can be used to perform neutronic and thermal-hydraulic calculation. - Abstract: In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.

  13. First detection of [N II] 205 micrometer absorption in interstellar gas

    Persson, C M; Mookerjea, B; Black, J H; Olberg, M; Goicoechea, J R; Hassel, G E; Falgarone, E; Levrier, F; Menten, K M; Pety, J

    2014-01-01

    We present high resolution [NII] 205 micrometer ^3P_1-^3P_0 spectra obtained with Herschel-HIFI towards a small sample of far-infrared bright star forming regions in the Galactic plane: W31C (G10.6-0.4), W49N (G43.2-0.1), W51 (G49.5-0.4) and G34.3+0.1. All sources display an emission line profile associated directly with the HII regions themselves. For the first time we also detect absorption of the [NII] 205 micrometer line by extended low-density foreground material towards W31C and W49N over a wide range of velocities. We attribute this absorption to the Warm Ionised Medium (WIM) and find N(N^+)\\approx 1.5x10^17 cm^-2 towards both sources. This is in agreement with recent Herschel-HIFI observations of [CII] 158 micrometer, also observed in absorption in the same sight-lines (Gerin et al. 2012, 2014), if ~10-13% of all C^+ ions exist in the WIM on average. Using an abundance ratio of [N]/[H] = 6.76x10^-5 in the gas-phase we find that the mean electron and proton densities are ~0.2-2 cm^-3 assuming a WIM fil...

  14. The warm ionized gas in CALIFA early-type galaxies: 2D emission-line patterns and kinematics for 32 galaxies

    Gomes, J M; Kehrig, C; Vílchez, J M; Lehnert, M D; Sánchez, S F; Ziegler, B; Breda, I; Reis, S N dos; Iglesias-Páramo, J; Bland-Hawthorn, J; Galbany, L; Bomans, D J; Rosales-Ortega, F F; Fernandes, R Cid; Walcher, C J; Falcón-Barroso, J; García-Benito, R; Márquez, I; del Olmo, A; Masegosa, J; Mollá, M; Marino, R A; Delgado, R M González; López-Sánchez, Á R

    2015-01-01

    The morphological, spectroscopic and kinematical properties of the warm interstellar medium (wim) in early-type galaxies (ETGs) hold key observational constraints to nuclear activity and the buildup history of these massive, quiescent systems. High-quality integral field spectroscopy (IFS) data with a wide spectral and spatial coverage, such as those from the CALIFA survey, offer an unprecedented opportunity for advancing our understanding of the wim in ETGs. This article centers on a 2D investigation of the wim component in 32 nearby (<~150Mpc) ETGs from CALIFA, complementing a previous 1D analysis of the same sample (Papaderos et al. 2013; P13). We include here H\\alpha\\ intensity and equivalent width (EW) maps and radial profiles, diagnostic emission-line ratios, besides ionized-gas and stellar kinematics. This study is supplemented by \\tau-ratio maps as an efficient means to quantify the role of photoionization by pAGB stars, as compared to other mechanisms (e.g., AGN, low-level star formation). Additio...

  15. The Survey for Ionization in Neutral-Gas Galaxies: III. Diffuse, Warm Ionized Medium and Escape of Ionizing Radiation

    Oey, M S; Yelda, S; Furst, E J; Caballero-Nieves, S M; Hanish, D J; Levesque, E M; Thilker, D A; Walth, G L

    2007-01-01

    We use the first data release from the SINGG H-alpha survey of HI-selected galaxies to study the quantitative behavior of the diffuse, warm ionized medium (WIM) across the range of properties represented by these 109 galaxies. The mean fraction f_WIM of diffuse ionized gas in this sample is 0.59+/- 0.19, slightly higher than found in previous samples. Since lower surface-brightness galaxies tend to have higher f_WIM, we believe that most of this difference is due to selection effects favoring large, optically-bright, nearby galaxies with high star-formation rates. As found in previous studies, there is no appreciable correlation with Hubble type or total star-formation rate. However, we find that starburst galaxies, defined here by an H-alpha surface brightness > 2.5x 10^39 erg s^-1 kpc^-2 within the H-alpha half-light radius, do show much lower fractions of diffuse H-alpha emission. The cause apparently is not dominated by a lower fraction of field OB stars. However, it is qualitatively consistent with an ex...

  16. Extended nebular emission in CALIFA early-type galaxies

    Gomes, J M; Kehrig, C; Vílchez, J M; Lehnert, M D

    2015-01-01

    The morphological, spectroscopic and kinematical properties of the warm interstellar medium (wim) in early-type galaxies (ETGs) hold key observational constraints to nuclear activity and the buildup history of these massive quiescent systems. High-quality integral field spectroscopy (IFS) data with a wide spectral and spatial coverage, such as those from the CALIFA survey, offer a precious opportunity for advancing our understanding in this respect. We use deep IFS data from CALIFA (califa.caha.es) to study the wim over the entire extent and optical spectral range of 32 nearby ETGs. We find that all ETGs in our sample show faint (H\\alpha\\ equivalent width EW~0.5...2 {\\AA}) extranuclear nebular emission extending out to >= 2 Petrosian_50 radii. Confirming and strengthening our conclusions in Papaderos et al. (2013) we argue that ETGs span a broad continuous sequence with regard to the properties of their wim, and they can be roughly subdivided into two characteristic classes. The first one (type i) comprises E...

  17. Definition and analysis of heavy water reactor benchmarks for testing new multigroup libraries

    A set of heavy water reactor benchmarks has been selected for testing new WIMS-D libraries. The libraries were constricted using data from ENDF/B-VI, Release 7, JENDL-3.2 and JEF-2.2 evaluated nuclear data files. The benchmarks cover a wide variety of reactor types and conditions, from fresh fuel to high burnup, and for natural and enriched uranium and Th-U fuels. The main parameters compared are the effective multiplication factor and other integral parameters, and isotopic composition of actinides on burnup cases. Besides, further investigations related with energy spectra used for preparation of WIMS-D libraries when applied on HWTR reactor calculations are included. Mostly of the benchmarks show a good agreement between experimental measurements and calculated values for all libraries. One exception is Th232 benchmark, were it is found that a library with JEND-3.2 Th232 data produces better results than ENDF/B-VI, R.7 and JEF-2.2 Th232 data. Results are slightly improved when HWTR spectra are used for weighting function to prepare the multi-group cross sections. This work is part of the International Atomic Energy Agency's Coordinated Research Project on 'Final Stage of WIMS-D Library Update Project'. (author)

  18. TRIGA MARK-II source term

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my; Rawi, M. Z. M. Rawi, E-mail: mark-dennis@nuclearmalaysia.gov.my; Abu, M. P., E-mail: mark-dennis@nuclearmalaysia.gov.my [Bahagian Teknologi Reaktor, Agensi Nuklear Malaysia, 43000 Kajang (Malaysia)

    2014-02-12

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  19. TRIGA MARK-II source term

    Usang, M. D.; Hamzah, N. S.; J. B., Abi M.; M. Z., M. Rawi; Abu, M. P.

    2014-02-01

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  20. Obtaining the neutronic and thermal hydraulic parameters of the VVER-1000 Bushehr nuclear reactor core by coupling nuclear codes

    In this research, the simulation of one-sixth of the VVER-1000 reactor core is carried out by the WIMS-D4 nuclear code, based on the symmetry of the core and information obtained from the Final Safety Analysis Report. The atomic densities of some important nuclear materials and fission poisons are calculated by the WIMS-D4 code at the end of the first fuel cycle. In addition, the cross sections of some nuclides are obtained by WIMS-D4, and they are transferred into the CITATION code as inputs. In the next stage, neutron flux and reactor power are calculated by the CITATION code in Cold Zero Power and Hot Full Power status, and subsequently the heat flux of the core is obtained. Then the results are returned again into the extended program cycle. Finally, the heat flux of the core is inputted into the COBRA code, and the temperatures of fuel, clad and coolant are calculated along the various distances applying the COBRA thermal hydraulic code, through the results of the CITATION code and also initial data as default created from the Final Safety Analysis Report. In conclusion, there are some interesting outcomes resulting form the obtained results.

  1. TRIGA MARK-II source term

    Full-text: ORIGEN 2.2 are employed to obtain data regarding g source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel. (author)

  2. Sampling optimization for high-speed weigh-in-motion measurements using in-pavement strain-based sensors

    Weigh-in-motion (WIM) measurement has been widely used for weight enforcement, pavement design, freight management, and intelligent transportation systems to monitor traffic in real-time. However, to use such sensors effectively, vehicles must exit the traffic stream and slow down to match their current capabilities. Hence, agencies need devices with higher vehicle passing speed capabilities to enable continuous weight measurements at mainline speeds. The current practices for data acquisition at such high speeds are fragmented. Deployment configurations and settings depend mainly on the experiences of operation engineers. To assure adequate data, most practitioners use very high frequency measurements that result in redundant samples, thereby diminishing the potential for real-time processing. The larger data memory requirements from higher sample rates also increase storage and processing costs. The field lacks a sampling design or standard to guide appropriate data acquisition of high-speed WIM measurements. This study develops the appropriate sample rate requirements as a function of the vehicle speed. Simulations and field experiments validate the methods developed. The results will serve as guidelines for future high-speed WIM measurements using in-pavement strain-based sensors. (paper)

  3. TRIGA MARK-II source term

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel

  4. A Study on the Development of the FEP and Scenario for the HLW Disposal in Korea

    The impacts influenced on the performance and safety of a repository are classified as units of Features, Events, and Processes (FEP), for the total system performance assessment (TSPA) related to the permanent disposal of HLW. The importance is evaluated in consideration of the frequency, consequence, regulation, suitability of a specific site, etc. and then these are grouped as a similar FEP. A scenario describing the migration of radionuclide from the repository to the biosphere is derived from understanding the interaction among these groups. KAERI has developed the KAERI FEP lists by review and collation of the foreign studies. The KAERI FEP list has been reviewed by several Korean experts. The five major scenarios describing possible future evolutions of the geological disposal system have been developed by RES and PID methods. Also the CYPRUS which is a KAERI integrated database management system for the total system performance assessment (TSPA) related to the permanent disposal of HLW has been developed and the results of the FEP and scenario development have been uploaded in this system.

  5. Radioactive wastes assay technique and equipment

    The waste inventory records such as the activities and radio- nuclides contained in the waste packages are to be submitted with the radioactive wastes packages for the final disposal. The nearly around 10,000 drums of waste stocked in KAERI now should be assayed for the preparation of the waste inventory records too. For the successive execution of the waste assay, the investigation into the present waste assay techniques and equipment are to be taken first. Also the installation of the waste assay equipment through the comprehensive design, manufacturing and procurement should be proceeded timely. As the characteristics of the KAERI-stocked wastes are very different from that of the nuclear power plant and those have no regular waste streams, the application of the in-direct waste assay method using the scaling factors are not effective for the KAERI-generated wastes. Considering for the versal conveniency including the accuracy over the wide range of waste forms and the combination of assay time and sensitivity, the TGS(Tomographic Gamma Scanner) is appropriate as for the KAERI -generated radioactive waste assay equipment

  6. Establishment of Korea-Russia bilateral research collaboration for studies on biological effects of cosmic ray and space radiation

    Lee, Juwoon; Kim, Dongho; Choi, Jongil; Song, Beomseok; Kim, Jaekyung; Kang, Oilhyun; Lee, Yoonjong; Kim, Jinhong; Jo, Minho

    2011-04-15

    {Omicron} KAERI-IBMP joint workshop on countermeasure and application researches to space environments - Sharing of state-of-the-art researches on space radiobiology using bio-satellites (BION-M1, Photon-soil) and ISS module (Bio-risk) was conducted - Sharing and discussion of state-of-the-art researches on dosimetry of space radiation and its affect on organisms were conducted. {Omicron} Making a contract on KAERI-IBMP Joint Research using Bio-risk module - Contract on KAERI-IBMP Joint Research to evaluate effect of space environment (microgravity and space radiation) on fermentative fungi (Aspergillus oryzae), Algae (Nostoc sp.), and plant seeds (rice, Arabidopsis thaliana, Brachypodium distachyon) was made in November, 2010. {Omicron} Discussion on new Joint Researches on evaluation of space radiation on organisms - Final step on Bion-M projects in terms of evaluation of physiological changes of lactic acid bacteria consumed by Mouse - Discussing new joint research on evaluation of physiological changes of primate by space radiation {Omicron} Establishment and management of the practical working group to invite a branch office of the IBMP in Korea - The system and the working group to implement cooperating researches between KAERI-IBMP on space radiation were established.

  7. Maintenance and fabrication of electronic equipment

    Solving the maintenance and repair problems of electronic instruments, we have supported the research and development work, and reduced operation cost of the pilot plants in KAERI. In addition, we have improved the maintainability of instruments to use effectively. (author). 10 refs., 18 tabs., 8 figs

  8. Maintenance and fabrication of electronic equipment

    Providing technical support to the maintenance and repair problems of electronic instruments, we have assisted the research and development work, and reduced operation cost of the pilot plants in KAERI. In addition, we have improved the performance of the data processing system of RMS, and also modified the cyclic box using PLC which is a facility for airborne monitoring in radiation area. (author)

  9. Nuclear instrumentation evaluation and analysis

    This project provides the program for improving instrumentation reliability as well as developing a cost-effective preventive maintenance activity through evaluation and analysis of nuclear instrumentation concerning pilot plants, large-scale test facilities and various laboratories on KAERI site. In addition, it discusses the program for enhancing safe operations and improving facility availability through establishment of maintenance technology. (Author)

  10. An Human Reliability Analysis to Identify Human Error Mechanisms for Reducing the Risks Associated with Human Errors in a Main Control Room of the SMART

    The research results are summarized as followed: (1) The task analysis performed on the EOGs of the SMART MMIS identified seven different human error mechanisms: Perception Error, Decision Error, Control-Identification Error, Control-Selection Error, Control-Execution Error, Communication Error, and Extraneous Error. The human error mechanisms includes 48 different human error types. 2) The design requirements were proposed to prevent 48 different possible human errors while running the HSI of SMART. 3) Sixteen different human errors were found for the SC designed by KAERI. Fifty six PSFs were also identified influencing the initiation of a human error mechanism. 4) Human factors design requirements were developed to hinder the human error mechanisms. CHED in KHU proposed a design alternative of the SC which took into account the human factors design requirements previously identified. 5) An human error quantification technique was applied to compare the CHED design with that the KAERI's in terms of the probabilities of the human errors caused by each design. The comparison showed that the CHD design was more effective than the KAERI's to reduce the human error probability from 0.0108 to 0.00004. It meant that 96.3% of the human error probability in the KAERI's was prevented by introducing the human factors design recommendations on the SC design

  11. Maintenance and fabrication of electronic equipment

    Han, Kwang Soo; Chung, Chong Eun; Park, Kwang Hyeon; Hong, Suk Boong; Moon, Je Sun; Choi, Myung Jin; Kim, Seung Bok; Kim, Jung Bok

    1996-12-01

    Solving the maintenance and repair problems of electronic instruments, we have supported the research and development work, and reduced operation cost of the pilot plants in KAERI. In addition, we have improved the maintainability of instruments to use effectively. (author). 10 refs., 18 tabs., 8 figs.

  12. Design report: An off gas trapping system for a voloxidizer in INL of US

    Jung, I. H.; Shin, J. M.; Park, J. J.; Park, G. I.; Lee, H. H

    2006-09-15

    This reports on the 'Development of Voloxidation Process for Treatment of LWR Spent Fuel', and it is the second year since it has started from June 2004 as a tripartite cooperation project among KAERI(Korea Atomic Energy Research Institute), INL(Idaho National Laboratory) and ORNL(Oak Ridge National Laboratory). This report is described mainly for the Task B2 accomplished during the second project year. The Task B2 in proposal contains two sub-tasks. The first one is design of an off-gas treatment system for a voloxidizer to be used in HFEF of INL. For this, KAERI team developed the design of INL OTS (Off-gas Treatment System) for hot experiment in the HFEF. INL team modified and completed the design of the INL OTS. The second task is manufacturing and test operation of the INL OTS for a voloxidizer in the INL. Manufacturing of the OTS is accomplished by INL team with co-work of KAERI. KAERI provided four sets of trapping filters needed for conducting hot experiment in the INL HFEF.

  13. Decontamination and decommissioning technology development of nuclear facilities

    Removal behaviour of an oxide which is similar in structure and composition to that on internal system of steam generator were investigated in low concentration chemical decontamination process [KAERI process]. In the AP solution (oxidative dissolution step), Cr dissolved fastly from the oxide in early stage and then dissolved very slowly in later stage. Dissolution behaviours of Fe from the oxides in the reductive dissolution process were similar to those of Cr in the oxidative dissolution process. Oxide dissolution behaviour in each process were discussed. In twice cyclic application of the oxidative and the reductive dissolution process(KAERI decontamination process), about 50% of the oxide was removed by chemical dissolution, about 40% by particulate detachment. The rest 10% oxide could be completely removed by ultrasonic decontamination. Corrosion acceptance guideline was established for the decontamination of domestic PWRs' steam generator. In the KAERI decontamination process, general corrosion to an Inconel-600 and 304 stainless steel was about 2.4 and 1.0% of general corrosion limit, respectively. And localized corrosion was not observed. Those results indicated that the KAERI decontamination process assured integrity of KNUs' steam generator. To evaluate the radioactive inventory for the decommissioning of nuclear facilities, general calculation methods of radioactive inventory, calculation and measurement of contact exposure rate, and confirmation of those results were reviewed. Feasibility for application of the above methods was examined by taking examples of radioactive inventory estimation in the Shippingport nuclear reactor vessel. (Author)

  14. Operation of the radioactive waste treatment facility

    The radioactive wasted generated at Korea Atomic Energy Research Institute (KAERI) in 1996 are about 118m3 of liquid waste and 204 drums of solid waste. Liquid waste were treated by the evaporation process, the bituminization process, and the solar evaporation process. In 1996, 100.5m3 of liquid waste was treated. (author). 84 tabs., 103 figs

  15. Establishment of Research Infrastructure for National Advanced Radiation Technology

    Kuk, Il Hiun; Byun, Myung Woo; Jeong, Il Yun; and others

    2007-07-15

    Construction of fundamental analysis system for RT/RFT advancement and pilot scale laboratory/facility for industry support and Assembly/installation of 30 MeV cyclotron for RI production and research utilizing positron beam, and construction of /distribution system for industrial and medical purpose were carried out for fast settlement for research environment of ARTI (a Jeongeup branch of KAERI)

  16. Maintenance and fabrication of electronic equipment

    Chung, Chong Eun; Moon, Byung Soo; Hong, Suk Boong; Kim, Jung Bok

    1999-12-01

    Providing technical support to the maintenance and repair problems of electronic instruments, we have assisted the research and development work, and reduced operation cost of the pilot plants in KAERI. In addition, we have improved the performance of the data processing system of RMS, and also modified the cyclic box using PLC which is a facility for airborne monitoring in radiation area. (author)

  17. A basic study on the international cooperation using international nuclear DB

    KAERI has collected the domestic data related to nuclear energy and sent the INIS a total of 1,296 items inputted according to the INIS bibliographic input rules. Korea inputted a total of 1,360 items to the IAEA INIS DB in 2000, and ranked 12th among the 103 INIS member countries in input amount. So Korea has reached to the level of advanced countries in input amount. In order to induce the INIS DB Mirror Site in Korea, the Korea INIS National Center has cooperated with KAERI and organizations concerned, contacted the INIS Secretariat and strengthened the relationship with Asian nations. As INIS liaison officers from each country agreed to induce the INIS DB Mirror Site in Korea, In 2002 the Mirror Site will be launched in Korea. The total records of our local OECD NEA documents are 2,742. KAERI will develop the OECD NEA documents retrival system and service will start. The KAERI has contributed to the improvement of domestic nuclear energy technology by inducing more OECD NEA computer codes from advanced countries and utilizing the codes

  18. Calculating Program for Decommissioning Work Productivity based on Decommissioning Activity Experience Data

    Song, Chan-Ho; Park, Seung-Kook; Park, Hee-Seong; Moon, Jei-kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    KAERI is performing research to calculate a coefficient for decommissioning work unit productivity to calculate the estimated time decommissioning work and estimated cost based on decommissioning activity experience data for KRR-2. KAERI used to calculate the decommissioning cost and manage decommissioning activity experience data through systems such as the decommissioning information management system (DECOMMIS), Decommissioning Facility Characterization DB System (DEFACS), decommissioning work-unit productivity calculation system (DEWOCS). In particular, KAERI used to based data for calculating the decommissioning cost with the form of a code work breakdown structure (WBS) based on decommissioning activity experience data for KRR-2.. Defined WBS code used to each system for calculate decommissioning cost. In this paper, we developed a program that can calculate the decommissioning cost using the decommissioning experience of KRR-2, UCP, and other countries through the mapping of a similar target facility between NPP and KRR-2. This paper is organized as follows. Chapter 2 discusses the decommissioning work productivity calculation method, and the mapping method of the decommissioning target facility will be described in the calculating program for decommissioning work productivity. At KAERI, research on various decommissioning methodologies of domestic NPPs will be conducted in the near future. In particular, It is difficult to determine the cost of decommissioning because such as NPP facility have the number of variables, such as the material of the target facility decommissioning, size, radiographic conditions exist.

  19. Present status and perspectives of residual stress measurement at HANARO reactor

    The residual stress instrument installed at HANARO reactor in KAERI is described and the results of residual stress measurements in different samples are described to depict the present status of the residual stress measurements at HANARO reactor. The perspectives for the future development of the instrument are outlined.

  20. Burn-Up Dependence of Bubble Morphology of Uranium Silicide Dispersion Fuels Used in Research Reactor

    Burn-up dependence of fission gas bubble morphology of U3Si2-Al and U3Si-Al dispersion fuels are reviewed with the data of ANL(Argonne Nation Laboratory) and KAERI(Korea Atomic Energy Research Institute

  1. Preliminary nuclear design for test MOX Fuel rods

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  2. Establishment of Research Infrastructure for National Advanced Radiation Technology

    Construction of fundamental analysis system for RT/RFT advancement and pilot scale laboratory/facility for industry support and Assembly/installation of 30 MeV cyclotron for RI production and research utilizing positron beam, and construction of /distribution system for industrial and medical purpose were carried out for fast settlement for research environment of ARTI (a Jeongeup branch of KAERI)

  3. Calculating Program for Decommissioning Work Productivity based on Decommissioning Activity Experience Data

    KAERI is performing research to calculate a coefficient for decommissioning work unit productivity to calculate the estimated time decommissioning work and estimated cost based on decommissioning activity experience data for KRR-2. KAERI used to calculate the decommissioning cost and manage decommissioning activity experience data through systems such as the decommissioning information management system (DECOMMIS), Decommissioning Facility Characterization DB System (DEFACS), decommissioning work-unit productivity calculation system (DEWOCS). In particular, KAERI used to based data for calculating the decommissioning cost with the form of a code work breakdown structure (WBS) based on decommissioning activity experience data for KRR-2.. Defined WBS code used to each system for calculate decommissioning cost. In this paper, we developed a program that can calculate the decommissioning cost using the decommissioning experience of KRR-2, UCP, and other countries through the mapping of a similar target facility between NPP and KRR-2. This paper is organized as follows. Chapter 2 discusses the decommissioning work productivity calculation method, and the mapping method of the decommissioning target facility will be described in the calculating program for decommissioning work productivity. At KAERI, research on various decommissioning methodologies of domestic NPPs will be conducted in the near future. In particular, It is difficult to determine the cost of decommissioning because such as NPP facility have the number of variables, such as the material of the target facility decommissioning, size, radiographic conditions exist

  4. Operation of the radioactive waste treatment facility

    Kim, Kil Jeong; Ahn, Seom Jin; Lee, Kang Moo; Lee, Young Hee; Sohn, Jong Sik; Bae, Sang Min; Kang, Kwon Ho; Lim, Kil Sung; Sohn, Young Joon; Kim, Tae Kook; Jeong, Kyung Hwan; Wi, Geum San; Park, Seung Chul; Park, Young Woong; Yoon, Bong Keun

    1996-12-01

    The radioactive wasted generated at Korea Atomic Energy Research Institute (KAERI) in 1996 are about 118m{sup 3} of liquid waste and 204 drums of solid waste. Liquid waste were treated by the evaporation process, the bituminization process, and the solar evaporation process. In 1996, 100.5m{sup 3} of liquid waste was treated. (author). 84 tabs., 103 figs.

  5. Operating procedures manual for fuel technology development facility

    To maintain the optimum condition of Fuel Technology Development Facility(FTDF) in KAERI, this report is described operating procedures manual for FTDF. The main topics of this report are as follows: - Policy - General management for FTDF - Electric facility - DG - UPS - HVAC - Fire protection system - Automatic control.

  6. Maintenance and fabrication of electronic equipment

    Han, Kwang Soo; Jeong, Jong Eun; Park, Kwang Hyun; Hong, Suk Bong; Moon, Je Sun; Choi, Myung Jin; Kim, Seung Bok; Kim, Jeong Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-12-01

    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.

  7. AN overview of the FLYSAFE datalink solution for the exchange of weather information: supporting aircrew decision making processes.

    Mirza, A.; Drouin, A.

    2009-09-01

    FLYSAFE is an Integrated Project of the 6th framework of the European Commission with the aim to improve flight safety through the development of an avionics solution the Next Generation Integrated Surveillance System (NGISS), which is supported by a ground based network of Weather Information Management Systems (WIMS) and access points in the form of the Ground Weather Processor (GWP). The NGISS provides information to the flight crew on the three major external hazards for aviation: weather, air traffic and terrain. The NGISS has the capability of displaying data about all three hazards on a single display screen, facilitating rapid appreciation of the situation by the flight crew. Weather Information Management Systems (WIMS) were developed to provide the NGISS and the flight crew with weather related information on in-flight icing, thunderstorms and clear-air turbulence. These products are generated on the ground from observations and model forecasts. WIMS will supply relevant information on three different scales: global, regional and local (over airport Terminal Manoeuvring Area). The Ground Weather Processor is a client-server architecture that utilises open source components, which include a geospatial database and web feature services. The GWP stores Weather Objects generated by the WIMS. An aviation user can retrieve on-demand all Weather Objects that intersect the volume of space that is of interest to them. The Weather Objects are fused with in-situ observation data and can be used by the flight management system to propose a route to avoid the hazard. In addition they can be used to display the current hazardous weather to the Flight Crew thereby raising their awareness. Within the FLYSAFE program, around 120 hours of flight trials were performed during February 2008 and August 2008. Two aircraft were involved each with separate objectives: - to assess FLYSAFE's innovative solutions for the data-link, on-board data-fusion and data-display and data

  8. Review on Overseas Contracts of a Nuclear Research Institute in Korea

    Since its establishment, Korea Atomic Energy Research Institute (KAERI) has made various contracts in research, design, engineering and consultation with a lot of foreign counterparts all over the world, including international organizations. As one of the global nuclear energy research leaders, KAERI can make a large scale contract because it has already procured a turnkey EPC (Engineering, Procurement, Construction) contract for a research and training reactor in the spring of 2010 by forming a consortium with a construction and engineering company. A contract in nuclear business industries is to be made under the limited control of regulatory authorities because the contractors must ensure nuclear safety and follow the international nuclear non-proliferation guidelines to secure the peaceful use of nuclear energy at an international level. The export and import of strategic technologies, products or materials (including nuclear materials) must be directly controlled by the authorities in accordance with the applicable law. In 2009, KAERI organized a new team to manage the overseas contracts and to make the limited control reflected in the contract documentation. In large scale project contracts, more attention shall be given to the contracts to prevent claims and also to the consideration of the regulatory requirements. In this context, the nature of the past KAERI contracts was reviewed. The conditions of several recent KAERI contracts were also individually reviewed based on the FIDIC (Federation Internationale des Ingenieurs-Conseils) model service agreement, which is generally accepted by service contractors. Ways to increase the quality of future contracts and to improve the standard model agreement which is used to prepare the draft contract were also considered

  9. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  10. Waste Information Management System-2012 - 12114

    The Waste Information Management System (WIMS) -2012 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 40 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and inter-modal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. WIMS continues to successfully accomplish the goals and objectives set forth by DOE for this project. It has

  11. Presence of inulin-type fructo-oligosaccharides and shift from raffinose family oligosaccharide to fructan metabolism in leaves of boxtree (Buxus sempervirens

    Wim eVan den Ende

    2016-03-01

    Full Text Available from raffinose family oligosaccharide to fructan metabolism in leaves of boxtree (Buxus sempervirens Wim Van den Ende1,* Marlies Coopman1, Rudy Vergauwen1, André Van Laere11 KU Leuven, Laboratory of Molecular Plant Biology, Institute of Botany and Microbiology, Kasteelpark Arenberg 31, B-3001 Leuven, Belgium* Correspondence: Wim Van den Ende, Laboratory of Molecular Plant Biology,Institute of Botany and Microbiology, Kasteelpark Arenberg 31, B-3001 Leuven, Belgium tel +32 16321952; fax +32 16321967;Wim.vandenende@bio.kuleuven.beKeywords: inulin, oligosaccharides, stress, RFO, fructanAbstractFructans are known to occur in 15% of flowering plants and their accumulation is often associated with stress responses. Typically, particular fructan types occur within particular plant families. The family of the Buxaceae, harbouring Pachysandra terminalis, an accumulator of graminan- and levan-type fructans, also harbours boxtree (Buxus sempervirens, a cold and drought tolerant species. Surprisingly, boxtree leaves do not accumulate the expected graminan- and levan-type fructans but small inulin fructo-oligosaccharides (FOS: 1-kestotriose and nystose and raffinose family oligosaccharides (RFO: raffinose and stachyose instead. The seasonal variation in concentrations of glucose, fructose, sucrose, FOS and RFO were followed. Raffinose and stachyose peaked during the winter months, while FOS peaked at a very narrow time-interval in spring, immediately preceded by a prominent sucrose accumulation. Sucrose may function as a reserve carbohydrate in winter and early spring leaves. The switch from RFO to fructan metabolism in spring strongly suggests that fructan and RFO fulfil distinct roles in boxtree leaves. RFO may play a key role in the cold acclimation of winter leaves while temporal fructan biosynthesis in spring might increase sink strength to sustain the formation of new shoots.

  12. Kinetic parameters evaluation of PWRs using static cell and core calculation codes

    Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((βeff)i)/((βeff)core) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((βeff)i)/((βeff)core) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.

  13. Study of the end flux peaking for the Candu fuel bundle types by transport methods

    The region separating the Candu fuel in two adjoining bundles in a channel is called the end region. The end of the last pellet in the fuel stack adjacent to the end region is called the fuel end. In the end region of the bundle the thermal neutron flux is higher than at the axial mid-point, because the end region of the bundle is made up of very low neutron absorption material: coolant and Zircaloy-4. For accurate evaluation of fuel performance, it is important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle, including the end region. The work reported here had two objectives. First, calculation of the flux distributions (axial and radial) and the end flux peaking factors for some Candu fuel bundles. Second objective is a comparative analysis of the obtained results. The Candu fuel bundles considered in this paper are NU37 (Natural Uranium, 37 elements) and SEU43 (Slightly Enriched Uranium, 43 elements, with 1.1wt% enrichment). For realization of the proposed objectives, a methodology based on WIMS, PIJXYZ and LEGENTR codes is used in this paper. WIMS is a standard lattice-cell code, based on transport theory and it is used for producing fuel cell multigroup macroscopic cross sections. For obtaining the flux distribution in Candu fuel bundles it is used PIJXYZ and LEGENTR respectively codes. These codes are consistent with WIMS lattice-cell calculations and allow a good geometrical representation of the Candu bundle in three dimensions. PIJXYZ is a 3D integral transport code using the first collision probability method and it has been developed for Candu cell geometry. LEGENTR is a 3D SN transport code based on projectors technique and can be used for 3D cell and 3D core calculations. (author)

  14. The Wisconsin Hα Mapper Northern Sky Survey

    Haffner, L. M.; Reynolds, R. J.; Madsen, G. J.; Tufte, S. L.; Jaehnig, K. P.; Percival, J. P.; Hausen, N. R.

    2001-12-01

    The ionized gas in the Milky Way has been fully surveyed from the Northern Hemisphere by the Wisconsin Hα Mapper (WHAM). The WHAM Northern Sky Survey (WHAM-NSS) has an angular resolution of one-degree and provides the first kinematically resolved map of the Warm Ionized Medium (WIM). With 12 km s-1 spectral resolution, we have removed atmospheric emission and zodiacal absorption features from each of the 37,565 spectra, leaving behind a fully resolved Galactic Hα profile. Galactic emission is detected in nearly every spectrum. Velocity channel maps from the survey show complex filamentary structure in the local WIM and in the nearest spiral arms. Some of these halo features are clearly associated with active star formation in the Galactic plane. High-latitude Hα emission at intermediate velocities traces out IVC complexes previously discovered through 21 cm observations. An initial analysis of the relationship between the high latitude Hα and 21 cm emission suggests that although the spatial extent and velocity profiles are quite similar, the intensities are completely uncorrelated. Our deep emission sensitivity also reveals several H 2 regions around early B stars and sdO stars, providing an indirect probe of their Lyman continuum and adding another ionizing source for the WIM. Total intensity maps, velocity channel maps, and full spectral profiles from the WHAM-NSS are available for download at http://www.astro.wisc.edu/wham/. WHAM was built and continues to explore the rich science of ionized gas through generous support of the National Science Foundation. This work is funded by grant AST96-19424.

  15. Waste Information Management System with 2012-13 Waste Streams - 13095

    Upadhyay, H.; Quintero, W.; Lagos, L.; Shoffner, P.; Roelant, D. [Applied Research Center, Florida International University, 10555 West Flagler Street, Suite 2100, Miami, FL 33174 (United States)

    2013-07-01

    The Waste Information Management System (WIMS) 2012-13 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 40 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and inter-modal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. (authors)

  16. Waste Information Management System with 2012-13 Waste Streams - 13095

    The Waste Information Management System (WIMS) 2012-13 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 40 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and inter-modal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. (authors)

  17. Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

    Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process

  18. A New Rose of Sharon Cultivar, 'Seonnyo' Developed by Mutation Breeding

    'Seonnyo' is a new Hibiscus variety developed by mutation breeding using gamma ray irradiation at Korea Atomic Energy Research Institute (KAERI). One hundred seeds of original variety, 'Gyewolhyang', collected in Namyangju of Gyeonggi Province were irradiated 10 Krad a-ray from a ∨60Co source at KAERI in 1993 (Fig. 1). The original variety, 'Gyewolhyang' showed the I-c that means single and bell-shaped flower type, and light purple with small red eye in its flower color. The irradiated seeds were sown in a field of the Atomic Experiment Farm in Namyangju, in April 1994. Forty-four out of one hundred seeds survived

  19. A Dwarf Type New Rose of Sharon Variety, 'Ggoma' Developed by a Mutation Breeding

    'Ggoma' is a new Hibiscus variety released by a mutation breeding using a gamma ray irradiation at the Korea Atomic Energy Research Institute (KAERI). One hundred seeds of the original native variety, 'Hongdansim 2', were collected from around a 35 year old plant, grown in a breeding field in Namyangju, Gyeonggi Province. The seeds were irradiated with a 100 Gy gamma ray from a ∨60Co source at KAERI in 1991 (Fig. 1). The original variety, 'Hongdansim 2' within the I-b category represents a single flower, with an intermediate petal width and a light purple color with a red eye in the center of its flower

  20. A New Rose of Sharon Cultivar, 'Daegoang' Developed by Mutation Breeding

    'Daegoang' is a new Hibiscus variety developed by mutation breeding using gamma ray irradiation at Korea Atomic Energy Research Institute (KAERI). One hundred seeds of original variety, 'Yongkwang', collected in Namyangju of Gyeonggi Province were irradiated 10 Krad γ-ray from a ∨60Co source at KAERI in 1994 (Fig. 1). The flower type and color of the variety 'Yongkwang' was I-b that means single and fully open, and light purple with red eye, respectively. The irradiated seeds were sown in a field of the Atomic Experiment Farm in Namyangju, in April 1994. The survival rate of seedling at the 0 and 100 Gy dose was 50% and 35%, respectively (Table 1)

  1. Environmental radiation monitoring around the nuclear facilities

    Lee, Chang Woo; Choi, Young Ho

    2000-02-01

    Environmental radiation monitoring was carried out with measurement of environment radiation and environmental radioactivity analysis around KAERI nuclear facilities and Seoul Research Reactor. The results of environmental radiation monitoring around KAERI nuclear facilities and Seoul Research Reactor are the follows: The average level of environmental radiation dose measured by NaI scintillation counter and accumulated radiation dose by TLD was almost some level compared with the past years. Gross {alpha}, {beta} radioactivity in environmental samples showed a environmental level. {gamma}-radionuclides in water sample were not detected. But only radionuclide K-40, which is natural radionuclide, was detected in the all samples and Cs-137 was detected in the surface soil and discharge sediment. The average level of environmental radiation dose around Seoul Research Reactor was almost same level compared with the past years, and Be-7 and Cs-137 were detected in some surface soil and discharge sediment by {gamma}-spectrometry. (author)

  2. Environmental radiation monitoring around the nuclear facilities

    Environmental radiation monitoring was carried out with measurement of environment radiation and environmental radioactivity analysis around KAERI nuclear facilities and Seoul research reactor. The results of environmental radiation monitoring around KAERI nuclear facilities and Seoul research reactor are the follows : The average level of environmental radiation dose measured by NaI scintillation counter and accumulated radiation dose by TLD was almost same level compared with the past years. Gross α, β radioactivity in environmental samples showed a environmental level. γ-radionuclides in water samples were not detected. But only radionuclide K-40, which is natural radionuclide, was detected in the all samples and Cs-137 was detected in the surface soil and discharge sediment. The average level of environmental radiation dose around Seoul research reactor was almost same level compared with the past years, and Be-7 and Cs-137 were detected in some surface soil and discharge sediment by γ-spectrometry. (author). 3 refs., 50 tabs., 12 figs

  3. RM Based Bilateral Regional Cooperation and its Perspective in Korea

    Remote monitoring (RM) is one alternative step to fulfill safeguards requirements in the member states. Korea installed a surveillance and unattended monitoring system in the ACPF (Advanced Spent Fuel Conditioning Process Facility) at the Korea Atomic Energy Research Institute (KAERI) in 2005. Data began to be shared through a virtual private network (VPN) started in 2006 with the Korea Institute of Nuclear Non-proliferation and Control (KINAC), KAERI, and the Sandia National Laboratories (SNL), as well. From 2009 the data are also being sent to the IAEA. Recently discussions have taken place to form a trilateral KINAC-SNL-JAEA (Japan Atomic Energy Agency) network using RM to strengthen the regional cooperative nonproliferation. The cooperation is supporting the basic ground of regional approaches for the peaceful use of nuclear energy. This paper addresses the main features of recent development to form a trilateral KINAC- SNL-JAEA network and a future prospective in nuclear nonproliferation and transparency via remote monitoring surveillance

  4. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    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

  5. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    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

  6. Qualification of the ONED90 code with measurements from Kori Unit 4

    A personal computer (PC) and workstation based one-dimensional computer program named ONED90 has been developed and used to simulate the core transient induced by a series of load-follow tests for Kori unit 4. The ONED90 code is evolved from the APCBOX code, which was jointly developed by Korea Atomic Energy Research Institute (KAERI) and Siemens-KWU. This paper highlights KAERI's verification effort of the ONED90 code. Results of comparisons with the actual measurement data obtained from Kori unit 4 are provided. These comparisons demonstrate that the code and methodology accurately model the anticipated core transient. The ONED90 code features modern methods that include the analytic nodal method with a nonlinear iteration scheme based on an equivalent discontinuity factor for a nuclear model, simplified thermal hydraulics, and an effective depletion model with cross-section derivatives with respect to the feedback mechanism for the accurate simulation of several days of reactor operation

  7. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for 330MWt SMART integral reactor

    The work reported in this document identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in a 330 MWt SMART integral reactor which is under development at KAERI. The result of this efforts is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts' knowledge and experience. The preliminary PIRT has been developed by the consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this report is intended for use to identify and integrate development areas of further experimental tests needed and thermal-hydraulic models and correlations and code improvements for the safety analysis of the SMART integral reactor. (author). 7 refs., 21 tabs., 22 figs

  8. Status of Fast Reactor and Pyroprocess Technology Development in the Republic of Korea

    A fast reactor system with pyroprocess technology is one of the most promising options for electricity generation, with an efficient utilization of uranium resources and a reduction of radioactive wastes. On the experience gained during the development of the conceptual designs for KALIMER reactors, Korea Atomic Energy Research Institute (KAERI) is currently developing advanced sodium cooled fast reactor (SFR) design concepts that can better meet the Generation IV technology goals. The long term SFR development plan will be carried out with the aim of constructing an advanced SFR demonstration plant by 2028. For the development of pyroprocess technology, KAERI is currently establishing a pyroprocess integrated inactive demonstration facility (PRIDE), a mock-up facility for pyroprocessing, to produce the engineering data to be incorporated into the design of an engineering scale pyrochemical process facility, which is scheduled to be constructed by 2016. (author)

  9. Prompt gamma activation analysis of boron in reference materials using diffracted polychromatic neutron beam

    Boron concentrations were analyzed for standard reference materials by prompt gamma activation analysis (PGAA). The measurements were performed at the SNU-KAERI PGAA facility installed at Hanaro, the research reactor of Korea Atomic Energy Research Institute (KAERI). The facility uses a diffracted polychromatic beam with a neutron flux of 7.9 x 107 n/cm2 s. Elemental sensitivity for boron was calibrated from the prompt gamma-ray spectra of boric acid samples containing 2-45 μg boron. The sensitivity of 2131 cps/mg-B was obtained from the linearity of the boron peak count rate versus the boron mass. The detection limit for boron was estimated to be 67 ng from an empty sample bag spectrum for a counting time of 10,000 s. The measured boron concentrations for standard reference materials showed good consistency with the certified or information values

  10. Biological efficiency of interaction between various radiation and chemicals

    This research project has been carried out jointly with INP (Poland) to develop technologies to assess the biological efficiency of interaction between radiation and chemicals. Through the cooperative project, KAERI and INP have established wide variety of bioassay techniques applicable to radiation bioscience, human monitoring, molecular epidemiology and environmental science. The joint experiment, in special, made it possible to utilize the merits of both institutes and to upgrade and verify KAERI's current technology level. All results of the cooperative research will be jointly published in high standard scientific journals listed in the Science Citation Index (SCI), which can make the role of fundamental basis for improving relationship between Korea and Poland. Research skills such as Trad-MCN assay, SCGE assay, immunohistochemical assay and molecular assay developed through joint research will be further elaborated and will be continuously used for the collaboration between two institutes

  11. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for SMART

    Chung, Bub Dong; Lee, Won Jae; Kim, Hee Cheol; Song, Jin Ho; Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART (System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART. 8 refs., 4 tabs (Author)

  12. A Study to Promote a Collaboration of R and D for Nuclear Energy Technology Development between Korea and Central Asia

    The goal of this research work is to promote the cooperation in the field of nuclear related works with Kazakhstan and Uzbekistan which are main countries in the central Asia. To make a basis on exchanging researchers, a staff in KAERI had visited INP NNC RK, touring research reactors, and make a discussion with the staffs in INP. With this, there is a efforts to make an arrangement between KAERI and NNC RK, signing will be made in near futures. To understand the level of nuclear technology in Uzbekistan, Dr. Rho had made a trip to INP. He visited the Gamma-radiation facility, a research reactor. In Nov. 2005, the chairman of Science Academy visited Korea to discuss the future cooperation in the field of peaceful use of the nuclear energy. By doing so, Korea will make an effort to cooperate with countries in the Central Asia

  13. A Study to Promote a Collaboration of R and D for Nuclear Energy Technology Development between Korea and Central Asia

    Rho, Si Pyo; Kim, Cheol Jung; Yoo, Bung Duk; Kim, Jae Woo; Lee, Myung Ho; Kim, Kyung Pyo

    2006-02-15

    The goal of this research work is to promote the cooperation in the field of nuclear related works with Kazakhstan and Uzbekistan which are main countries in the central Asia. To make a basis on exchanging researchers, a staff in KAERI had visited INP NNC RK, touring research reactors, and make a discussion with the staffs in INP. With this, there is a efforts to make an arrangement between KAERI and NNC RK, signing will be made in near futures. To understand the level of nuclear technology in Uzbekistan, Dr. Rho had made a trip to INP. He visited the Gamma-radiation facility, a research reactor. In Nov. 2005, the chairman of Science Academy visited Korea to discuss the future cooperation in the field of peaceful use of the nuclear energy. By doing so, Korea will make an effort to cooperate with countries in the Central Asia.

  14. Web-based networking within the framework of ANENT

    The Korea Atomic Energy Research Institute (KAERI) is actively participating in the Asian Network for Education in Nuclear Technology (ANENT), which is an IAEA activity to promote nuclear knowledge management. This has led KAERI to conduct a web-based networking for nuclear education and training in Asia. The networking encompasses the establishment of a relevant website and a system for a sustainable operation of the website. The established ANENT website features function as a database providing collected information, a link facilitating a systematic worldwide access to relevant websites, and an activity implementation for supporting the individual tasks of ANENT. The required information is being collected and loaded onto the database, and the website will be improved step by step. Consequently, networking is expected to play an important role, through cooperating with other networks, and thus contributing to a future global network for a sustainable development of nuclear technology. (author)

  15. The Korean nuclear ODA policy development

    Korean nuclear Official Development Assistance (ODA) is established with support from institutes such as the Korea International Cooperation Agency (KOICA) and the Korea Atomic Energy Research Institute (KAERI). KOICA's grant aid mainly made through the activities including IAEA's training program, and KAERI currently runs the inter-regional education and training cooperation called Asian Network for Education in Nuclear Technology(ANENT) which aimed to achieve the goal of encouraging web based education training network via cooperation with IAEA. Yet now these programs are focusing more on assisting nuclear infrastructure rather than highlighting nuclear education and training. This paper aims to, first, do a self-evaluation about the Korean ODA policy; second, to study the transition of the international nuclear atmosphere; and third, by apprehending the trend of the subjects of Korean nuclear ODA policy, to discuss the overall appropriate trajectory of Korean nuclear ODA

  16. Development of nuclear safety standards in Korea

    In 1982, the Nuclear Safety Center was established as an outfit of KAERI to perform Government-entrusted duties of nuclear safety regulation. In Feb. 1990, this Center was separated from KAERI and became an independent juristic body called the 'Korea Institute of Nuclear Safety' (hereinafter referred to as 'KINS'). KINS has developed various safety standards which are required for safety regulation of nuclear facilities, radioactive materials, etc. as a part of the entrusted duties since 1982. By the end of 1994, safety standards developed by KINS totalled 61 cases and among them 41 cases were notified by the Minister of Science and Technology. This paper discusses the concept of safety standards, the role between MOST (the Ministry of Science and Technology) and KINS in developing standards, details of standards developed and finally, future direction for the improvement of standards development

  17. Design report for cask transportation equipment

    In Korea, the spent fuels stored in the spent fuel storage pools in the domestic nuclear power plants significantly affects the continuation of the power plant operation. To solve this problem, KAERI has developed KSC-4 spent fuel shipping cask, which can transport 4 PWR spent fuel assemblies. Besides the development of the cask, KAERI developed transportation equipment which needed to use of KSC-4 cask. These equipment consist of cask handling tools such as lifting yoke, lid handling tool and spent fuel handling tool, etc. and transportation equipment such as trailer. In this report the usages, structures and functions of these tools and equipment were described, and the safety evaluation was carried out for each equipment

  18. Large break loss-of-coolant accident analysis for China Qinshan-2 nuclear power plant

    Large break LOCA analysis for China Qinshan-2 nuclear power plant has been performed using realistic evaluation model which has been being developed by KAERI. RELAP5/MOD3/KAERI code, which is a modified version of RELAP5/MOD3, is coupled with CONTEMPT4/MOD5 and is used as a best estimate code to predict the thermal hydraulic behavior of the system. PCT uncertainty which stems from code uncertainty, plant application uncertainty, scaling uncertainty and PCT bias are discussed. Among them, plant application uncertainty is described in detail. The licensing PCT is calculated by adding all the uncertainties to the best-estimate PCT. The result indicates the Qinshan-2 nuclear power plant has at least 37 deg C safety margin for large break LOCA. (Author) 10 refs., 47 figs., 14 tabs

  19. Development of the Decommissioning Project Management System, DECOMMIS

    Chung, U. S.; Park, J. H.; Lee, K. W.; Hwang, D. S.; Park, S. K.; Hwang, S. T.; Paik, S. T.; Choi, Y. D.; Chung, K. H.; Lee, K. I.; Hong, S. B

    2007-03-15

    At the Korea Atomic Energy Research Institute(KAERI), two projects for decommissioning of the research reactors and uranium conversion plant are carried out. The management of the projects can be defined as 'the decision of the changes of the decommissioning methodologies for the more efficient achievement of the project at an adequate time and to an improved method'. The correct decision comes from the experiences on the decommissioning project and the systematic experiences can be obtained from the good management of the decommissioning information. For this, a project management tool, DECOMMIS, was developed in the D and D Technology Division, which has the charge of the decommissioning projects at the KAERI, and its purpose was extended to following fields; generation of reports on the dismantling waste for WACID, record keeping for the next decommissioning projects of nuclear facilities, provision of fundamental data for the R and D of the decommissioning technologies.

  20. Preliminary evaluation of irradiation characteristics of new K alloys irradiated in HANARO

    Korean Atomic Energy Research Institute (KAERI) is trying various tests to develop zirconium based new alloys for nuclear fuel, which has better performance than that of Zircaloy-4 alloy. To evaluate the in-pile performance of newly developed K alloys preliminarily, KAERI had prepared the test specimens of K alloys, irradiated them upto the fluence of 8.63-9.27 x 1019n/cm2 at 320 ± 7 .deg. C in HANARO, and performed the hardness and tensile tests in IMEF. After the irradiation the hardness of K alloys increased from 24% to 37%, the yield strength from 17% to 37%, the ultimate tensile strength from 12% to 21% with the decease of maximum elongation from 6 to 39%