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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. 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

  5. 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.

  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. 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)

  18. 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

  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

    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)

  3. 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)

  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. 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)

  17. 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...

  18. 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)

  19. 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)

  20. 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)

  1. 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

  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. 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.

  4. 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

  5. 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)

  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. 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

  18. 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

  19. 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)

  20. 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.

  1. 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

  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. 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

  8. 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)

  9. 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)

  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. 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)

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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.)

  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 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.)

  19. 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.)

  20. 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.)

  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. 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.

  3. 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)

  4. 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.)

  5. 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)

  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

    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.

  10. 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

  11. 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

  12. 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.

  13. 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)

  14. 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)

  15. 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

  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. 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

  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. (author)

  20. 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.

  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. 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

  4. 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)

  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 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)

  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 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

  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. 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)

  18. 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

  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. 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

  3. 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

  4. 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)

  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. 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.

  10. 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)

  11. 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

  12. 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.

  13. 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.

  14. 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)

  15. 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

  16. 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.

  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. 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.

  7. 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

  8. 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.

  9. 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

  10. 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.)

  11. 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

  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. 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)

  14. 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

  15. 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).

  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. 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.

  2. 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

  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. 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

  5. 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)

  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. 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

  20. 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

  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. 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

  6. 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.

  7. 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.

  8. 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.

  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. 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

  13. 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...

  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. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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.

  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. 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)

  7. 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...

  8. 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...

  9. 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

  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. 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)

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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.

  3. 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

  4. 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

  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 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

  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