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Sample records for kaeri

  1. Development of tritium technologies at KAERI

    SciTech Connect

    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.; Yunn, S.H.; Jung, K.J.

    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.

  2. Radioecological studies in Korea atomic energy research institute, KAERI.

    PubMed

    Choi, Yong-Ho; Lim, Kwang-Muk; Jun, In; Keum, Dong-Kwon; Han, Moon-Hee

    2011-07-01

    Regarding the assessment of the terrestrial food chain dose to man, radioecology may be the field that is focused on the transfer of radionuclides from environmental media to food crops. In Korea, the environmental transfer of radionuclides to staple food crops have been investigated at Korea Atomic Energy Research Institute (KAERI) for the last 25 y mainly through radiotracer experiments in greenhouses. As a result, several hundreds of parameter values for the prediction of the radionuclide transfer have been produced. Many of them appear in two recent publications of International Atomic Energy Agency. This paper outlines the KAERI's past radioecological work and introduces the ongoing research and future plans. PMID:21525043

  3. Mapping Fractures in KAERI Underground Research Tunnel using Ground Penetrating Radar

    NASA Astrophysics Data System (ADS)

    Baek, Seung-Ho; Kim, Seung-Sep; Kwon, Jang-Soon

    2016-04-01

    The proportion of nuclear power in the Republic of Korea occupies about 40 percent of the entire electricity production. Processing or disposing nuclear wastes, however, remains one of biggest social issues. Although low- and intermediate-level nuclear wastes are stored temporarily inside nuclear power plants, these temporary storages can last only up to 2020. Among various proposed methods for nuclear waste disposal, a long-term storage using geologic disposal facilities appears to be most highly feasible. Geological disposal of nuclear wastes requires a nuclear waste repository situated deep within a stable geologic environment. However, the presence of small-scale fractures in bedrocks can cause serious damage to durability of such disposal facilities because fractures can become efficient pathways for underground waters and radioactive wastes. Thus, it is important to find and characterize multi-scale fractures in bedrocks hosting geologic disposal facilities. In this study, we aim to map small-scale fractures inside the KAERI Underground Research Tunnel (KURT) using ground penetrating radar (GPR). The KURT is situated in the Korea Atomic Energy Research Institute (KAERI). The survey target is a section of wall cut by a diamond grinder, which preserves diverse geologic features such as dykes. We conducted grid surveys on the wall using 500 MHz and 1000 MHz pulseEKKO PRO sensors. The observed GPR signals in both frequencies show strong reflections, which are consistent to form sloping planes. We interpret such planar features as fractures present in the wall. Such fractures were also mapped visually during the development of the KURT. We confirmed their continuity into the wall from the 3D GPR images. In addition, the spatial distribution and connectivity of these fractures are identified from 3D subsurface images. Thus, we can utilize GPR to detect multi-scale fractures in bedrocks, during and after developing underground disposal facilities. This study was

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

    SciTech Connect

    Choi, W. K.; Kim, K. N.; Won, H. J.; Jung, C. H.; Oh, W. Z.

    2003-02-25

    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 {approx} 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.

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

    SciTech Connect

    Lee, B.C.; Jeong, Y.U.; Cho, S.O.

    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.

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

    SciTech Connect

    Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H.

    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.

  7. High Power Proton Accelerator Development at KAERI and its Vacuum System

    NASA Astrophysics Data System (ADS)

    Choi, Byung-Ho; Park, Mi Young; Kim, Kui Young; Kim, Kye Ryung; Kim, Jun Yeon; Cho, Yong-Sub

    The Proton Engineering Frontier Project (PEFP), approved and launched by the Korean government in July 2002, includes a 100 MeV proton linear accelerator (linac) development and programs for its utilization and application. The main goals in the first phase of the project, spanning from 2002 to 2005, were the design of a 100 MeV proton linac and the development of a 20 MeV linac consisting of a 50 keV proton injector, a 3 MeV radio frequency quadrupole (RFQ), and a 20 MeV drift tube linac (DTL). The 50 keV injector and 3 MeV RFQ have been installed and tested, and the 20 MeV DTL is being assembled, tuned and under a beam test. At the same time, the utilization programs using the proton beam have been planned, and some are now under way. The vacuum system of the 20 MeV proton linac and its related issues, especially in operation with a high duty, are discussed in detail.

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

    SciTech Connect

    Reass, William A; Baca, David M; Gribble, Robert F

    2009-01-01

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

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

    SciTech Connect

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

    2012-01-30

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

  10. Thermal conductivity modeling of U-Mo/Al dispersion fuel

    NASA Astrophysics Data System (ADS)

    Kim, Yeon Soo; Cho, Byoung Jin; Sohn, Dong-Seong; Park, Jong Man

    2015-11-01

    A dataset for the thermal conductivity of U-Mo/Al dispersion fuel made available by KAERI was reanalyzed. Using this dataset, an analytical model was obtained by expanding the Bruggeman model. The newly developed model incorporates thermal resistances at the interface between the U-Mo particles and the Al matrix and the defects within the Al matrix (grain boundaries, cracks, and dislocations). The interfacial resistances are expressed as functions of U-Mo particle size and Al grain size obtained empirically by fitting to measured data from KAERI. The model was then validated against an independently measured dataset from ANL.

  11. Development of an ACP facility

    SciTech Connect

    Gil-Sung You; Won-Myung Choung; Jeong-Hoe Ku; il-Je Cho; Dong-Hak Kook; Kie-Chan Kwon; Eun-Pyo Lee; Ji-Sup Yoon; Seong-Won Park; Won-Kyung Lee

    2007-07-01

    KAERI has been developing an advanced spent fuel conditioning process (ACP). The ACP facility for a process demonstration consists of two air-sealed type hot cells. The safety analysis results showed that the facility was designed safely. The relevant integrated performance tests were also carried out successfully. (authors)

  12. Research of a Supercritical Pressure Water Cooled Reactor in Korea

    SciTech Connect

    Bae, Yoon-Yeong; Joo, Hyung-Kook; Jang, Jinsung; Jeong, Yong-Hwan; Song, Jin-ho; Yoon, Han-Young; Yoo, Jung-Yul

    2004-07-01

    In this paper the activities on the supercritical pressure water-cooled reactor (SCWR) in Korea are briefly introduced. Four projects on a SCWR are being conducted in Korea. Three of them are supported by the I-NERI program while one is by KAERI. Two of the I-NERI-supported projects concern suitable materials for supercritical pressure and temperature, and radiation environment. The other I-NERI-supported project surveys numerically and experimentally the proper turbulence modeling for the numerical calculation of heat transfer phenomena at a supercritical condition. Heat transfer at a supercritical condition is being studied at KAERI experimentally using carbon dioxide as a coolant. The test loop is to be completed by the end of 2004. (authors)

  13. Experience and Lessons Learned from Conditioning of Spent Sealed Sources in Singapore - 13107

    SciTech Connect

    Hong, Dae-Seok; Kang, Il-Sik; Jang, Kyung-Duk; Jang, Won-Hyuk; Hoo, Wee-Teck

    2013-07-01

    In 2010, IAEA requested KAERI (Korea Atomic Energy Research Institute) to support Singapore for conditioning spent sealed sources. Those that had been used for a lightning conductor, check source, or smoke detector, various sealed sources had been collected and stored by the NEA (National Environment Agency) in Singapore. Based on experiences for the conditioning of Ra-226 sources in some Asian countries since 2000, KAERI sent an expert team to Singapore for the safe management of spent sealed sources in 2011. As a result of the conditioning, about 575.21 mCi of Am-241, Ra-226, Co-60, and Sr-90 were safely conditioned in 3 concrete lining drums with the cooperation of the KAERI expert team, the IAEA supervisor, the NEA staff and local laborers in Singapore. Some lessons were learned during the operation: (1) preparations by a local authority are very helpful for an efficient operation, (2) a preliminary inspection by an expert team is helpful for the operation, (3) brief reports before and after daily operation are useful for communication, and (4) a training opportunity is required for the sustainability of the expert team. (authors)

  14. Fragility Analysis Methodology for Degraded Structures and Passive Components in Nuclear Power Plants - Illustrated using a Condensate Storage Tank

    SciTech Connect

    Nie, J.; Braverman, J.; Hofmayer, C.; Choun, Y.; Kim, M.; Choi, I.

    2010-06-30

    The Korea Atomic Energy Research Institute (KAERI) is conducting a five-year research project to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). The KAERI research project includes three specific areas that are essential to seismic probabilistic risk assessment (PRA): (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. Since 2007, Brookhaven National Laboratory (BNL) has entered into a collaboration agreement with KAERI to support its development of seismic capability evaluation technology for degraded structures and components. The collaborative research effort is intended to continue over a five year period. The goal of this collaboration endeavor is to assist KAERI to develop seismic fragility analysis methods that consider the potential effects of age-related degradation of structures, systems, and components (SSCs). The research results of this multi-year collaboration will be utilized as input to seismic PRAs. In the Year 1 scope of work, BNL collected and reviewed degradation occurrences in US NPPs and identified important aging characteristics needed for the seismic capability evaluations. This information is presented in the Annual Report for the Year 1 Task, identified as BNL Report-81741-2008 and also designated as KAERI/RR-2931/2008. The report presents results of the statistical and trending analysis of this data and compares the results to prior aging studies. In addition, the report provides a description of U.S. current regulatory requirements, regulatory guidance documents, generic communications, industry standards and guidance, and past research related to aging degradation of SSCs. In the Year 2 scope of work, BNL carried out a research effort to identify and assess degradation models for the long-term behavior of dominant materials that are

  15. KJRR-FAI Hydraulic Flow Testing Input Package

    SciTech Connect

    N.E. Woolstenhulme; R.B. Nielson; D.B. Chapman

    2013-12-01

    The INL, in cooperation with the KAERI via Cooperative Research And Development Agreement (CRADA), undertook an effort in the latter half of calendar year 2013 to produce a conceptual design for the KJRR-FAI campaign. The outcomes of this effort are documented in further detail elsewhere [5]. The KJRR-FAI was designed to be cooled by the ATR’s Primary Coolant System (PCS) with no provision for in-pile measurement or control of the hydraulic conditions in the irradiation assembly. The irradiation assembly was designed to achieve the target hydraulic conditions via engineered hydraulic losses in a throttling orifice at the outlet of the irradiation vehicle.

  16. Identification and Assessment of Material Models for Age-Related Degradation of Structures and Passive Components in Nuclear Power Plants

    SciTech Connect

    Nie,J.; Braverman, J.; Hofmayer, C.; Kim, M. K.; Choi, I-K.

    2009-04-27

    When performing seismic safety assessments of nuclear power plants (NPPs), the potential effects of age-related degradation on structures, systems, and components (SSCs) should be considered. To address the issue of aging degradation, the Korea Atomic Energy Research Institute (KAERI) has embarked on a five-year research project to develop a realistic seismic risk evaluation system which will include the consideration of aging of structures and components in NPPs. Three specific areas that are included in the KAERI research project, related to seismic probabilistic risk assessment (PRA), are probabilistic seismic hazard analysis, seismic fragility analysis including the effects of aging, and a plant seismic risk analysis. To support the development of seismic capability evaluation technology for degraded structures and components, KAERI entered into a collaboration agreement with Brookhaven National Laboratory (BNL) in 2007. The collaborative research effort is intended to continue over a five year period with the goal of developing seismic fragility analysis methods that consider the potential effects of age-related degradation of SSCs, and using these results as input to seismic PRAs. In the Year 1 scope of work BNL collected and reviewed degradation occurrences in US NPPs and identified important aging characteristics needed for the seismic capability evaluations that will be performed in the subsequent evaluations in the years that follow. This information is presented in the Annual Report for the Year 1 Task, identified as BNL Report-81741-2008 and also designated as KAERI/RR-2931/2008. The report presents results of the statistical and trending analysis of this data and compares the results to prior aging studies. In addition, the report provides a description of U.S. current regulatory requirements, regulatory guidance documents, generic communications, industry standards and guidance, and past research related to aging degradation of SSCs. This report

  17. Major results from safety-related integral effect tests with VISTA-ITL for the SMART design

    SciTech Connect

    Park, H. S.; Min, B. Y.; Shin, Y. C.; Yi, S. J.

    2012-07-01

    A series of integral effect tests (IETs) was performed by the Korea Atomic Energy Research Inst. (KAERI) using the VISTA integral test loop (VISTA-ITL) as a small-scale IET program. Among them this paper presents major results acquired from the safety-related IETs with the VISTA-ITL facility for the SMART design. Three small-break loss-of-coolant accident (SBLOCA) tests of safety injection system (SIS) line break, shutdown cooling system (SCS) line break and pressurizer safety valve (PSV) line break were successfully performed and the transient characteristics of a complete loss of flowrate (CLOF) was simulated properly with the VISTA-ITL facility. (authors)

  18. Korea's developmental program for superconductivity

    NASA Technical Reports Server (NTRS)

    Hong, Gye-Won; Won, Dong-Yeon; Kuk, Il-Hyun; Park, Jong-Chul

    1995-01-01

    Superconductivity research in Korea was firstly carried out in the late 70's by a research group in Seoul National University (SNU), who fabricated a small scale superconducting magnetic energy storage system under the financial support from Korea Electric Power Company (KEPCO). But a few researchers were involved in superconductivity research until the oxide high Tc superconductor was discovered by Bednorz and Mueller. After the discovery of YBaCuO superconductor operating above the boiling point of liquid nitrogen (77 K)(exp 2), Korean Ministry of Science and Technology (MOST) sponsored a special fund for the high Tc superconductivity research to universities and national research institutes by recognizing its importance. Scientists engaged in this project organized 'High Temperature Superconductivity Research Association (HITSRA)' for effective conducting of research. Its major functions are to coordinate research activities on high Tc superconductivity and organize the workshop for active exchange of information. During last seven years the major superconductivity research has been carried out through the coordination of HITSRA. The major parts of the Korea's superconductivity research program were related to high temperature superconductor and only a few groups were carrying out research on conventional superconductor technology, and Korea Atomic Energy Research Institute (KAERI) and Korea Electrotechnology Research Institute (KERI) have led this research. In this talk, the current status and future plans of superconductivity research in Korea will be reviewed based on the results presented in interim meeting of HITSRA, April 1-2, 1994. Taejeon, as well as the research activity of KAERI.

  19. Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

    SciTech Connect

    B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

    2008-07-01

    A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

  20. Fuel Safety Activities in Korea

    SciTech Connect

    Auh, Geun-Sun; Shin, A.D.; Lee, J.S.; Woo, S.W.; Ryu, Y.H.; Kim, Jun-Hwan; Kim, S.K.; Jeong, Y.H.

    2007-07-01

    The current regulatory requirements for fuel performance were based on earlier test data of fresh or low burnup Zircaloy fuels of less than 40 GWD/MTU. Most countries have not changed the current regulatory requirements even if they are actively investigating the high burnup and new cladding alloy effects. Korea agrees with commonly accepted international consensus that although there are technical issues requiring resolutions, these issues do not constitute immediate safety concerns. The high burnup fuel reactor performance experiences of Korea do not show any major problems even if there have been some burnup related fuel failures which are described in the paper. KINS has recommended the industry to have lower fuel failure rates than 1-2 per 50,000 fuel rods. A research project of High Burnup Fuel Safety Tests and Evaluations has started in 2002 under a joint cooperation of KAERI/KNFC/KEPRI and KINS to obtain performance results of high burnup fuel and to develop evaluation technologies of high burnup fuel safety issues. From 1998, KINS has closely monitored and actively participated in international activities such as OECD/NEA CABRI Water Loop Program to reflect on regulatory requirements if needed. KINS will closely monitor the high burnup fuel performances of Korea to strength the regulatory activities if needed. The research activities in Korea including of LOCA and RIA being performed at KAERI with active supports of the industry are summarized in the paper. (authors)

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

    PubMed

    Kim, J L; Kim, B H; Chang, S Y; Lee, J K

    1997-12-01

    This paper describes establishment by computational and experimental methods of the American National Standard Institute (ANSI) N13.11 X-ray radiation fields by the Korea Atomic Energy Research Institute (KAERI). These fields were used in the standard irradiations of various personal dosimeters for the personal dosimetry performance test program performed by the Ministry of Science and Technology of Korea in the autumn of 1995. Theoretical X-ray spectra produced from two KAERI X-ray generators were estimated using a modified Kramers' theory with target attenuation and backscatter correction and their spectral distributions experimentally measured by a high-purity germanium semiconductor detector through proper corrections for measured pulse height distributions with photopeak efficiency, Compton fraction, and K-escape fraction. The average energies and conversion coefficients obtained from the computation and experimental methods, when compared with ANSI N13.11 and the recently published National Institute of Standards and Technology X-ray beams, appeared to be in good agreement--(+/-)3% between corresponding values--and thus, could be satisfactorily applied in the performance test of personal dosimeters. PMID:9467054

  2. Gamma-ray generation using laser-accelerated electron beam

    NASA Astrophysics Data System (ADS)

    Park, Seong Hee; Lee, Ho-Hyung; Lee, Kitae; Cha, Yong-Ho; Lee, Ji-Young; Kim, Kyung-Nam; Jeong, Young Uk

    2011-06-01

    A compact gamma-ray source using laser-accelerated electron beam is being under development at KAERI for nuclear applications, such as, radiography, nuclear activation, photonuclear reaction, and so on. One of two different schemes, Bremsstrahlung radiation and Compton backscattering, may be selected depending on the required specification of photons and/or the energy of electron beams. Compton backscattered gamma-ray source is tunable and quasimonochromatic and requires electron beams with its energy of higher than 100 MeV to produced MeV photons. Bremsstrahlung radiation can generate high energy photons with 20 - 30 MeV electron beams, but its spectrum is continuous. As we know, laser accelerators are good for compact size due to localized shielding at the expense of low average flux, while linear RF accelerators are good for high average flux. We present the design issues for a compact gamma-ray source at KAERI, via either Bremsstrahlung radiation or Compton backscattering, using laser accelerated electron beams for the potential nuclear applications.

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

    PubMed Central

    Kim, J L; Kim, B H; Chang, S Y; Lee, J K

    1997-01-01

    This paper describes establishment by computational and experimental methods of the American National Standard Institute (ANSI) N13.11 X-ray radiation fields by the Korea Atomic Energy Research Institute (KAERI). These fields were used in the standard irradiations of various personal dosimeters for the personal dosimetry performance test program performed by the Ministry of Science and Technology of Korea in the autumn of 1995. Theoretical X-ray spectra produced from two KAERI X-ray generators were estimated using a modified Kramers' theory with target attenuation and backscatter correction and their spectral distributions experimentally measured by a high-purity germanium semiconductor detector through proper corrections for measured pulse height distributions with photopeak efficiency, Compton fraction, and K-escape fraction. The average energies and conversion coefficients obtained from the computation and experimental methods, when compared with ANSI N13.11 and the recently published National Institute of Standards and Technology X-ray beams, appeared to be in good agreement--(+/-)3% between corresponding values--and thus, could be satisfactorily applied in the performance test of personal dosimeters. PMID:9467054

  4. Development of a neutron measurement system in unified non-destructive assay for the PRIDE facility

    NASA Astrophysics Data System (ADS)

    Seo, Hee; Park, Se-Hwan; Won, Byung-Hee; Ahn, Seong-Kyu; Shin, Hee-Sung; Na, Sang-Ho; Song, Dae-Yong; Kim, Ho-Dong; Lee, Seung Kyu

    2013-12-01

    The Korea Atomic Energy Research Institute (KAERI) has made an effort to develop pyroprocessing technology to resolve an on-going problem in Korea, i.e., the management of spent nuclear fuels. To this end, a test-bed facility for pyroprocessing, called PRIDE (PyRoprocessing Integrated inactive DEmonstration facility), is being constructed at KAERI. The main objective of PRIDE is to evaluate the performance of the unit processes, remote operation, maintenance, and proliferation resistance. In addition, integrating all unit processes into a one-step process is also one of the main goals. PRIDE can also provide a good opportunity to test safeguards instrumentations for a pyroprocessing facility such as nuclear material accounting devices, surveillance systems, radiation monitoring systems, and process monitoring systems. In the present study, a non-destructive assay (NDA) system for the testing of nuclear material accountancy of PRIDE was designed by integrating three different NDA techniques, i.e., neutron, gamma-ray, and mass measurements. The developed neutron detection module consists of 56 3He tubes and 16 AMPTEK A111 signal processing circuits. The amplifiers were matched in terms of the gain and showed good uniformity after a gain-matching procedure (%RSD=0.37%). The axial and the radial efficiency distributions within the cavity were then measured using a 252Cf neutron source and were compared with the MCNPX calculation results. The measured efficiency distributions showed excellent agreement with the calculations, which confirmed the accuracy of the MCNPX model of the system.

  5. Travel time simulation for radionuclide transport at the Korean underground research facility, KURT

    NASA Astrophysics Data System (ADS)

    Ko, N.; Hwang, Y.; Jeong, J.; Kim, K.

    2013-12-01

    For the research on the deep geological disposal of radioactive waste, it is necessary to understand the underground environment, including the geology and hydrogeology. In Korea, KURT (KAERI Underground Research Tunnel) was constructed in 2006 at KAERI (Korea Atomic Energy Research Institute). Geological and hydrogeological field data have been obtained from the facility, and the groundwater flow system was simulated. Based on the data observed and analyzed on a groundwater flow system, the transport of potential radionuclides, which were assumed to be released at the supposed position, was then calculated in order to prepare the fundamental data for a safety assessment of a hypothetical underground repository. Several pathways with highly water-conductive features were selected to evaluate the elapsed times of radionuclide transport. The transport times were calculated using a TDRW (Time-Domain Random Walk) method. The matrix diffusion and sorption mechanisms in the host rock, as well as the advection-dispersion processes, were considered under the KURT field conditions. To reflect the radioactive decay, some decay chains were selected. The simulation results indicate that the main factors for the shapes of the mass discharge of the radionuclides were the half-life and distribution coefficient. This shows that the long-lived radionuclides must be treated accurately at the steps of determining radioactive waste source term as well as considering the transport process, and intensified research is required for the sorption between radionuclides and host rocks for making the safety assessment process more reliable and less uncertain.

  6. Dosimetric properties of the newly developed KLT-300 (LiF:Mg,Cu,Na,Si) TL detector.

    PubMed

    Lee, J I; Kim, J L; Chang, S Y; Chung, K S; Choe, H S

    2004-01-01

    The dosimetric properties of the newly developed KLT-300 (KAERI LiF:Mg,Cu,Na,Si TL detector) in KAERI (Korea Atomic Energy Research Institute) were investigated. The sensitivity of the TL detector was about 30 times higher than that of the TLD-100 by light integration. In the study of the dose linearity of the detector, the dose response was very linear up to 10 Gy and a sublinear response was observed at higher doses. The energy response of the detector was studied for photon energies from 20 to 662 keV. The results show that a maximum response of 1.004 at 53 keV and a minimum response of 0.825 at 20 keV were observed. The reproducibility study for the TL detector was also carried out. The coefficients of variation for each detector separately did not exceed 0.016, and for all the 10 detectors collectively it was 0.0054. IEC Standard requires that the coefficient of variation shall not exceed 0.075. So, the reproducibility of this new TL detector sufficiently satisfied the IEC requirements. A detection threshold of the detector was investigated and found to be 70 nGy by Harshaw 4500 TLD Reader. PMID:15856584

  7. Evaluation of Water-Mineral Interaction Using Microfluidic Tests with Thin Sections

    NASA Astrophysics Data System (ADS)

    Oh, Y. S.; Ryu, J. H.; Koh, Y. K.; Jo, H. Y.

    2014-12-01

    For the geological disposal of radioactive wastes, geological settings and groundwater conditions are significantly important because of their effects on a radionuclide migration. One of the preferred host rocks for the radioactive waste disposal is crystalline rock. Fractures in crystalline rocks are main fluid pathways. Groundwater reacts with fracture filling minerals in fracture zones, resulting in physicochemical changes in the minerals and groundwater. In this study, fracture filling mineral-groundwater interactions were investigated by conducting microfluidic tests using thin sections at various conditions (i.e., fluid chemistry and flow rate). Groundwater and rock core samples collected from the KAERI Underground Research Tunnel (KURT) located in the Korea Atomic Energy Research Institute (KAERI) were used in this study. Dominant bedrock is two-mica granite, which contains both biotite and muscovite. Secondary minerals (e.g., chlorite, calcite and clay minerals) occur in fracture and alteration zones. In nature, water-mineral interactions generally take long time. Microfluidic tests were conducted to simulate water-mineral interactions in shorter time with smaller scale. Thin sections of fracture filling minerals, minerals from alteration zones, and natural and synthetic groundwater samples were used for the microfluidic tests. Results showed that water-mineral interactions at various conditions caused changes in groundwater chemistry, dissolution of minerals, precipitation of secondary minerals, and formation of colloids, which can affect radionuclide migration. In addition, the fluid chemistry and flow rate affected characteristics of water-rock interactions.

  8. Microbiology and Biogeochemical Study of Underground Research Tunnel for the Geological Disposal of Nuclear Waste

    NASA Astrophysics Data System (ADS)

    Roh, Y.; Oh, J.; Seo, H.; Rhee, S.

    2007-12-01

    The Underground Research Tunnel (URT) located in Korea Atomic Energy Research Institute (KAERI), Daejeon, South Korea was recently constructed as an experimental site to study radionuclide transport, biogeochemistry, radionuclide-mineral interactions for the geological disposal of high level nuclear waste. Groundwater sampled from URT was used to examine microbial diversity and to enrich metal reducing bacteria for studying microbe- metal interactions. Genomic analysis indicated that the groundwater contained diverse microorganisms such as metal reducers, metal oxidizers, anaerobic denitrifying bacteria, and bacteria for reductive dechlorination. Metal- reducing bacteria enriched from the groundwater was used to study metal reduction and biomineralization. The metal-reducing bacteria enriched with acetate or lactate as the electron donors showed the bacteria reduced Fe(III)-citrate, Fe(III) oxyhydroxides, Mn(IV) oxide, and Cr(VI) as the electron acceptors. Preliminary study indicated that the enriched bacteria were able to use glucose, lactate, acetate, and hydrogen as electron donors while reducing Fe(III)-citrate or Fe(III) oxyhydroxide as the electron acceptor. The bacteria exhibited diverse mineral precipitation capabilities including the formation of magnetite, siderite, and rhodochrosite. The results indicated that Fe(III)- and metal-reducing communities are present in URT at the KAERI.

  9. A Study on the Tritium Behavior in the Rice Plant after a Short-Term Exposure of HTO

    SciTech Connect

    Yook, D-S.; Lee, K. J.; Choi, Y-H.

    2002-02-26

    In many Asian countries including Korea, rice is a very important food crop. Its grain is consumed by humans and its straw is used to feed animals. In Korea, there are four CANDU type reactors that release relatively large amounts of tritium into the environment. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out the experimental studies to obtain domestic data on various parameters concerning the direct contamination of plant. In this study, the behavior of tritium in the rice plant is predicted and compared with the measurement performed at KAERI. Using the conceptual model of the soil-plant-atmosphere tritiated water transport system which was suggested by Charles E. Murphy, tritium concentrations in the soil and in leaves to time were derived. If the effect of tritium concentration in the soil is considered, the tritium concentration in leaves is described as a double exponential model. On the other hand if the tritium concentration in the soil is disregarded, the tritium concentration in leaves is described by a single exponential term as other models (e.g. Belot's or STAR-H3 model). Also concentration of organically bound tritium in the seed is predicted and compared with measurements. The results can be used to predict the tritium concentration in the rice plant at a field around the site and the ingestion dose following the release of tritium to the environment.

  10. International Nuclear Energy Research Initiative Development of Computational Models for Pyrochemical Electrorefiners of Nuclear Waste Transmutation Systems

    SciTech Connect

    M.F. Simpson; K.-R. Kim

    2010-12-01

    In support of closing the nuclear fuel cycle using non-aqueous separations technology, this project aims to develop computational models of electrorefiners based on fundamental chemical and physical processes. Spent driver fuel from Experimental Breeder Reactor-II (EBR-II) is currently being electrorefined in the Fuel Conditioning Facility (FCF) at Idaho National Laboratory (INL). And Korea Atomic Energy Research Institute (KAERI) is developing electrorefining technology for future application to spent fuel treatment and management in the Republic of Korea (ROK). Electrorefining is a critical component of pyroprocessing, a non-aqueous chemical process which separates spent fuel into four streams: (1) uranium metal, (2) U/TRU metal, (3) metallic high-level waste containing cladding hulls and noble metal fission products, and (4) ceramic high-level waste containing sodium and active metal fission products. Having rigorous yet flexible electrorefiner models will facilitate process optimization and assist in trouble-shooting as necessary. To attain such models, INL/UI has focused on approaches to develop a computationally-light and portable two-dimensional (2D) model, while KAERI/SNU has investigated approaches to develop a computationally intensive three-dimensional (3D) model for detailed and fine-tuned simulation.

  11. Perform Tests and Document Results and Analysis of Oxide Layer Effects and Comparisons

    SciTech Connect

    Collins, E. D.; DelCul, G. D.; Spencer, B. B.; Hunt, R. D.; Ausmus, C.

    2014-08-30

    During the initial feasibility test using actual used nuclear fuel (UNF) cladding in FY 2012, an incubation period of 30–45 minutes was observed in the initial dry chlorination. The cladding hull used in the test had been previously oxidized in a dry air oxidation pretreatment prior to removal of the fuel. The cause of this incubation period was attributed to the resistance to chlorination of an oxide layer imparted by the dry oxidation pretreatment on the cladding. Subsequently in 2013, researchers at the Korea Atomic Energy Institute (KAERI) reported on their chlorination study [R1] on ~9-gram samples of unirradiated ZirloTM cladding tubes that had been previously oxidized in air at 500oC for various time periods to impart oxide layers of varying thickness. In early 2014, discussions with Indefinite Delivery, Indefinite Quantity (IDIQ) contracted technical consultants from Westinghouse described their previous development (and patents) [R2] on methods of chemical washing to remove some or all of the hydrous oxide layer imparted on UNF cladding during irradiation in light water reactors (LWRs) . Thus, the Oak Ridge National Laboratory (ORNL) study, described herein, was planned to extend the KAERI study on the effects of anhydrous oxide layers, but on larger ~100-gram samples of unirradiated zirconium alloy cladding tubes, and to investigate the effects of various methods of chemical pretreatment prior to chlorination with 100% chlorine on the average reaction rates and Cl2 usage efficiencies.

  12. Current and anticipated uses of thermal hydraulic codes in Korea

    SciTech Connect

    Kim, Kyung-Doo; Chang, Won-Pyo

    1997-07-01

    In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.

  13. Preparation of conducting silver paste with Ag nanoparticles prepared by e-beam irradiation

    NASA Astrophysics Data System (ADS)

    Sohn, Jong Hwa; Pham, Long Quoc; Kang, Hyun Suk; Park, Ji Hyun; Lee, Byung Cheol; Kang, Young Soo

    2010-11-01

    Conducting silver paste was prepared by using Ag nanoparticles which were synthesized by e-beam irradiation method (from KAERI); its conductivity was comparatively determined with Ag nanoparticles which were prepared by thermolysis method (commercial). The silver nanoparticles with the diameter of approximately 150 nm size prepared by e-beam irradiation were mixed with glass frit and sintered for 1 h at 500 °C. It is presumably concluded that the wt% of silver nanoparticle, size distribution and homogenous dispersibility of Ag nanoparticles in the pastes are the critical factors for the high conductivity of the paste. Among the various wt% of silver nanoparticle in the conducting silver pastes, silver paste with 90 wt% of silver nanoparticle has the highest conductivity as 1.6×10 4 S cm -1. This conductivity value is 1.6 times higher than the Ag pastes which were prepared with silver nanoparticles obtained by thermolysis method.

  14. High Density Fuel Development for Research Reactors

    SciTech Connect

    Daniel Wachs; Dennis Keiser; Mitchell Meyer; Douglas Burkes; Curtis Clark; Glenn Moore; Jan-Fong Jue; Totju Totev; Gerard Hofman; Tom Wiencek; Yeon So Kim; Jim Snelgrove

    2007-09-01

    An international effort to develop, qualify, and license high and very high density fuels has been underway for several years within the framework of multi-national RERTR programs. The current development status is the result of significant contributions from many laboratories, specifically CNEA in Argentina, AECL in Canada, CEA in France, TUM in Germany, KAERI in Korea, VNIIM, RDIPE, IPPE, NCCP and RIARR in Russia, INL, ANL and Y-12 in USA. These programs are mainly engaged with UMo dispersion fuels with densities from 6 to 8 gU/cm3 (high density fuel) and UMo monolithic fuel with density as high as 16 gU/cm3 (very high density fuel). This paper, mainly focused on the French & US programs, gives the status of high density UMo fuel development and perspectives on their qualification.

  15. Nuclear Instrumentation and Control Cyber Testbed Considerations – Lessons Learned

    SciTech Connect

    Jonathan Gray; Robert Anderson; Julio G. Rodriguez; Cheol-Kwon Lee

    2014-08-01

    Abstract: Identifying and understanding digital instrumentation and control (I&C) cyber vulnerabilities within nuclear power plants and other nuclear facilities, is critical if nation states desire to operate nuclear facilities safely, reliably, and securely. In order to demonstrate objective evidence that cyber vulnerabilities have been adequately identified and mitigated, a testbed representing a facility’s critical nuclear equipment must be replicated. Idaho National Laboratory (INL) has built and operated similar testbeds for common critical infrastructure I&C for over ten years. This experience developing, operating, and maintaining an I&C testbed in support of research identifying cyber vulnerabilities has led the Korean Atomic Energy Research Institute of the Republic of Korea to solicit the experiences of INL to help mitigate problems early in the design, development, operation, and maintenance of a similar testbed. The following information will discuss I&C testbed lessons learned and the impact of these experiences to KAERI.

  16. AFIP-1 Irradiation Summary Report

    SciTech Connect

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-05-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-1 was designed to demonstrate the performance of second-generation dispersion fuels at a prototypic scale with a length of 21.5 inches (54.6 cm), width of 2.25 inches (5.75 cm) and a thickness of 0.050 inch (0.13 cm). The experiment was fabricated using commercially standard practices at BWX Technology, Inc. (BWXT). The U-7Mo fuel particles were supplied by the Korean Atomic Energy Research Institute (KAERI) using equipment intended for commercial supply. Two fuel plates were tested that incorporated two different matrix compositions, Al-2Si and Al-4043.1 The following report summarizes the life of the AFIP-1 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results

  17. Developmental Status of Beam Position and Phase Monitor for PEFP Proton Linac

    NASA Astrophysics Data System (ADS)

    Park, Sungju; Park, Jangho; Yu, Inha; Kim, Dotae; Hwang, Jung-Yun; Nam, Sanghoon

    2004-11-01

    The PEFP (Proton Engineering Frontier Project) at the KAERI (Korea Atomic Energy Research Institute) is building a high-power proton linear accelerator aiming to generate 100-MeV proton beams with 20-mA peak current. (Pulse width and max. repetition rate of 1 ms and 120 Hz respectively.) We have developed the Beam Position and Phase Monitor (BPPM) for the machine that features the button-type PU, the full-analog processing electronics, and the EPICS-based control system. The beam responses of the button-type PU have been obtained using the MAGIC (Particle-In-Cell) code. The processing electronics has been developed in collaboration with Bergoz Instrumentation. In this article, we report the present status of the system developments except the control system.

  18. Development of the integrated control system for the microwave ion source of the PEFP 100-MeV proton accelerator

    NASA Astrophysics Data System (ADS)

    Song, Young-Gi; Seol, Kyung-Tae; Jang, Ji-Ho; Kwon, Hyeok-Jung; Cho, Yong-Sub

    2012-07-01

    The Proton Engineering Frontier Project (PEFP) 20-MeV proton linear accelerator is currently operating at the Korea Atomic Energy Research Institute (KAERI). The ion source of the 100-MeV proton linac needs at least a 100-hour operation time. To meet the goal, we have developed a microwave ion source that uses no filament. For the ion source, a remote control system has been developed by using experimental physics and the industrial control system (EPICS) software framework. The control system consists of a versa module europa (VME) and EPICS-based embedded applications running on a VxWorks real-time operating system. The main purpose of the control system is to control and monitor the operational variables of the components remotely and to protect operators from radiation exposure and the components from critical problems during beam extraction. We successfully performed the operation test of the control system to confirm the degree of safety during the hardware performance.

  19. Advanced spent fuel conditioning process (ACP) progress with respect to remote operation and maintenance

    SciTech Connect

    Lee, Hyo Jik; Lee, Jong Kwang; Park, Byung Suk; Yoon, Ji Sup

    2007-07-01

    Korea Atomic Energy Research Institute (KAERI) has been developing an Advanced Spent Fuel Conditioning Process (ACP) to reduce the volume of spent fuel, and the construction of the ACP facility (ACPF) for a demonstration of its technical feasibility has been completed. In 2006 two inactive demonstrations were performed with simulated fuels in the ACPF. Accompanied by process equipment performance tests, its remote operability and maintainability were also tested during that time. Procedures for remote operation tasks are well addressed in this study and evaluated thoroughly. Also, remote maintenance and repair tasks are addressed regarding some important modules with a high priority order. The above remote handling test's results provided a lot of information such as items to be revised to improve the efficiency of the remote handling tasks. This paper deals with the current status of ACP and the progress of remote handling of ACPF. (authors)

  20. Review of Recent Aging-Related Degradation Occurrences of Structures and Passive Components in U.S. Nuclear Power Plants

    SciTech Connect

    Nie,J.; Braverman, J.; Hofmayer, C.; Choun, Y.-S.; Kim, M.K.; Choi, I.-K.

    2009-04-02

    The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are collaborating to develop seismic capability evaluation technology for degraded structures and passive components (SPCs) under a multi-year research agreement. To better understand the status and characteristics of degradation of SPCs in nuclear power plants (NPPs), the first step in this multi-year research effort was to identify and evaluate degradation occurrences of SPCs in U.S. NPPs. This was performed by reviewing recent publicly available information sources to identify and evaluate the characteristics of degradation occurrences and then comparing the information to the observations in the past. Ten categories of SPCs that are applicable to Korean NPPs were identified, comprising of anchorage, concrete, containment, exchanger, filter, piping system, reactor pressure vessel, structural steel, tank, and vessel. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

  1. DEVELOPMENT OF ELECTROCHEMICAL REDUCTION TECHNOLOGY FOR SPENT OXIDE FUELS

    SciTech Connect

    Hur, Jin-Mok; Seo, Chung-Seok; Kim, Ik-Soo; Hong, Sun-Seok; Kang, Dae-Seung; Park, Seong-Won

    2003-02-27

    The Advanced Spent Fuel Conditioning Process (ACP) has been under development at Korea Atomic Energy Research Institute (KAERI) since 1997. The concept is to convert spent oxide fuel into metallic form and to remove high heat-load fission products such as Cs and Sr from the spent fuel. The heat power, volume, and radioactivity of spent fuel can decrease by a factor of a quarter via this process. For the realization of ACP, a concept of electrochemical reduction of spent oxide fuel in Li2O-LiCl molten salt was proposed and several cold tests using fresh uranium oxides have been carried out. In this new electrochemical reduction process, electrolysis of Li2O and reduction of uranium oxide are taking place simultaneously at the cathode part of electrolysis cell. The conversion of uranium oxide to uranium metal can reach more than 99% ensuring the feasibility of this process.

  2. An approach to developing an integrated pyroprocessing simulator

    SciTech Connect

    Lee, Hyo Jik; Ko, Won Il; Choi, Sung Yeol; Kim, Sung Ki; Kim, In Tae; Lee, Han Soo

    2014-02-12

    Pyroprocessing has been studied for a decade as one of the promising fuel recycling options in Korea. We have built a pyroprocessing integrated inactive demonstration facility (PRIDE) to assess the feasibility of integrated pyroprocessing technology and scale-up issues of the processing equipment. Even though such facility cannot be replaced with a real integrated facility using spent nuclear fuel (SF), many insights can be obtained in terms of the world's largest integrated pyroprocessing operation. In order to complement or overcome such limited test-based research, a pyroprocessing Modelling and simulation study began in 2011. The Korea Atomic Energy Research Institute (KAERI) suggested a Modelling architecture for the development of a multi-purpose pyroprocessing simulator consisting of three-tiered models: unit process, operation, and plant-level-model. The unit process model can be addressed using governing equations or empirical equations as a continuous system (CS). In contrast, the operation model describes the operational behaviors as a discrete event system (DES). The plant-level model is an integrated model of the unit process and an operation model with various analysis modules. An interface with different systems, the incorporation of different codes, a process-centered database design, and a dynamic material flow are discussed as necessary components for building a framework of the plant-level model. As a sample model that contains methods decoding the above engineering issues was thoroughly reviewed, the architecture for building the plant-level-model was verified. By analyzing a process and operation-combined model, we showed that the suggested approach is effective for comprehensively understanding an integrated dynamic material flow. This paper addressed the current status of the pyroprocessing Modelling and simulation activity at KAERI, and also predicted its path forward.

  3. Electrolytic Reduction of Spent Nuclear Oxide Fuel -- Effects of Fuel Form and Cathode Containment Materials on Bench-Scale Operations

    SciTech Connect

    S. D. Herrmann

    2007-09-01

    A collaborative effort between the Idaho National Laboratory (INL) and Korea Atomic Energy Research Institute (KAERI) is underway per an International Nuclear Energy Research Initiative to advance the development of a pyrochemical process for the treatment of spent nuclear oxide fuel. To assess the effects of specific process parameters that differ between oxide reduction operations at INL and KAERI, a series of 4 electrolytic reduction runs will be performed with a single salt loading of LiCl-Li2O at 650 °C using a test apparatus located inside of a hot cell at INL. The spent oxide fuel for the tests will be irradiated UO2 that has been subjected to a voloxidation process to form U3O8. The primary variables in the 4 electrolytic reduction runs will be fuel basket containment material and Li2O concentration in the LiCl salt. All 4 runs will be performed with comparable fuel loadings (approximately 50 g) and fuel compositions and will utilize a platinum anode and a Ni/NiO reference electrode. The first 2 runs will elucidate the effect of fuel form on the electrolytic reduction process by comparison of the above test results with U3O8 versus results from previous tests with UO2. The first 3 runs will investigate the impact that the cathode containment material has on the electrolytic reduction of spent oxide fuel. The 3rd and 4th runs will investigate the effect of Li2O concentration on the reduction process with a porous MgO cathode containment.

  4. An approach to developing an integrated pyroprocessing simulator

    NASA Astrophysics Data System (ADS)

    Lee, Hyo Jik; Ko, Won Il; Choi, Sung Yeol; Kim, Sung Ki; Kim, In Tae; Lee, Han Soo

    2014-02-01

    Pyroprocessing has been studied for a decade as one of the promising fuel recycling options in Korea. We have built a pyroprocessing integrated inactive demonstration facility (PRIDE) to assess the feasibility of integrated pyroprocessing technology and scale-up issues of the processing equipment. Even though such facility cannot be replaced with a real integrated facility using spent nuclear fuel (SF), many insights can be obtained in terms of the world's largest integrated pyroprocessing operation. In order to complement or overcome such limited test-based research, a pyroprocessing Modelling and simulation study began in 2011. The Korea Atomic Energy Research Institute (KAERI) suggested a Modelling architecture for the development of a multi-purpose pyroprocessing simulator consisting of three-tiered models: unit process, operation, and plant-level-model. The unit process model can be addressed using governing equations or empirical equations as a continuous system (CS). In contrast, the operation model describes the operational behaviors as a discrete event system (DES). The plant-level model is an integrated model of the unit process and an operation model with various analysis modules. An interface with different systems, the incorporation of different codes, a process-centered database design, and a dynamic material flow are discussed as necessary components for building a framework of the plant-level model. As a sample model that contains methods decoding the above engineering issues was thoroughly reviewed, the architecture for building the plant-level-model was verified. By analyzing a process and operation-combined model, we showed that the suggested approach is effective for comprehensively understanding an integrated dynamic material flow. This paper addressed the current status of the pyroprocessing Modelling and simulation activity at KAERI, and also predicted its path forward.

  5. SECOND GENERATION EXPERIMENTAL EQUIPMENT DESIGN TO SUPPORT VOLOXIDATION TESTING AT INL

    SciTech Connect

    Dennis L. Wahlquit; Kenneth J. Bateman; Brian R. Westphal

    2008-05-01

    Voloxidation is a potential head-end process used prior to aqueous or pyrochemical spent-oxide-fuel treatment. The spent oxide fuel is heated to an elevated temperature in oxygen or air to promote separation of the fuel from the cladding as well as volatize the fission products. The Idaho National Laboratory (INL) and the Korea Atomic Energy Research Institute (KAERI) have been collaborating on voloxidation research through a joint International Nuclear Energy Research Initiative (I-NERI). A new furnace and off-gas trapping system (OTS) with enhanced capability was necessary to perform further testing. The design criteria for the OTS were jointly agreed upon by INL and KAERI. First, the equipment must accommodate the use of spent nuclear fuel and be capable of operating in the Hot Fuel Examination Facility (HFEF) at the INL. This primarily means the furnace and OTS must be remotely operational and maintainable. The system requires special filters and distinctive temperature zones so that the fission products can be uniquely captured. The OTS must be sealed to maximize the amount of fission products captured. Finally, to accommodate the largest range of operating conditions, the OTS must be capable of handling high temperatures and various oxidizing environments. The constructed system utilizes a vertical split-tube furnace with four independently controlled zones. One zone is capable of reaching 1200°C to promote the release of volatile fission products. The three additional zones that capture fission products can be controlled to operate between 100-1100°C. A detailed description of the OTS will be presented as well as some initial background information on high temperature seal options.

  6. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    SciTech Connect

    Cole, C.M.; Dirk, W.J.; Cottam, R.E.; Paik, S.T.

    1997-10-09

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; and (5) coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included.

  7. Radiation hardness characteristics of Si-PIN radiation detectors

    NASA Astrophysics Data System (ADS)

    Jeong, Manhee; Jo, Woo Jin; Kim, Han Soo; Ha, Jang Ho

    2015-06-01

    The Korea Atomic Energy Research Institute (KAERI) has fabricated Si-PIN radiation detectors with low leakage current, high resistivity (>11 kΩ cm) and low capacitance for high-energy physics and X-ray spectroscopy. Floating-zone (FZ) 6-in. diameter N-type silicon wafers, with <1 1 1> crystal orientation and 675 μm thick, were used in the detector fabrication. The active areas are 3 mm×3 mm, 5 mm×5 mm and 10 mm×10 mm. We used a double deep-diffused structure at the edge of the active area for protection from the surface leakage path. We also compared the electrical performance of the Si-PIN detector with anti-reflective coating (ARC). For a detector with an active area of 3 mm×3 mm, the leakage current is about 1.9 nA and 7.4 nA at a 100 V reverse bias voltage, and 4.6 pF and 4.4 pF capacitance for the detector with and without an ARC, respectively. In addition, to compare the energy resolution in terms of radiation hardness, we measured the energy spectra with 57Co and 133Ba before the irradiation. Using developed preamplifiers (KAERI-PA1) that have ultra-low noise and high sensitivity, and a 3 mm×3 mm Si-PIN radiation detector, we obtained energy resolutions with 122 keV of 57Co and 81 keV of 133Ba of 0.221 keV and 0.261 keV, respectively. After 10, 100, 103, 104 and 105 Gy irradiation, we tested the characteristics of the radiation hardness on the Si-PIN radiation detectors in terms of electrical and energy spectra performance changes. The fabricated Si-PIN radiation detectors are working well under high dose irradiation conditions.

  8. Seismic Fragility Analysis of a Degraded Condensate Storage Tank

    SciTech Connect

    Nie, J.; Braverman, J.; Hofmayer, C.; Choun, Y-S.; Kim, M.K.; Choi, I-K.

    2011-05-16

    The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory are conducting a collaborative research project to develop seismic capability evaluation technology for degraded structures and components in nuclear power plants (NPPs). One of the goals of this collaboration endeavor is to develop seismic fragility analysis methods that consider the potential effects of age-related degradation of structures, systems, and components (SSCs). The essential part of this collaboration is aimed at achieving a better understanding of the effects of aging on the performance of SSCs and ultimately on the safety of NPPs. A recent search of the degradation occurrences of structures and passive components (SPCs) showed that the rate of aging related degradation in NPPs was not significantly large but increasing, as the plants get older. The slow but increasing rate of degradation of SPCs can potentially affect the safety of the older plants and become an important factor in decision making in the current trend of extending the operating license period of the plants (e.g., in the U.S. from 40 years to 60 years, and even potentially to 80 years). The condition and performance of major aged NPP structures such as the containment contributes to the life span of a plant. A frequent misconception of such low degradation rate of SPCs is that such degradation may not pose significant risk to plant safety. However, under low probability high consequence initiating events, such as large earthquakes, SPCs that have slowly degraded over many years could potentially affect plant safety and these effects need to be better understood. As part of the KAERI-BNL collaboration, a condensate storage tank (CST) was analyzed to estimate its seismic fragility capacities under various postulated degradation scenarios. CSTs were shown to have a significant impact on the seismic core damage frequency of a nuclear power plant. The seismic fragility capacity of the CST was developed

  9. I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels

    SciTech Connect

    S. Frank

    2009-09-01

    An attractive alternative to the once-through disposal of electrorefiner salt is to selectively remove the active fission products from the salt and recycle the salt back to the electrorefiner (ER). This would allow salt reuse for some number of cycles before ultimate disposal of the salt in a ceramic waste form. Reuse of ER salt would, thus, greatly reduce the volume of ceramic waste produced during the pyroprocessing of spent nuclear fuel. This final portion of the joint I-NERI research project is to demonstrate the separation of fission products from molten ER salt by two methods previously selected during phase two (FY-08) of this project. The two methods selected were salt/zeolite contacting and rare-earth fission product precipitation by oxygen bubbling. The ER salt used in these tests came from the Mark-IV electrorefiner used to anodically dissolved driver fuel from the EBR-II reactor on the INL site. The tests were performed using the Hot Fuel Dissolution Apparatus (HFDA) located in the main cell of the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels complex on the INL site. Results from these tests were evaluated during a joint meeting of KAERI and INL investigators to provide recommendations as to the future direction of fission product removal from electrorefiner salt that accumulate during spent fuel treatment. Additionally, work continued on kinetic measurements of surrogate quaternary salt systems to provide fundamental kinetics on the ion exchange system and to expand the equilibrium model system developed during the first two phases of this project. The specific objectives of the FY09 I-NERI research activities at the INL include the following: • Perform demonstration tests of the selected KAERI precipitation and INL salt/zeolite contacting processes for fission product removal using radioactive, fission product loaded ER salt • Continue kinetic studies of the quaternary Cs/Sr-LiCl-KCl system to determine the rate of ion

  10. Development of an S-band cavity-type beam position monitor for a high power THz free-electron laser

    SciTech Connect

    Noh, Seon Yeong; Kim, Eun-San Hwang, Ji-Gwang; Heo, A.; Won, Jang Si; Vinokurov, Nikolay A.; Jeong, Young UK Hee Park, Seong; Jang, Kyu-Ha

    2015-01-15

    A cavity-type beam position monitor (BPM) has been developed for a compact terahertz (THz) free-electron laser (FEL) system and ultra-short pulsed electron Linac system at the Korea Atomic Energy Research Institute (KAERI). Compared with other types of BPMs, the cavity-type BPM has higher sensitivity and faster response time even at low charge levels. When electron beam passes through the cavity-type BPM, it excites the dipole mode of the cavity of which amplitude depends linearly on the beam offset from the center of the cavity. Signals from the BPM were measured as a function of the beam offset by using an oscilloscope. The microtron accelerator for the KAERI THz FEL produces the electron beam with an energy of 6.5 MeV and pulse length of 5 μs with a micropulse of 10-20 ps at the frequency of 2.801 GHz. The macropulse beam current is 40 mA. Because the microtron provides multi-bunch system, output signal would be the superposition of each single bunch. So high output signal can be obtained from superposition of each single bunch. The designed position resolution of the cavity-type BPM in multi-bunch is submicron. Our cavity-type BPM is made of aluminum and vacuum can be maintained by indium sealing without brazing process, resulting in easy modification and cost saving. The resonance frequency of the cavity-type BPM is 2.803 GHz and the cavity-type BPM dimensions are 200 × 220 mm (length × height) with a pipe diameter of 38 mm. The measured position sensitivity was 6.19 (mV/mm)/mA and the measured isolation between the X and Y axis was −39 dB. By measuring the thermal noise of system, position resolution of the cavity-type BPM was estimated to be less than 1 μm. In this article, we present the test results of the S-band cavity-type BPM and prove the feasibility of the beam position measurement with high resolution using this device.

  11. Radioactivity determination of sealed pure beta-sources by surface dose measurements and Monte Carlo simulations

    NASA Astrophysics Data System (ADS)

    Choi, Chang Heon; Jung, Seongmoon; Choi, Kanghyuk; Son, Kwang-Jae; Lee, Jun Sig; Ye, Sung-Joon

    2016-04-01

    This study aims to determine the activity of a sealed pure beta-source by measuring the surface dose rate using an extrapolation chamber. A conversion factor (cGy s-1 Bq-1), which was defined as the ratio of surface dose rate to activity, can be calculated by Monte Carlo simulations of the extrapolation chamber measurement. To validate this hypothesis the certified activities of two standard pure beta-sources of Sr/Y-90 and Si/P-32 were compared with those determined by this method. In addition, a sealed test source of Sr/Y-90 was manufactured by the HANARO reactor group of KAERI (Korea Atomic Energy Research Institute) and used to further validate this method. The measured surface dose rates of the Sr/Y-90 and Si/P-32 standard sources were 4.615×10-5 cGy s-1 and 2.259×10-5 cGy s-1, respectively. The calculated conversion factors of the two sources were 1.213×10-8 cGy s-1 Bq-1 and 1.071×10-8 cGy s-1 Bq-1, respectively. Therefore, the activity of the standard Sr/Y-90 source was determined to be 3.995 kBq, which was 2.0% less than the certified value (4.077 kBq). For Si/P-32 the determined activity was 2.102 kBq, which was 6.6% larger than the certified activity (1.971 kBq). The activity of the Sr/Y-90 test source was determined to be 4.166 kBq, while the apparent activity reported by KAERI was 5.803 kBq. This large difference might be due to evaporation and diffusion of the source liquid during preparation and uncertainty in the amount of weighed aliquot of source liquid. The overall uncertainty involved in this method was determined to be 7.3%. We demonstrated that the activity of a sealed pure beta-source could be conveniently determined by complementary combination of measuring the surface dose rate and Monte Carlo simulations.

  12. Melting characteristics of the stainless steel generated from the uranium conversion plant

    SciTech Connect

    Choi, W.K.; Song, P.S.; Oh, W.Z.; Jung, C.H.; Min, B.Y.

    2007-07-01

    The partition ratio of cerium (Ce) and uranium (U) in the ingot, slag and dust phases has been investigated for the effect of the slag type, slag concentration and basicity in an electric arc melting process. An electric arc furnace (EAF) was used to melt the stainless steel wastes, simulated by uranium oxide and the real wastes from the uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). The composition of the slag former used to capture the contaminants such as uranium, cerium, and cesium during the melt decontamination process generally consisted of silica (SiO{sub 2}), calcium oxide (CaO) and aluminum oxide (Al{sub 2}O{sub 3}). Also, Calcium fluoride (CaF{sub 2} ), nickel oxide (NiO), and ferric oxide (Fe{sub 2}O{sub 3}) were added to provide an increase in the slag fluidity and oxidative potential. Cerium was used as a surrogate for the uranium because the thermochemical and physical properties of cerium are very similar to those of uranium. Cerium was removed from the ingot phase to slag phase by up to 99% in this study. The absorption ratio of cerium was increased with an increase of the amount of the slag former. And the maximum removal of cerium occurred when the basicity index of the slag former was 0.82. The natural uranium (UO{sub 2}) was partitioned from the ingot phase to the slag phase by up to 95%. The absorption of the natural uranium was considerably dependent on the basicity index of the slag former and the composition of the slag former. The optimum condition for the removal of the uranium was about 1.5 for the basicity index and 15 wt% of the slag former. According to the increase of the amount of slag former, the absorption of uranium oxide in the slag phase was linearly increased due to an increase of its capacity to capture uranium oxide within the slag phase. Through experiments with various slag formers, we verified that the slag formers containing calcium fluoride (CaF{sub 2}) and a high amount of silica were more

  13. Development of an S-band cavity-type beam position monitor for a high power THz free-electron laser

    NASA Astrophysics Data System (ADS)

    Noh, Seon Yeong; Kim, Eun-San; Hwang, Ji-Gwang; Heo, A.; won Jang, Si; Vinokurov, Nikolay A.; Jeong, Young UK; Hee Park, Seong; Jang, Kyu-Ha

    2015-01-01

    A cavity-type beam position monitor (BPM) has been developed for a compact terahertz (THz) free-electron laser (FEL) system and ultra-short pulsed electron Linac system at the Korea Atomic Energy Research Institute (KAERI). Compared with other types of BPMs, the cavity-type BPM has higher sensitivity and faster response time even at low charge levels. When electron beam passes through the cavity-type BPM, it excites the dipole mode of the cavity of which amplitude depends linearly on the beam offset from the center of the cavity. Signals from the BPM were measured as a function of the beam offset by using an oscilloscope. The microtron accelerator for the KAERI THz FEL produces the electron beam with an energy of 6.5 MeV and pulse length of 5 μs with a micropulse of 10-20 ps at the frequency of 2.801 GHz. The macropulse beam current is 40 mA. Because the microtron provides multi-bunch system, output signal would be the superposition of each single bunch. So high output signal can be obtained from superposition of each single bunch. The designed position resolution of the cavity-type BPM in multi-bunch is submicron. Our cavity-type BPM is made of aluminum and vacuum can be maintained by indium sealing without brazing process, resulting in easy modification and cost saving. The resonance frequency of the cavity-type BPM is 2.803 GHz and the cavity-type BPM dimensions are 200 × 220 mm (length × height) with a pipe diameter of 38 mm. The measured position sensitivity was 6.19 (mV/mm)/mA and the measured isolation between the X and Y axis was -39 dB. By measuring the thermal noise of system, position resolution of the cavity-type BPM was estimated to be less than 1 μm. In this article, we present the test results of the S-band cavity-type BPM and prove the feasibility of the beam position measurement with high resolution using this device.

  14. Development of the IPRO-zone for fire PSA and its applications

    SciTech Connect

    Kang, D. I.; Han, S. H.

    2012-07-01

    A PSA analyst has been manually determining fire-induced component failure modes and modeling them into the PSA logics. These can be difficult and time-consuming tasks as they need much information and many events are to be modeled. KAERI has been developing the IPRO-ZONE (interface program for constructing zone effect table) to facilitate fire PSA works for identifying and modeling fire-induced component failure modes, and to construct a one top fire event PSA model. With the output of the IPRO-ZONE, the AIMS-PSA, and internal event one top PSA model, one top fire events PSA model is automatically constructed. The outputs of the IPRO-ZONE include information on fire zones/fire scenarios, fire propagation areas, equipment failure modes affected by a fire, internal PSA basic events corresponding to fire-induced equipment failure modes, and fire events to be modeled. This paper introduces the IPRO-ZONE, and its application results to fire PSA of Ulchin Unit 3 and SMART(System-integrated Modular Advanced Reactor). (authors)

  15. Development of a gadolinium-loaded liquid scintillator for the Hanaro short baseline prototype detector

    NASA Astrophysics Data System (ADS)

    Yeo, In Sung; Joo, Kyung Kwang; So, Sun Heang; Song, Sook Hyung; Kim, Hong Joo; So, Jung Ho; Park, Kang Soon; Ma, Kyung Ju; Jeon, Eun Ju; Kim, Jin Yu; Kim, Young Duk; Lee, Jason; Lee, Jeong-Yeon; Sun, Gwang-Min

    2014-02-01

    We propose a new experiment on the site of the Korea Atomic Energy Research Institute (KAERI) located at Daejeon, Korea. The Hanaro short baseline (SBL) nuclear reactor with a thermal power output 30 MW is used to investigate a reactor neutrino anomaly. A Hanaro SBL prototype detector having a 60- l volume has been constructed ˜6 m away from the reactor core. A gadolinium (Gd)-loaded liquid scintillator (LS) is used as an active material to trigger events. The selection of the LS is guided by physical and technical requirements, as well as safety considerations. A linear alkyl benzene (LAB) is used as a base solvent of the Hanaro SBL prototype detector. Three g/ l of PPO and 30 mg/ l of bis-MSB are dissolved to formulate the LAB-based LS. Then, a 0.5% gadolinium (Gd) complex with carboxylic acid is loaded into the LAB-based LS by using the liquidliquid extraction method. In this paper, we will summarize all the characteristics of the Gd-loaded LAB-based LS for the Hanaro prototype detector.

  16. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  17. On the multidimensional modeling of fluid flow and heat transfer in SCWRS

    SciTech Connect

    Gallaway, T.; Antal, S. P.; Podowski, M. Z.

    2012-07-01

    The Supercritical Water Reactor (SCWR) has been proposed as one of the six Generation IV reactor design concepts under consideration. The key feature of the SCWR is that water at supercritical pressures is used as the reactor coolant. Although at such pressures, fluids do not undergo phase change as they are heated, the fluid properties experience dramatic variations throughout what is known as the pseudo-critical region. Highly nonuniform temperature and fluid property distributions are expected in the reactor core, which will have a significant impact on turbulence and heat transfer in future SCWRs. The goal of the present work has been to understand and predict the effects of these fluid property variations on turbulence and heat transfer throughout the reactor core. Spline-type property models have been formulated for water at supercritical pressures in order to include the dependence of properties on both temperature and pressure into a numerical solver. New models of turbulence and heat transfer for variable-property fluids have been developed and implemented into the NPHASE-CMFD software. The results for these models have been compared to experimental data from the Korea Atomic Energy Research Inst. (KAERI) for various heat transfer regimes. It is found that the Low-Reynolds {kappa}-{epsilon} model performs best at predicting the experimental data. (authors)

  18. The severe accident research programme PHEBUS F.P.: First results and future tests

    SciTech Connect

    Schwarz, M.; Hardt, P. von der

    1996-03-01

    PHEBUS FP is an international programme, managed by the French Institut de Protection et de Surete Nucleaire, Electricite de France and the European Commission in close collaboration with the USNRC (US), COG (Canada), NUPEC and JAERI (Japan) and KAERI (South Korea). Its objective is to investigate through a series of in-pile integral experiments, key phenomena involved in LWR severe accident such as the degradation of core materials up to molten pool, the subsequent release of fission products and of structural materials, their transport in the cooling system and their deposition in the containment with a special emphasis on the volatility of iodine. After a general programme description, the paper focuses on the status of analysis of the first test FPT-0, which involved trace irradiated fuel and which has shown some quite unexpected results regarding fuel degradation and iodine behaviour, and on the upcoming test FPT-1 which will use irradiated fuel. The status of the preparation of the remaining tests of the programme is also presented.

  19. Feasibility study of a gamma camera for monitoring nuclear materials in the PRIDE facility

    NASA Astrophysics Data System (ADS)

    Jo, Woo Jin; Kim, Hyun-Il; An, Su Jung; Lee, Chae Young; Song, Han-Kyeol; Chung, Yong Hyun; Shin, Hee-Sung; Ahn, Seong-Kyu; Park, Se-Hwan

    2014-05-01

    The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocessing technology, in which actinides are recovered together with plutonium. There is no pure plutonium stream in the process, so it has an advantage of proliferation resistance. Tracking and monitoring of nuclear materials through the pyroprocess can significantly improve the transparency of the operation and safeguards. An inactive engineering-scale integrated pyroprocess facility, which is the PyRoprocess Integrated inactive DEmonstration (PRIDE) facility, was constructed to demonstrate engineering-scale processes and the integration of each unit process. the PRIDE facility may be a good test bed to investigate the feasibility of a nuclear material monitoring system. In this study, we designed a gamma camera system for nuclear material monitoring in the PRIDE facility by using a Monte Carlo simulation, and we validated the feasibility of this system. Two scenarios, according to locations of the gamma camera, were simulated using GATE (GEANT4 Application for Tomographic Emission) version 6. A prototype gamma camera with a diverging-slat collimator was developed, and the simulated and experimented results agreed well with each other. These results indicate that a gamma camera to monitor the nuclear material in the PRIDE facility can be developed.

  20. Modeling in-situ transport of uranine and colloids in the fracture network in KURT.

    PubMed

    Kim, Jung-Woo; Lee, Jae-Kwang; Baik, Min-Hoon; Jeong, Jongtae

    2015-02-01

    An in-situ dipole migration experiment was conducted using the conservative tracer uranine and latex colloids in KAERI (Korea Atomic Energy Research Institute) Underground Research Tunnel (KURT). The location and dimensions of the fractures between the two boreholes were estimated using the results of a borehole image processing system (BIPS) investigation, and the connectivity of the fractures was evaluated by a packer test. To investigate the flow and transport of uranine and colloids through an in-situ fracture network, a fracture network transport model was newly developed. The model consists of a series of one-dimensional advection-dispersion-matrix diffusion equations for each channel of the fracture network. Using the fracture network transport model, the most probable representation and the hydrologic parameters of the fracture network can be estimated by fitting the breakthrough of uranine. While the fracture network might not be unique, the representation chosen was adequate to describe the breakthrough of uranine and it represents a reasonable approach to modeling transport in the fracture network. An additional evaluation showed that the colloid transport in this study was influenced by filtration on the fracture surface rather than the enhancement of the colloid velocity. Overall, the model can explain successfully the in-situ experimental results of uranine and colloid transports through the fracture network. PMID:25543462

  1. Technology Verification of the Advanced Integral Reactor SMART

    SciTech Connect

    Si-Hwan Kim; Young-Dong Hwang; Hee-Chul Kim; Sung-Quun Zee

    2006-07-01

    SMART(System-Integrated Modular Advanced Reactor) is an integral type advanced pressurized water reactor with a rated thermal power of 330 MW, developed at KAERI (Korea Atomic Energy Research Institute) for a seawater desalination and small scale electricity generation. Safety and economic improvement are the two most important considerations in the design of the SMART. The SMART design combines firmly established commercial reactor design technologies with advanced design features. The advanced design features and technologies implemented into the SMART design have been proven or will be qualified through the technology verification program of SMART. Technology verification program of SMART consists of basic thermal-hydraulic experiments, separate effect test, major components performance test, system integrated tests of safety system and one fifth scaled pilot plant construction project. The overall performance and safety of SMART will be demonstrated through the SMART-pilot plant (SMART-P). The SMART-P plant construction project is currently underway and will be complete the construction by 2010. (authors)

  2. Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization

    SciTech Connect

    Versey, Joshua R.

    2013-08-01

    This work contributes to the development of pyroprocessing technology as an economically viable means of separating used nuclear fuel from fission products and cladding materials. Electrolytic oxide reduction is used as a head-end step before electrorefining to reduce oxide fuel to metallic form. The electrolytic medium used in this technique is molten LiCl-Li2O. Groups I and II fission products, such as cesium (Cs) and strontium (Sr), have been shown to partition from the fuel into the molten LiCl-Li2O. Various approaches of separating these fission products from the salt have been investigated by different research groups. One promising approach is based on a layer crystallization method studied at the Korea Atomic Energy Research Institute (KAERI). Despite successful demonstration of this basic approach, there are questions that remain, especially concerning the development of economical and scalable operating parameters based on a comprehensive understanding of heat and mass transfer. This research explores these parameters through a series of experiments in which LiCl is purified, by concentrating CsCl in a liquid phase as purified LiCl is crystallized and removed via an argon-cooled cold finger.

  3. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    NASA Astrophysics Data System (ADS)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-10-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

  4. Development progresses of radio frequency ion source for neutral beam injector in fusion devices

    NASA Astrophysics Data System (ADS)

    Chang, D. H.; Jeong, S. H.; Kim, T. S.; Park, M.; Lee, K. W.; In, S. R.

    2014-02-01

    A large-area RF (radio frequency)-driven ion source is being developed in Germany for the heating and current drive of an ITER device. Negative hydrogen ion sources are the major components of neutral beam injection systems in future large-scale fusion experiments such as ITER and DEMO. RF ion sources for the production of positive hydrogen (deuterium) ions have been successfully developed for the neutral beam heating systems at IPP (Max-Planck-Institute for Plasma Physics) in Germany. The first long-pulse ion source has been developed successfully with a magnetic bucket plasma generator including a filament heating structure for the first NBI system of the KSTAR tokamak. There is a development plan for an RF ion source at KAERI to extract the positive ions, which can be applied for the KSTAR NBI system and to extract the negative ions for future fusion devices such as the Fusion Neutron Source and Korea-DEMO. The characteristics of RF-driven plasmas and the uniformity of the plasma parameters in the test-RF ion source were investigated initially using an electrostatic probe.

  5. Smart measurement system for an environmental radiation monitoring

    NASA Astrophysics Data System (ADS)

    Lee, Wanno; Kim, Hee-Reyoung; Chung, Kun-Ho; Kim, Eun-Han; Cho, Young Hyun; Choi, Geun Sik; Lee, Chang Woo

    2007-08-01

    A smart measurement system for an on-line gamma monitoring has been developed to overcome the problems of a conventional system which cannot automatically restore the previous-lost data of several posts by a radio telemetry. It is similar to the conventional system except for a new electronic circuit board and an integrated operation program. The new electronic circuit board is able to store the radiation data with a time tag of 6 or more months if the recording interval time is 10 s. The operation program automatically sends the time correction command to the six monitoring posts for a daily synchronization between the monitoring posts and the central control computer as a Korean mean time. The previous-lost radiation data for 6 or more months could be restored by using two components with the functions of a time tag and a daily synchronization without any additional equipment. It was tested for more than 1 year, from which the test results, the data collection rate was dramatically improved without any tedious manual work, which was almost about 100% for 1 year. The smart measurement system has been applied for an effective gamma monitoring around the nuclear facilities at KAERI since it was developed and tested in 2003.

  6. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    NASA Astrophysics Data System (ADS)

    LaFleur, Adrienne M.; Menlove, Howard O.

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies.

  7. Determination of the DFN modeling domain size based on ensemble variability of equivalent permeability

    NASA Astrophysics Data System (ADS)

    Ji, S. H.; Koh, Y. K.

    2015-12-01

    Conceptualization of the fracture network in a disposal site is important for the safety assessment of a subsurface repository for radioactive waste. To consider the uncertainty of the stochastically conceptualized discrete fracture networks (DFNs), the ensemble variability of equivalent permeability was evaluated by defining different network structures with various fracture densities and characterization levels, and analyzing the ensemble mean and variability of the equivalent permeability of the networks, where the characterization level was defined as the ratio of the number of deterministically conceptualized fractures to the total number of fractures in the domain. The results show that the hydraulic property of the generated fractures were similar among the ensembles when the fracture density was larger than the specific fracture density where the domain size was equal to the correlation length of a given fracture network. In a sparsely fracture network where the fracture density was smaller than the specific fracture density, the ensemble variability was too large to ensure the consistent property from the stochastic DFN modeling. Deterministic information for a portion of a fracture network could reduce the uncertainty of the hydraulic property only when the fracture density was larger than the specific fracture density. Based on these results, the DFN modeling domain size for KAERI's (Korea Atomic Energy Research Institute) URT (Underground Research Tunnel) site to guarantee a less variable hydraulic property of the fracture network was determined by calculating the correlation length, and verified by evaluating the ensemble variability of the equivalent permeability.

  8. Development progresses of radio frequency ion source for neutral beam injector in fusion devices.

    PubMed

    Chang, D H; Jeong, S H; Kim, T S; Park, M; Lee, K W; In, S R

    2014-02-01

    A large-area RF (radio frequency)-driven ion source is being developed in Germany for the heating and current drive of an ITER device. Negative hydrogen ion sources are the major components of neutral beam injection systems in future large-scale fusion experiments such as ITER and DEMO. RF ion sources for the production of positive hydrogen (deuterium) ions have been successfully developed for the neutral beam heating systems at IPP (Max-Planck-Institute for Plasma Physics) in Germany. The first long-pulse ion source has been developed successfully with a magnetic bucket plasma generator including a filament heating structure for the first NBI system of the KSTAR tokamak. There is a development plan for an RF ion source at KAERI to extract the positive ions, which can be applied for the KSTAR NBI system and to extract the negative ions for future fusion devices such as the Fusion Neutron Source and Korea-DEMO. The characteristics of RF-driven plasmas and the uniformity of the plasma parameters in the test-RF ion source were investigated initially using an electrostatic probe. PMID:24593580

  9. Korea Research Reactor -1 & 2 Decommissioning Project in Korea

    SciTech Connect

    Park, S. K.; Chung, U. S.; Jung, K. J.; Park, J. H.

    2003-02-24

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D & D) project of these two research reactors, the first D & D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 & 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000.

  10. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  11. 3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan

    SciTech Connect

    Ku, J.H.; Choung, W.M.; You, G.S.; Moon, S.I.; Park, S.H.; Kim, H.D.

    2013-07-01

    The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluation plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.

  12. CFD study of isothermal water flow in rod bundle with split-type spacer grid

    NASA Astrophysics Data System (ADS)

    Batta, A.; Class, A. G.

    2014-06-01

    The design of rod bundles in nuclear application nowadays is assessed by CFD (computational fluid dynamics). The accuracy of CFD models need validation. Within the OECD/NEA benchmark MATiS-H (Measurement and Analysis of Turbulent Mixing in Sub-channels - Horizontal) a single-phase water flow in a 5x5 rod bundle is studied. In the benchmark, two types of spacer grids are tested, the swirl type and the split type, where the current study focuses on the split type spacer grid. Comparison of CFD results obtained at Karlsruhe Institut of Technology (KIT) with experimental results of KAERI (Korea Atomic Energy Research Institute) are presented. In the benchmark velocities components along selected lines downstream of the spacer grid are measured and compared to CFD results. The CFD code STAR CCM+ with the Realized k-ɛ model is used. Comparisons with experimental results show quantitative and qualitative agreement for the averaged values of velocity components. Comparisons of results to other benchmark partners using different modeling show that the selected mesh size and models for the analysis of the current case gives relatively accurate results. However, the used turbulent model (Realized k-ɛ does not capture the turbulent intensity correctly. Computation shows that the flow has very high mixing due to the spacer grid, which does not decay within the measurements domain (z/ DH =0-10 downstream of spacer grid). The same conclusion can be drawn from experimental data.

  13. Casting evaluation of U-Zr alloy system fuel slug for SFR prepared by injection casting method

    SciTech Connect

    Song, Hoon; Kim, Jong-Hwan; Kim, Ki-Hwan; Lee, Chan-Bock

    2013-07-01

    Metal fuel slugs of U-Pu-Zr alloys for Sodium-cooled Fast Reactor (SFR) have conventionally been fabricated by a vacuum injection casting method. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long-lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. In order to prevent the evaporation of volatile elements such as Am, alternative fabrication methods of metal fuel slugs have been studied applying gravity casting, and improved injection casting in KAERI, including melting under inert atmosphere. And then, metal fuel slugs were examined with casting soundness, density, chemical analysis, particle size distribution and microstructural characteristics. Based on these results there is a high level of confidence that Am losses will also be effectively controlled by application of a modest amount of overpressure. A surrogate fuel slug was generally soundly cast by improved injection casting method, melted fuel material under inert atmosphere.

  14. Characteristics of a Frisch collar grid CdZnTe radiation detector grown by low-pressure Bridgman method

    NASA Astrophysics Data System (ADS)

    Jeong, Manhee; Jo, Woo Jin; Kim, Han Soo; Ha, Jang Ho

    2015-06-01

    A single-polarity charge-sensing method was studied by using a CdZnTe Frisch collar grid detector grown by using a low-pressure Bridgeman furnace at the Korea Atomic Energy Research Institute (KAERI). The Frisch collar grid CdZnTe detector has an active volume of 5 × 5 × 10 mm3, and was fabricated from a single crystal, Teflon tape and copper tape used as a Frisch collar grid. A room-temperature energy resolution of 6% full width at half maximum (FWHM) was obtained for the 662keV peak of Cs-137 without any additional electrical corrections. The detector's fabrication process is described, and its characteristics are discussed. Finally, the charge transport properties and gamma-ray energy resolution of the fabricated Frisch collar grid detector are compared with those of two other Frisch collar detectors grown by using different crystal growth methods and purchased from eV-products and Redlen technology.

  15. Development of a Compton camera for safeguards applications in a pyroprocessing facility

    NASA Astrophysics Data System (ADS)

    Park, Jin Hyung; Kim, Young Su; Kim, Chan Hyeong; Seo, Hee; Park, Se-Hwan; Kim, Ho-Dong

    2014-11-01

    The Compton camera has a potential to be used for localizing nuclear materials in a large pyroprocessing facility due to its unique Compton kinematics-based electronic collimation method. Our R&D group, KAERI, and Hanyang University have made an effort to develop a scintillation-detector-based large-area Compton camera for safeguards applications. In the present study, a series of Monte Carlo simulations was performed with Geant4 in order to examine the effect of the detector parameters and the feasibility of using a Compton camera to obtain an image of the nuclear material distribution. Based on the simulation study, experimental studies were performed to assess the possibility of Compton imaging in accordance with the type of the crystal. Two different types of Compton cameras were fabricated and tested with a pixelated type of LYSO (Ce) and a monolithic type of NaI(Tl). The conclusions of this study as a design rule for a large-area Compton camera can be summarized as follows: 1) The energy resolution, rather than position resolution, of the component detector was the limiting factor for the imaging resolution, 2) the Compton imaging system needs to be placed as close as possible to the source location, and 3) both pixelated and monolithic types of crystals can be utilized; however, the monolithic types, require a stochastic-method-based position-estimating algorithm for improving the position resolution.

  16. Regulatory Experiences for the Decommissioning of the Research Reactor in Korea

    SciTech Connect

    CHOI, Kyung-Woo

    2008-01-15

    The first research reactor in Korea (KRR-1, TRIGA Mark-II) has operated since 1962, and the second one (KRR-2, TRIGA Mark-III), since 1972. Both of them were phased out in 1995 due to their lives and the operation of a new research reactor, HANARO (30 MW thermal power) operated by KAERI (Korea Atomic Energy Research Institute). After deciding the shutdown by the Nuclear Development and Utilization Committee in March 1996, KAERI began to prepare the decommissioning plan, including the environmental impact assessment, and submitted the plan to the Ministry of Science and Technology (MOST) in December 1998. Korea Institute of Nuclear Safety (KINS) reviewed document and prepared the review report in 1999. KINS is an organization of technical expertise which performs regulatory functions, entrusted by the MOST in accordance with the Atomic Energy Act and its Enforcement Decree. The review report written by KINS was consulted by the Special Committee on Nuclear Safety in January 2000. The committee submitted their consultation results to the Nuclear Safety Commission for the final approval by the Minister of MOST. The license was issued in November 2000. With the consent of the Korean government to the US Record of Decision, the spent fuel of KRR-1 and 2 was safely transported to the United States in July 1998. The decontamination and dismantling of KRR-2 was completed at the end of 2005 but the decommissioning of KRR-1 has been suspended by the problem for the memorial of the reactor. After the decommissioning of the research reactor is finished, the site will be returned to the site owner, Korea Electric Power Corporation (KEPCO). In this paper, the state-of-art and lessons learnt from recent regulatory activities for decommissioning of KRR- 2 are summarized. In conclusion: since the shutdown of KRR-1 and 2 had been decided, the safe assessment and licensing review were carried out after applying for decommissioning plan of those research reactors by operator. Through

  17. Final report-passive safety optimization in liquid sodium-cooled reactors.

    SciTech Connect

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  18. Zone Freezing Study for Pyrochemical Process Waste Minimization

    SciTech Connect

    Ammon Williams

    2012-05-01

    Pyroprocessing technology is a non-aqueous separation process for treatment of used nuclear fuel. At the heart of pyroprocessing lies the electrorefiner, which electrochemically dissolves uranium from the used fuel at the anode and deposits it onto a cathode. During this operation, sodium, transuranics, and fission product chlorides accumulate in the electrolyte salt (LiCl-KCl). These contaminates change the characteristics of the salt overtime and as a result, large volumes of contaminated salt are being removed, reprocessed and stored as radioactive waste. To reduce the storage volumes and improve recycling process for cost minimization, a salt purification method called zone freezing has been proposed at Korea Atomic Energy Research Institute (KAERI). Zone freezing is melt crystallization process similar to the vertical Bridgeman method. In this process, the eutectic salt is slowly cooled axially from top to bottom. As solidification occurs, the fission products are rejected from the solid interface and forced into the liquid phase. The resulting product is a grown crystal with the bulk of the fission products near the bottom of the salt ingot, where they can be easily be sectioned and removed. Despite successful feasibility report from KAERI on this process, there were many unexplored parameters to help understanding and improving its operational routines. Thus, this becomes the main motivation of this proposed study. The majority of this work has been focused on the CsCl-LiCl-KCl ternary salt. CeCl3-LiCl-KCl was also investigated to check whether or not this process is feasible for the trivalent species—surrogate for rare-earths and transuranics. For the main part of the work, several parameters were varied, they are: (1) the retort advancement rate—1.8, 3.2, and 5.0 mm/hr, (2) the crucible lid configurations—lid versus no-lid, (3) the amount or size of mixture—50 and 400 g, (4) the composition of CsCl in the salt—1, 3, and 5 wt%, and (5) the

  19. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    NASA Astrophysics Data System (ADS)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  20. Spent fuel management status perspectives in Korea

    SciTech Connect

    Park, H.S.; Lee, J.S.; Kim, B.T. )

    1992-01-01

    Concomitant with steadily increasing nuclear power program in Korea, a national radioactive waste management program has been in initial implementation stage for several years. In late 1990, however, a serious confrontation was witnessed at Anmyon area where residents expressed strong opposition against any possibility to consider that site as a potential candidate for waste disposal by the Authority. As far as spent fuel management is concerned, an interim storage policy was adopted by Korean Atomic Energy Commission. A decision to build a centralized wet storage facility was made followed by a conceptual design. Due to the incident at Anmyon site, the public has became more concerned about radioactive wastes management. Parallel efforts are being made to ameliorate public acceptance in regard to radioactive waste management and in particular to spent fuel management. There are substantial uncertainties, however, whether any site could be found given that precarious mood has been prevailing against radioactive wastes throughout the world. In the meantime waiting for successful siting, various research and development for future perspectives are in order. Of particular importance in such endeavor is to provide technological impetus for future perspectives as well as public acceptance through safety demonstrations of certain viable technology alternatives. The dry storage option, for instance, is acclaimed for intrinsic safety and lower cost as prospective alternative. Combined with rod consolidation, dry storage technologies which have not extensively applied in the past, could be considered as a technological basis for longer term management of spent fuel. Conscious of such global trend, some appropriate programs in preparation for such perspectives have been launched by KAERI.

  1. Incorporation of a Helical Tube Heat Transfer Model in the MARS Thermal Hydraulic Systems Analysis Code for the T/H Analyses of the SMART Reactor

    SciTech Connect

    Young Jin Lee; Bub Dong Chung; Jong Chull Jo; Hho Jung Kim; Un Chul Lee

    2004-07-01

    SMART is a medium sized integral type advanced pressurized water reactor currently under development at KAERI. The steam generators of SMART are designed with helically coiled tubes and these are designed to produce superheated steam. The helical shape of the tubes can induce strong centrifugal effect on the secondary coolant as it flows inside the tubes. The presence of centrifugal effect is expected to enhance the formation of cross-sectional circulation flows within the tubes that will increase the overall heat transfer. Furthermore, the centrifugal effect is expected to enhance the moisture separation and thus make it easier to produce superheated steam. MARS is a best-estimate thermal-hydraulic systems analysis code with multi-phase, multi-dimensional analysis capability. The MARS code was produced by restructuring and merging the RELAP5 and the COBRA-TF codes. However, MARS as well as most other best-estimate systems analysis codes in current use lack the detailed models needed to describe the thermal hydraulics of helically coiled tubes. In this study, the heat transfer characteristics and relevant correlations for both the tube and shell sides of helical tubes have been investigated, and the appropriate models have been incorporated into the MARS code. The newly incorporated helical tube heat transfer package is available to the MARS users via selection of the appropriate option in the input. A performance analysis on the steam generator of SMART under full power operation was carried out using the modified MARS code. The results of the analysis indicate that there is a significant improvement in the code predictability. (authors)

  2. NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

    SciTech Connect

    OH,S.Y.; CHANG,J.; MUGHABGHAB,S.

    2000-05-11

    Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.

  3. First neutral beam injection experiments on KSTAR tokamak.

    PubMed

    Jeong, S H; Chang, D H; Kim, T S; In, S R; Lee, K W; Jin, J T; Chang, D S; Oh, B H; Bae, Y S; Kim, J S; Park, H T; Watanabe, K; Inoue, T; Kashiwagi, M; Dairaku, M; Tobari, H; Hanada, M

    2012-02-01

    The first neutral beam (NB) injection system of the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak was partially completed in 2010 with only 1∕3 of its full design capability, and NB heating experiments were carried out during the 2010 KSTAR operation campaign. The ion source is composed of a JAEA bucket plasma generator and a KAERI large multi-aperture accelerator assembly, which is designed to deliver a 1.5 MW, NB power of deuterium at 95 keV. Before the beam injection experiments, discharge, and beam extraction characteristics of the ion source were investigated. The ion source has good beam optics in a broad range of beam perveance. The optimum perveance is 1.1-1.3 μP, and the minimum beam divergence angle measured by the Doppler shift spectroscopy is 0.8°. The ion species ratio is D(+):D(2)(+):D(3)(+) = 75:20:5 at beam current density of 85 mA/cm(2). The arc efficiency is more than 1.0 A∕kW. In the 2010 KSTAR campaign, a deuterium NB power of 0.7-1.5 MW was successfully injected into the KSTAR plasma with a beam energy of 70-90 keV. L-H transitions were observed within a wide range of beam powers relative to a threshold value. The edge pedestal formation in the T(i) and T(e) profiles was verified through CES and electron cyclotron emission diagnostics. In every deuterium NB injection, a burst of D-D neutrons was recorded, and increases in the ion temperature and plasma stored energy were found. PMID:22380259

  4. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  5. I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

    SciTech Connect

    S.M. Frank

    2011-09-01

    Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project

  6. Steady-state operation of a large-area high-power RF ion source for the neutral beam injector

    NASA Astrophysics Data System (ADS)

    Chang, Doo-Hee; Park, Min; Jeong, Seung Ho; Kim, Tae-Seong; Lee, Kwang Won; In, Sang Ryul

    2014-10-01

    A large-area high-power RF-driven ion source is being developed in Germany for the heating and current drive (H&CD) of an ITER device. Negative hydrogen ion sources are the major components of neutral beam injection systems in future large-scale fusion devices such as an the ITER and the DEMO. The first and the second long-pulse ion sources (LPIS-1 and LPIS-2) have been successfully developed with a magnetic-bucket plasma generator, including a filament heating structure for the first NBI (NBI-1) system of the KSTAR tokamak. A development plan exists for a large-area high-power RF ion source for steady-state operation (more than 300 seconds) at the Korea Atomic Energy Research Institute (KAERI) to extract positive ions, which can be used for the NBI heating and current drive systems, and to extract negative ions for future fusion devices such as a Fusion Neutron Source and Korea — DEMO. The RF ion source consists of a driver region, including a helical antenna and a discharge chamber, and an expansion region (magnetic bucket of the prototype LPIS-1). RF power can be transferred at up to 10 kW with a fixed frequency of 2 MHz through an optimized RF matching system. An actively water-cooled Faraday shield is located inside the driver region of the ion source for stable and steady-state operation of the RF discharge. The uniformities of the plasma parameters are measured at the lowest area of the expansion bucket by using two RF-compensated electrostatic probes along the directions of the short and the long dimensions of the expansion region.

  7. Thermoluminescence emission spectra for the LiF:Mg,Cu,Na,Si thermoluminescent materials with various concentrations of the dopants (3-D measurement).

    PubMed

    Lee, J I; Lee, D; Kim, J L; Chang, S Y

    2006-01-01

    The thermoluminescence (TL) emission spectra from LiF TL materials, called KLT-300 (LiF:Mg,Cu,Na,Si) with various dopant concentrations are measured and analysed. These KLT-300 materials were developed by the Korea Atomic Energy Research Institute (KAERI) to achieve an enhancement of the thermal stability in TL readings. Six types of samples are prepared with different dopant concentrations in the following ranges; Mg (0-0.20 mol%), Cu (0-0.05 mol%), Na and Si (0-0.9 mol%). The spectra measurements are carried out for the six types of samples using a TL emission spectra measurement device. The spectra measurement device consists of a monochromator, photomultiplier tube and temperature control unit to thermally stimulate the samples. The measured data shows the light emission during heating of the sample as a function of temperature and wavelength (three-dimensional TL spectra). The spectra were analysed using a method of deconvolution based on gaussian curve. The wavelength of a main peak of the emission spectra changes depending on the existence of the Cu dopant, while intensity of the spectra rapidly changes with the Cu dopant concentrations. The 385 nm emission is mainly observed in all the spectra from the samples with the Cu dopant, but in those from the samples without the Cu dopant a very weak 401 nm emission is mainly observed. However, any change in the wavelength at a main peak of the TL emission spectra from the sample materials with Na and Si dopants is not observed but that in the intensity at a peak of the spectra is observed. PMID:16644972

  8. Analysis of Phenix end-of-life natural convection test with the MARS-LMR code

    SciTech Connect

    Jeong, H. Y.; Ha, K. S.; Lee, K. L.; Chang, W. P.; Kim, Y. I.

    2012-07-01

    The end-of-life test of Phenix reactor performed by the CEA provided an opportunity to have reliable and valuable test data for the validation and verification of a SFR system analysis code. KAERI joined this international program for the analysis of Phenix end-of-life natural circulation test coordinated by the IAEA from 2008. The main objectives of this study were to evaluate the capability of existing SFR system analysis code MARS-LMR and to identify any limitation of the code. The analysis was performed in three stages: pre-test analysis, blind posttest analysis, and final post-test analysis. In the pre-test analysis, the design conditions provided by the CEA were used to obtain a prediction of the test. The blind post-test analysis was based on the test conditions measured during the tests but the test results were not provided from the CEA. The final post-test analysis was performed to predict the test results as accurate as possible by improving the previous modeling of the test. Based on the pre-test analysis and blind test analysis, the modeling for heat structures in the hot pool and cold pool, steel structures in the core, heat loss from roof and vessel, and the flow path at core outlet were reinforced in the final analysis. The results of the final post-test analysis could be characterized into three different phases. In the early phase, the MARS-LMR simulated the heat-up process correctly due to the enhanced heat structure modeling. In the mid phase before the opening of SG casing, the code reproduced the decrease of core outlet temperature successfully. Finally, in the later phase the increase of heat removal by the opening of the SG opening was well predicted with the MARS-LMR code. (authors)

  9. A Study on Cost Allocation in Nuclear Power Coupled with Desalination

    SciTech Connect

    Lee, ManKi; Kim, SeungSu; Moon, KeeHwan; Lim, ChaeYoung

    2004-07-01

    As for a single-purpose desalination plant, there is no particular difficulty in computing the unit cost of the water, which is obtained by dividing the annual total costs by the output of fresh water. When it comes to a dual-purpose plant, cost allocation is needed between the two products. No cost allocation is needed in some cases where two alternatives producing the same water and electricity output are to be compared. In these cases, the consideration of the total cost is then sufficient. This study assumes MED (Multi-Effect Distillation) technology is adopted when nuclear power is coupled with desalination. The total production cost of the two commodities in dual-purpose plant can easily be obtained by using costing methods, if the necessary raw data are available. However, it is not easy to calculate a separate cost for each product, because high-pressure steam plant costs cannot be allocated to one or the other without adopting arbitrary methods. Investigation on power credit method is carried out focusing on the cost allocation of combined benefits due to dual production, electricity and water. The illustrative calculation is taken from Preliminary Economic Feasibility Study of Nuclear Desalination in Madura Island, Indonesia. The study is being performed by BATAN (National Nuclear Energy Agency), KAERI (Korean Atomic Energy Research Institute) and under support of the IAEA (International Atomic Energy Agency) started in the year 2002 in order to perform a preliminary economic feasibility in providing the Madurese with sufficient power and potable water for the public and to support industrialization and tourism in Madura Region. The SMART reactor coupled with MED is considered to be an option to produce electricity and potable water. This study indicates that the correct recognition of combined benefits attributable to dual production is important in carrying out economics of desalination coupled with nuclear power. (authors)

  10. The Necessity of a New Type Test Rig for the Development of an Evaluation Method in Grid Fretting Problems

    SciTech Connect

    Lee, Young-Ho; Kim, Hyung-Kyu

    2007-07-01

    A grid fretting problem is recognized as one of the most important degradation mechanisms even though the examination results of fretting experiments could be applied to the development and design of spacer grid structures. This is because it is difficult to develop a fretting wear model for a grid fretting problem due to the various wear mechanisms involved according to the mechanical and environmental variables, the contact condition with a spring/dimple and the material properties. A number of spring shapes has been developed in KAERI and their performance tests such as fretting wear, flow-induced vibration (FIV) tests, etc. have been carried out from a part unit to a full assembly scale. From the unit part fretting test results, one of the noticeable results is that the contacting force (normal load) was gradually decreased with increasing number of fretting cycles due to a depth increase and this behavior was closely related to the contacting spring shape. When considering the actual contact condition between a fuel rod and a spring/dimple, if a fretting wear progresses due to a FIV under a specific normal load exerted on the fuel rod by an elastic deformation of the spring, the contacting force between the fuel rod and dimple that are located in the opposite side should be decreased. Consequently, an evaluation of developed spacer grids against fretting wear damage should be performed with the results of 1x1 cell unit experiments because a contacting force is one of the most important variables that influences a fretting wear mechanism. The discussion was focused on the development procedure of a new test rig and its performance by using a 1x1 cell unit test rig. (authors)

  11. STAR: The Secure Transportable Autonomous Reactor System - Encapsulated Fission Heat Source

    SciTech Connect

    Ehud Greenspan

    2003-10-31

    OAK-B135 The Encapsulated Nuclear Heat Source (ENHS) is a novel 125 MWth fast spectrum reactor concept that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor. It uses Pb-Bi or other liquid-metal coolant and is intended to be factory manufactured in large numbers to be economically competitive. It is anticipated to be most useful to developing countries. The US team studying the feasibility of the ENHS reactor concept consisted of the University of California, Berkeley, Argonne National Laboratory (ANL), Lawrence Livermore National Laboratory (LLNL) and Westinghouse. Collaborating with the US team were three Korean organizations: Korean Atomic Energy Research Institute (KAERI), Korean Advanced Institute for Science and Technology (KAIST) and the University of Seoul, as well as the Central Research Institute of the Electrical Power Industry (CRIEPI) of Japan. Unique features of the ENHS include at least 20 years of operation without refueling; no fuel handling in the host country; no pumps and valves; excess reactivity does not exceed 1$; fully passive removal of the decay heat; very small probability of core damaging accidents; autonomous operation and capability of load-following over a wide range; very long plant life. In addition it offers a close match between demand and supply, large tolerance to human errors, is likely to get public acceptance via demonstration of superb safety, lack of need for offsite response, and very good proliferation resistance. The ENHS reactor is designed to meet the requirements of Generation IV reactors including sustainable energy supply, low waste, high level of proliferation resistance, high level of safety and reliability, acceptable risk to capital and, hopefully, also competitive busbar cost of electricity.

  12. Forced and mixed convection heat transfer to supercritical CO{sub 2} vertically flowing in a uniformly-heated circular tube

    SciTech Connect

    Bae, Yoon-Yeong; Kim, Hwan-Yeol; Kang, Deog-Ji

    2010-11-15

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

  13. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  14. An Experimental Study of Critical Heat Flux in Narrow Gap With Two-Dimensional Slices

    SciTech Connect

    Yong Hoon Kim; Suh, Kune Y.; Rae Joon Park; San Baek Kim; Hee Dong Kim

    2002-07-01

    A cooling mechanism due to boiling in a gap between the debris crust and the reactor pressure vessel (RPV) wall was proposed for the TMI-2 reactor accident analysis. If there is enough heat transfer through the gap to cool the outer surface of the debris and the inner surface of the wall, the RPV wall may preserve its integrity during a severe core melt accident. If the heat removal through gap cooling relative to the counter-current flow limitation (CCFL) is pronounced, the safety margin of the reactor can be far greater than what had been previously known in the severe accident management arena. Should a severe accident take place, the RPV integrity will be maintained because of the inherent nature of degraded core coolability inside the lower head due to boiling in a narrow gap between the debris crust and the RPV wall. As a defense-in-depth measure, the heat removal capability by gap cooling coupled with external cooling can be examined for the Korean Standard Nuclear Power Plant (KSNPP) and the Advanced Power Reactor 1400 MWe (APR1400) in light of the TMI-2 vessel survival. A number of studies were carried out to investigate the complex heat transfer mechanisms for the debris cooling in the lower plenum. However, these heat transfer mechanisms have not been clearly understood yet. The CHFG (Critical Heat Flux in Gap) experiments at KAERI were carried out to develop the critical heat flux (CHF) correlation in a hemispherical gap, which is the upper limit of the heat transfer. According to the CHFG experiments performed with a pool boiling condition, the CHF in a parallel gap was reduced by 1/30 compared with the value measured in the open pool boiling condition. The correlation developed from the CHFG experiment is based on the fact that the CHF in a hemispherical gap is governed by the CCFL and a Kutateladze type CCFL parameter correlates CCFL data well in hemispherical gap geometry. However, the results of the CHFG experiments appear to be limited in their

  15. Hydrogeological Characteristics of Fractured Rocks around the In-DEBS Test Borehole at the Underground Research Facility (KURT)

    NASA Astrophysics Data System (ADS)

    Ko, Nak-Youl; Kim, Geon Young; Kim, Kyung-Su

    2016-04-01

    In the concept of the deep geological disposal of radioactive wastes, canisters including high-level wastes are surrounded by engineered barrier, mainly composed of bentonite, and emplaced in disposal holes drilled in deep intact rocks. The heat from the high-level radioactive wastes and groundwater inflow can influence on the robustness of the canister and engineered barrier, and will be possible to fail the canister. Therefore, thermal-hydrological-mechanical (T-H-M) modeling for the condition of the disposal holes is necessary to secure the safety of the deep geological disposal. In order to understand the T-H-M coupling phenomena at the subsurface field condition, "In-DEBS (In-Situ Demonstration of Engineered Barrier System)" has been designed and implemented in the underground research facility, KURT (KAERI Underground Research Tunnel) in Korea. For selecting a suitable position of In-DEBS test and obtaining hydrological data to be used in T-H-M modeling as well as groundwater flow simulation around the test site, the fractured rock aquifer including the research modules of KURT was investigated through the in-situ tests at six boreholes. From the measured data and results of hydraulic tests, the range of hydraulic conductivity of each interval in the boreholes is about 10‑7-10‑8 m/s and that of influx is about 10‑4-10‑1 L/min for NX boreholes, which is expected to be equal to about 0.1-40 L/min for the In-DEBS test borehole (diameter of 860 mm). The test position was determined by the data and availability of some equipment for installing In-DEBS in the test borehole. The mapping for the wall of test borehole and the measurements of groundwater influx at the leaking locations was carried out. These hydrological data in the test site will be used as input of the T-H-M modeling for simulating In-DEBS test.

  16. Computational Fluid Dynamic simulations of pipe elbow flow.

    SciTech Connect

    Homicz, Gregory Francis

    2004-08-01

    One problem facing today's nuclear power industry is flow-accelerated corrosion and erosion in pipe elbows. The Korean Atomic Energy Research Institute (KAERI) is performing experiments in their Flow-Accelerated Corrosion (FAC) test loop to better characterize these phenomena, and develop advanced sensor technologies for the condition monitoring of critical elbows on a continuous basis. In parallel with these experiments, Sandia National Laboratories is performing Computational Fluid Dynamic (CFD) simulations of the flow in one elbow of the FAC test loop. The simulations are being performed using the FLUENT commercial software developed and marketed by Fluent, Inc. The model geometry and mesh were created using the GAMBIT software, also from Fluent, Inc. This report documents the results of the simulations that have been made to date; baseline results employing the RNG k-e turbulence model are presented. The predicted value for the diametrical pressure coefficient is in reasonably good agreement with published correlations. Plots of the velocities, pressure field, wall shear stress, and turbulent kinetic energy adjacent to the wall are shown within the elbow section. Somewhat to our surprise, these indicate that the maximum values of both wall shear stress and turbulent kinetic energy occur near the elbow entrance, on the inner radius of the bend. Additional simulations were performed for the same conditions, but with the RNG k-e model replaced by either the standard k-{var_epsilon}, or the realizable k-{var_epsilon} turbulence model. The predictions using the standard k-{var_epsilon} model are quite similar to those obtained in the baseline simulation. However, with the realizable k-{var_epsilon} model, more significant differences are evident. The maximums in both wall shear stress and turbulent kinetic energy now appear on the outer radius, near the elbow exit, and are {approx}11% and 14% greater, respectively, than those predicted in the baseline calculation

  17. Modeling of flow and heat transfer for fluids at supercritical conditions

    NASA Astrophysics Data System (ADS)

    Gallaway, Tara

    2011-12-01

    The Supercritical Water Reactor (SCWR) has been proposed as one of the six Generation IV reactor design concepts under consideration. The key feature of the SCWR is that water at supercritical pressures is used as the reactor coolant. At supercritical pressures, the working fluid does not undergo phase change as it is heated, but rather the fluid properties experience dramatic variations throughout what is known as the pseudo-critical region. Highly nonuniform temperature and uid property distributions are expected in the reactor core, which will have a significant impact on turbulence and heat transfer as well as stability limits for future SCWRs. The goal of this work is to understand and predict the effects of these fluid property variations on turbulence and heat transfer throughout the reactor core and to predict the potential onset of dynamic instabilities. CO2 at supercritical conditions is included in the current study due in some part to its use as a viable simulant fluid in place of water for experimental studies. The use of CO2 at supercritical conditions as a reactor coolant has also gained popularity in recent years. Spline-type property models have been developed for both water and CO2 at supercritical pressures in order to include the property variations into a numerical solver. Turbulence and heat transfer models for fluids at supercritical conditions have been developed and implemented into the NPHASE-CMFD computer code. The results of predictions using the proposed models have been compared to experimental data from the Korea Atomic Energy Research Institute (KAERI) for various heat transfer regimes. While no model is without some deficiency, the Chien Low-Reynolds k -- epsilon model performs best at predicting the experimental data. A stability model has been developed and is presented in this dissertation as well. This model utilizes three different solution methods and tests the effects of inlet temperature, mass flow rate, local loss

  18. Melting of the metallic wastes generated by dismantling retired nuclear research facilities

    SciTech Connect

    Chong-Hun Jung; Pyung-Seob Song; Byung-Youn Min; Wang-Kyu Choi

    2008-01-15

    The decommissioning of nuclear installations results in considerably large amounts of radioactive metallic wastes such as stainless steel, carbon steel, aluminum, copper etc. It is known that the reference 1,000 MWe PWR and 881 MWe PHWR will generate metal wastes of 24,800 ton and 26,500 ton, respectively. In Korea, the D and D of KRR-2 and a UCP at KAERI have been performed. The amount of metallic wastes from the KRR-1 and UCP was about 160 ton and 45 ton, respectively, up to now. These radioactive metallic wastes will induce problems of handling and storing these materials from environmental and economical aspects. For this reason, prompt countermeasures should be taken to deal with the metal wastes generated by dismantling retired nuclear facilities. The most interesting materials among the radioactive metal wastes are stainless steel (SUS), carbon steel (CS) and aluminum wastes because they are the largest portions of the metallic wastes generated by dismantling retired nuclear research facilities. As most of these steels are slightly contaminated, if they are properly treated they are able to be recycled and reused in the nuclear field. In general, the technology of a metal melting is regarded as one of the most effective methods to treat metallic wastes from nuclear facilities. In conclusion: The melting of metal wastes (Al, SUS, carbon steel) from a decommissioning of research reactor facilities was carried out with the use of a radioisotope such as cobalt and cesium in an electric arc furnace. In the aluminum melting tests, the cobalt was captured at up to 75% into the slag phase. Most of the cesium was completely eliminated from the aluminum ingot phase and moved into the slag and dust phases. In the melting of the stainless steel wastes, the {sup 60}Co could almost be retained uniformly in the ingot phase. However, we found that significant amounts of {sup 60}Co remained in the slag at up to 15%. However the removal of the cobalt from the ingot phase was

  19. I-NERI Annual Technical Progress Report 2007-004-K Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing

    SciTech Connect

    S. Frank

    2010-09-01

    The current method for the immobilization of fission products that accumulate in electrorefiner salt during the electrochemical processing of used metallic nuclear fuel is to encapsulate the electrorefiner salt in a glass-bonded sodalite ceramic waste form. This process was developed by Argonne National Laboratory in the USA and is currently performed at the Idaho National Laboratory for the treatment of Experimental Breeder Reactor-II (EBR-II) used fuel. This process utilizes a “once-through” option for the disposal of spent electrorefiner salt; where, after the treatment of the EBR-II fuel, the electrorefiner salt containing the active fission products will be disposed of in the ceramic waste form (CWF). The CWF produced will have low fission product loading of approximately 2 to 5 weight percent due to the limited fuel inventory currently being processed. However; the design and implementation of advanced electrochemical processing facilities to treat used fuel would process much greater quantities fuel. With an advanced processing facility, it would be necessary to selectively remove fission products from the electrorefiner salt for salt recycle and to concentrate the fission products to reduce the volume of high-level waste from the treatment facility. The Korean Atomic Energy Research Institute and the Idaho National Laboratory have been collaborating on I-NERI research projects for a number of years to investigate both aspects of selective fission product separation from electrorefiner salt, and to develop advanced waste forms for the immobilization of the collected fission products. The first joint KAERI/INL I-NERI project titled: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels, was successfully completed in 2009 by concentrating and isolating fission products from actual electrorefiner salt used for the treated used EBR-II fuel. Two separation methods were tested and from these tests were

  20. Charge-dependent conformations and dynamics of pamam dendrimers revealed by neutron scattering and molecular dynamics

    NASA Astrophysics Data System (ADS)

    Wu, Bin

    spatial instrumental scales, understanding experimental results involves extensive and difficult data analysis based on liquid theory and condensed matter physics. Therefore, a model that successfully describes the inter- and intra-dendrimer correlations is crucial in obtaining and delivering reliable information. On the other hand, making meaningful comparisons between molecular dynamics and neutron scattering is a fundamental challenge to link simulations and experiments at the nano-scale. This challenge stems from our approach to utilize MD simulation to explain the underlying mechanism of experimental observation. The SANS measurements were conducted on a series of SANS spectrometers including the Extended Q-Range Small-Angle Neutron Scattering Diffractometer (EQ-SANS) and the General-Purpose Small-Angle Neutron Scattering Diffractometer (GP-SANS) at the Oak Ridge National Laboratory (ORNL), and NG7 Small Angle Neutron Scattering Spectrometer at National Institute of Standards (NIST) and Technology in U.S.A., large dynamic range small-angle diffractometer D22 at Institut Laue-Langevin (ILL) in France, and 40m-SANS Spectrometer at Korea Atomic Energy Research Institute (KAERI) in Korea. On the other hand, the Amber molecular dynamics simulation package is utilized to carry out the computational study. In this dissertation, the following observations have been revealed. The previously developed theoretical model for polyelectrolyte dendrimers are adopted to analyze SANS measurements and superb model fitting quality is found. Coupling with advanced contrast variation small angle neutron scattering (CVSANS) data analysis scheme reported recently, the intra-dendrimer hydration and hydrocarbon components distributions are revealed experimentally. The results indeed indicate that the maximum density is located in the molecular center rather than periphery, which is consistent to previous SANS studies and the back-folding picture of PAMAM dendrimers. According to this picture

  1. BIOPROTA: an international forum for environmental modelling in support of long-term radioactive waste management

    SciTech Connect

    Smith, K.L.; Smith, G.; Laciok, A.

    2007-07-01

    An international Forum, BIOPROTA, has been set up and maintained which allows common long-term environmental radiological assessment problems, such as post-closure modelling studies to be identified and then addressed. The focus of the Forum is to address key uncertainties in environmental modelling and related dose assessment with special reference to evaluation of the long-term impact of contaminant releases associated with radioactive waste management. The application of shared resources results in effective resource management and the development of common solutions to common problems. The Forum began in 2002 and has benefited from the knowledge and experience of organisations from Belgium (SCK.CEN), Czech Republic (NRI), Canada (OPG), Finland (Posiva), France (ANDRA, EdF), Japan (NUMO), Korea (KAERI), Norway (NRPA), Spain (ENRESA, CIEMAT), Sweden (SKB, SSI), Switzerland (Nagra), UK (Nirex, Nexia, UKAEA) and the USA (EPRI). These organisations include a mixture of operators, regulators and research institutes, and hence, including the participation of their technical support organizations, constitutes a very broad-based Forum. Enviros has acted as the technical secretariat to the Forum since its formation. Initially the Forum focused on three themes aimed at advancing knowledge and improving model predictions relating to performance and safety assessments: Theme 1 Development of a database to meet the key biosphere assessment information deficiencies. Theme 2 Implementation of a series of tasks to address key modelling issues, including uncertainties and inconsistencies in the modelling of inhalation, irrigation and soil contamination dose pathways; and approaches to the modelling the transfer of radionuclides across the geosphere-biosphere interface zone (GBIZ). Theme 3 Provision of guidance on site characterisation and experimental and monitoring protocols relevant to improving confidence in the biosphere component of the overall performance assessment

  2. The feasibility study of hot cell decontamination by the PFC spray method

    SciTech Connect

    Hui-Jun Won; Chong-Hun Jung; Jei-Kwon Moon

    2008-01-15

    The characteristics of per-fluorocarbon compounds (PFC) are colorless, non-toxic, easily vaporized and nonflammable. Also, some of them are liquids of a high density, low surface tension, low latent heat and low specific heat. These particular chemical and physical properties of fluoro-organic compounds permit their use in very different fields such as electronics, medicine, tribology, nuclear and material science. The Sonatol process was developed under a contract with the DOE. The Sonatol process uses an ultrasonic agitation in a PFC solution that contains a fluorinated surfactant to remove radioactive particles from surfaces. Filtering the suspended particles allows the solutions to be reused indefinitely. They applied the Sonatol process to the decontamination of a heterogeneous legacy Pu-238 waste that exhibited an excessive hydrogen gas generation, which prevents a transportation of such a waste to a Waste Isolation Pilot Plant. Korea Atomic Energy Research Institute (KAERI) is developing dry decontamination technologies applicable to a decontamination of a highly radioactive area loosely contaminated with radioactive particles. This contamination has occurred as a result of an examination of a post-irradiated material or the development of the DUPIC process. The dry decontamination technologies developed are the carbon dioxide pellet spray method and the PFC spray method. As a part of the project, PFC ultrasonic decontamination technology was developed in 2004. The PFC spray decontamination method which is based on the test results of the PFC ultrasonic method has been under development since 2005. The developed PFC spray decontamination equipment consists of four modules (spray, collection, filtration and distillation). Vacuum cup of the collection module gathers the contaminated PFC solution, then the solution is moved to the filtration module and it is recycled. After a multiple recycling of the spent PFC solution, it is purified in the distillation

  3. FY2012 summary of tasks completed on PROTEUS-thermal work.

    SciTech Connect

    Lee, C.H.; Smith, M.A.

    2012-06-06

    resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes

  4. Preface

    NASA Astrophysics Data System (ADS)

    Gorse, D.; Boutard, J.-L.

    2002-09-01

    interest for the next generation of LM spallation targets in EU, U.S.A. and Japan. These proceedings contain manuscripts from 90% of the presented papers. The organizers would like to thank all their Colleagues who presented papers, contributed with manuscripts and attended the sessions at the symposium. For sake of clarity, this volume is divided into five sections: 1) general R& D for spallation targets, 2) irradiation effects in liquid metal spallation targets, 3) oxygen control: thermodynamics and monitoring, 4) resistance to liquid metal corrosion and embrittlement of structural materials for spallation targets and 5) basic studies of intergranular penetration and liquid metal embrittlement. Section 1 begins with a description of the spallation neutron source facility SINQ and of ongoing R& D programs at PSI (Switzerland), including MEGAPIE, the joint initiative by six European research institutions and JAERI (Japan), DOE (USA) and KAERI (Korea) to design, build, operate and assess the performance of a liquid lead-bismuth spallation target for 1MW of beam power (G. Bauer et al.). The materials aspects related to the MEGAPIE target and to the LiSoR (Liquid Solid Reactions under irradiation) experiment are reviewed by T. Auger et al. The advantages and drawbacks of solid tungsten spallation targets, compared to liquid Pb-Bi eutectic spallation targets are examined by R. Enderlé et al., presenting the CEA point of view. Section 2 is dedicated to irradiation effects in Liquid Metal (LM) spallation targets structure, a crucial problem for the feasibility of ADS. P. Jung is pointing out the specificity of the irradiation conditions in LM targets by comparison with fast neutron fission and fusion reactors, and the metallurgical consequences like irradiation and helium-induced embrittlement. The author emphasizes the importance of spallation residues whose deleterious effects on in-service properties of target container and window are largely unknown. Until recently, say

  5. SCC analysis of Alloy 600 tubes from a retired steam generator

    NASA Astrophysics Data System (ADS)

    Hwang, Seong Sik; Kim, Hong Pyo

    2013-09-01

    Steam generators (SG) equipped with Alloy 600 tubes of a Korean nuclear power plants were replaced with a new one having Alloy 690 tubes in 1998 after 20 years of operation. To set up a guide line for an examination of the other SG tubes, a metallographic examination of the defected tubes was carried out. A destructive analysis on 71 tubes was addressed, and a relation among the stress corrosion crack (SCC) defect location, defect depth, and location of the sludge pile was obtained. Tubes extracted from the retired SG were transferred to a hot laboratory. Detailed nondestructive analysis examinations were taken again at the laboratory, and the tubes were then destructively examined. The types and sizes of the cracks were characterized. The location and depth of the SCC were evaluated in terms of the location and height of the sludge. Most axial cracks were in the sludge pile, whereas the circumferential ones were around the top of the tube sheet (TTS) or below the TTS. Average defect depth of the axial cracks was deeper than that of the circumferential ones. Axial cracks at tube support plate (TSP) seem to be related with corrosion/sludge in crevice like at the TTS region. Circumferential cracks at TSP seem to be caused by tube denting at the upper part of the TSP. Tubes not having clear ECT signals for quantifying an ECT data-base. Tubes having no ECT signal. Tubes with a large ECT signal. Tubes with various types and sizes of flaws (primary water stress corrosion cracking (PWSCC), outside diameter stress corrosion cracking (ODSCC), Pit). Tubes with distinct PWSCC or ODSCC. Tubes were extracted from the RSG based on the field ECT with the criteria, and transferred to a hot laboratory at the Korea Atomic Energy Research Institute (KAERI) for destructive examination. A comprehensive ECT inspection was performed again at the hot laboratory to confirm the location of the cracks obtained from a field inspection. These exact locations of the defects were marked on the

  6. Preface

    NASA Astrophysics Data System (ADS)

    Gorse, D.; Boutard, J.-L.

    2002-09-01

    interest for the next generation of LM spallation targets in EU, U.S.A. and Japan. These proceedings contain manuscripts from 90% of the presented papers. The organizers would like to thank all their Colleagues who presented papers, contributed with manuscripts and attended the sessions at the symposium. For sake of clarity, this volume is divided into five sections: 1) general R& D for spallation targets, 2) irradiation effects in liquid metal spallation targets, 3) oxygen control: thermodynamics and monitoring, 4) resistance to liquid metal corrosion and embrittlement of structural materials for spallation targets and 5) basic studies of intergranular penetration and liquid metal embrittlement. Section 1 begins with a description of the spallation neutron source facility SINQ and of ongoing R& D programs at PSI (Switzerland), including MEGAPIE, the joint initiative by six European research institutions and JAERI (Japan), DOE (USA) and KAERI (Korea) to design, build, operate and assess the performance of a liquid lead-bismuth spallation target for 1MW of beam power (G. Bauer et al.). The materials aspects related to the MEGAPIE target and to the LiSoR (Liquid Solid Reactions under irradiation) experiment are reviewed by T. Auger et al. The advantages and drawbacks of solid tungsten spallation targets, compared to liquid Pb-Bi eutectic spallation targets are examined by R. Enderlé et al., presenting the CEA point of view. Section 2 is dedicated to irradiation effects in Liquid Metal (LM) spallation targets structure, a crucial problem for the feasibility of ADS. P. Jung is pointing out the specificity of the irradiation conditions in LM targets by comparison with fast neutron fission and fusion reactors, and the metallurgical consequences like irradiation and helium-induced embrittlement. The author emphasizes the importance of spallation residues whose deleterious effects on in-service properties of target container and window are largely unknown. Until recently, say