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.


    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. Beam characterization at the KAERI UED beamline

    NASA Astrophysics Data System (ADS)

    Setiniyaz, Sadiq; Kim, Hyun Woo; Baek, In-Hyung; Nam, Jinhee; Chae, MoonSik; Han, Byung-Heon; Gudkov, Boris; Jang, Kyu Ha; Park, Sunjeong; Jeong, Young Uk; Miginsky, Sergey; Vinokurov, Nikolay


    The UED (ultrafast electron diffraction) beamline of the KAERI's (the Korea Atomic Energy Research Institute's) WCI (World Class Institute) Center has been successfully commissioned. We have measured the beam emittance by using the quadrupole scan technique and the charge by using a novel measurement system we have developed. In the quadrupole scan, a larger drift distance between the quadrupole and the screen is preferred because it gives a better thin-lens approximation. A high bunch-charge beam, however, will undergo emittance growth in the long drift caused by the space-charge force. We present a method that mitigates this growth by introducing a quadrupole scan with a short drift and without using the thin-lens approximation. The quadrupole in this method is treated as a thick lens, and the emittance is extracted by using the thick-lens equations. Apart from being precise, our method can be readily applied without making any change to the beamline and has no need for a big drift space. For charge measurement, we have developed a system consisting of an in-air Faraday cup (FC) and a preamplifier. Tests performed utilizing 3.3-MeV electrons show that the system was able to measure bunches with pulse durations of tens of femtoseconds at 10 fC sensitivity.

  3. Spectroscopic measurements and AMO data center in KAERI

    NASA Astrophysics Data System (ADS)

    Rhee, Yongjoo; Kim, S. K.; Park, H. M.; Lee, Jongmin


    An introduction to the AMO (Atomic, Molecular, and Optical) database system in KAERI is given with some experimental aspects related to the AMO data production. Data sources, constructing concepts and current status of the database are described. An example of measured data which have been compiled internally is given. .

  4. Arc plasma simulation of the KAERI large ion source.


    In, S R; Jeong, S H; Kim, T S


    The KAERI large ion source, developed for the KSTAR NBI system, recently produced ion beams of 100 keV, 50 A levels in the first half campaign of 2007. These results seem to be the best performance of the present ion source at a maximum available input power of 145 kW. A slight improvement in the ion source is certainly necessary to attain the final goal of an 8 MW ion beam. Firstly, the experimental results were analyzed to differentiate the cause and effect for the insufficient beam currents. Secondly, a zero dimensional simulation was carried out on the ion source plasma to identify which factors control the arc plasma and to find out what improvements can be expected.

  5. Construction of the 1 kJ Nd: glass laser facility at KAERI

    NASA Astrophysics Data System (ADS)

    Lim, C.; Hong, S.-K.; Ko, K.; Jin, J.-T.; Kim, M.; Yun, D.-H.; Li, L.-J.; Lee, D.-W.; Lee, K.-T.; Kim, C.-J.


    We report on the design and present status of a 1 kJ Nd:Glass laser facility for basic research on quantum engineering at KAERI (Korea Atomic Energy Research Institute). By applying a newly designed spatial filter with a serrated aperture, we improved the diffracted Gaussian spatial profile of an oscillator into a flat-top one. The laser system consists of 4 beam lines, each with the energies of more than 200 J at the nano-second regime. We measured the gain and spatial profiles of each amplification stage. A spectral shaping by a two-stage OPCPA (Optical Parametric Chirped Amplifier) for a pico-second front end was studied to compensate for gain narrowing in multi-stage amplifier chains.

  6. 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


    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

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

    SciTech Connect

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


    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.

  8. 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.


    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.

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

    SciTech Connect

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


    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.

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

    SciTech Connect

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


    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.

  11. 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.


    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.

  12. 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


    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.

  13. 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


    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)

  14. 78 FR 72072 - Proposed Subsequent Arrangement

    Federal Register 2010, 2011, 2012, 2013, 2014


    ... Concerning Civil Uses of Atomic Energy and the Agreement for Cooperation in the Peaceful Uses of Nuclear.... KAERI originally obtained the material from the U.S. Department of Energy/National Nuclear Security... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF...

  15. 77 FR 34367 - Proposed Subsequent Arrangement

    Federal Register 2010, 2011, 2012, 2013, 2014


    ... Cooperation may be effectively applied for the alteration in form or content of U.S.-origin nuclear material..., ``DUPIC Fuel Fabrication Using Spent PWR Fuel at KAERI,'' dated February 2012. These facilities are found... defense and security. Dated: May 28, 2012. For the Department of Energy. Anne M. Harrington,...

  16. Seismic Fragility Analysis of a Condensate Storage Tank with Age-Related Degradations

    SciTech Connect

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


    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. This report describes the research effort performed by BNL for the Year 4 scope of work. This report was developed as an update to the Year 3 report by incorporating a major supplement to the Year 3 fragility analysis. In the Year 4 research scope, an additional study was carried out to consider an additional degradation scenario, in which the three basic degradation scenarios, i.e., degraded tank shell, degraded anchor bolts, and cracked anchorage concrete, are combined in a non-perfect correlation manner. A representative operational water level is used for this effort. Building on the same CDFM procedure implemented for the Year 3 Tasks, a simulation method was applied using optimum Latin Hypercube samples to characterize the deterioration behavior of the fragility capacity as a function of age-related degradations. The results are summarized in Section 5

  17. Nuclear Proliferation: A Historical Overview

    DTIC Science & Technology


    Pakistan’s Nuclear Weapons Program: Turning Points and Nuclear Choices, International Security, Vol. 24, No. 3 (Spring 1999), p. 183. 1971: The CANDU ... and CANDU fuel technology. Key timeline data: 1959: Office of Atomic Energy was established. See, “Country Profile: South Korea...Research Institute (KAERI) has developed both pressurized water reactor (PWR) and CANDU fuel technology. South Korea is has ratified the CTBT and the

  18. 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


    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)

  19. Resonance Region Nuclear Data Analysis to Support Advanced Fuel Cycle Development

    SciTech Connect

    Dunn, Michael E; Derrien, Herve; Leal, Luiz C; Gil, Choong-Sup; Kim, D.


    The Oak Ridge National Laboratory (ORNL) and the Korean Atomic Energy Research Institute (KAERI) are performing collaborative research as part of a three-year United States (U.S.) / Republic of Korea (ROK) International Nuclear Energy Research Initiative (I-NERI) project to provide improved neutron cross-section data with uncertainty or covariance data important for advanced fuel cycle and nuclear safeguards applications. ORNL and KAERI have initiated efforts to prepare new cross-section evaluations for 240Pu, 237Np, and the stable Cm isotopes. At the current stage of the I-NERI project, ORNL has recently completed a preliminary resonance-region cross-section evaluation with covariance data for 240Pu and initiated resonance evaluation efforts for 237Np and 244Cm. Likewise, KAERI is performing corresponding high-energy cross-section analyses (i.e., above the resonance region) for the noted isotopes. The paper provides results pertaining to the new resonance region evaluation efforts with emphasis on the new 240Pu evaluation.

  20. Development of cold neutron depth profiling system at HANARO

    NASA Astrophysics Data System (ADS)

    Park, B. G.; Sun, G. M.; Choi, H. D.


    A neutron depth profiling (NDP) system has been designed and developed at HANARO, a 30 MW research reactor at the Korea Atomic Energy Research Institute (KAERI). The KAERI-NDP system utilizes cold neutrons that are transported along the CG1 neutron guide from the cold neutron source and it consists of a neutron beam collimator, a target chamber, a beam stopper, and charged particle detectors along with NIM-standard modules for charged particle pulse-height analysis. A 60 cm in diameter stainless steel target chamber was designed to control the positions of the sample and detector. The energy distribution of the cold neutron beam at the end of the neutron guide was calculated by using the Monte Carlo simulation code McStas, and a neutron flux of 1.8×108 n/cm2 s was determined by using the gold foil activation method at the sample position. The performance of the charged particle detection of the KAERI-NDP system was tested by using Standard Reference Materials. The energy loss spectra of alpha particles and Li ions emitted from 10B, which was irradiated by cold neutrons, were measured. The measured peak concentration and the areal density of 10B in the Standard Reference Material are consistent with the reference values within 1% and 3.4%, respectively.

  1. 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.


    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

  2. Preliminary calibration of the ACP safeguards neutron counter

    NASA Astrophysics Data System (ADS)

    Lee, T. H.; Kim, H. D.; Yoon, J. S.; Lee, S. Y.; Swinhoe, M.; Menlove, H. O.


    The Advanced Spent Fuel Conditioning Process (ACP), a kind of pyroprocess, has been developed at the Korea Atomic Energy Research Institute (KAERI). Since there is no IAEA safeguards criteria for this process, KAERI has developed a neutron coincidence counter to make it possible to perform a material control and accounting (MC&A) for its ACP materials for the purpose of a transparency in the peaceful uses of nuclear materials at KAERI. The test results of the ACP Safeguards Neutron Counter (ASNC) show a satisfactory performance for the Doubles count measurement with a low measurement error for its cylindrical sample cavity. The neutron detection efficiency is about 21% with an error of ±1.32% along the axial direction of the cavity. Using two 252Cf neutron sources, we obtained various parameters for the Singles and Doubles rates for the ASNC. The Singles, Doubles, and Triples rates for a 252Cf point source were obtained by using the MCNPX code and the results for the ft8 cap multiplicity tally option with the values of ɛ, fd, and ft measured with a strong source most closely match the measurement results to within a 1% error. A preliminary calibration curve for the ASNC was generated by using the point model equation relationship between 244Cm and 252Cf and the calibration coefficient for the non-multiplying sample is 2.78×10 5 (Doubles counts/s/g 244Cm). The preliminary calibration curves for the ACP samples were also obtained by using an MCNPX simulation. A neutron multiplication influence on an increase of the Doubles rate for a metal ingot and UO2 powder is clearly observed. These calibration curves will be modified and complemented, when hot calibration samples become available. To verify the validity of this calibration curve, a measurement of spent fuel standards for a known 244Cm mass will be performed in the near future.

  3. 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.


    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

  4. 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.


    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)

  5. KJRR-FAI Hydraulic Flow Testing Input Package

    SciTech Connect

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


    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.

  6. Korea's developmental program for superconductivity

    NASA Technical Reports Server (NTRS)

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


    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.

  7. 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.


    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.

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

    SciTech Connect

    Kim, Kyung-Doo; Chang, Won-Pyo


    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.

  9. 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


    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.

  10. 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


    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.

  11. 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.


    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.

  12. 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.


    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.

  13. International collaboration on used fuel disposition crystalline rocks

    SciTech Connect

    Wang, Yifeng; Gardner, Payton; Kim, Geon-Young; Ji, Sung-Hoon


    Active participation in international R&D is crucial for achieving the UFD long-term goals of conducting “experiments to fill data needs and confirm advanced modeling approaches” (by 2015) and of having a “robust modeling and experimental basis for evaluation of multiple disposal system options” (by 2020). DOE’s Office of Nuclear Energy (NE) and its Office of Used Fuel Disposition Research and Development (UFD) have developed a strategic plan to advance cooperation with international partners. The international collaboration on the evaluation of crystalline disposal media at Sandia National Laboratories (SNL) in FY16 focused on the following four activities: (1) thermal-hydrologic-mechanical-chemical modeling single fracture evolution; (2) simulations of flow and transport in Bedrichov Tunnel, Czech Republic, (3) completion of streaming potential testing at Korean Atomic Energy Research Institute (KAERI), and (4) technical data exchange with KAERI on thermal-hydrologic-mechanical (THM) properties and specifications of bentonite buffer materials. The first two activities are part of the Development of Coupled Models and their Validation against Experiments (DECOVALEX-2015) project.

  14. 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


    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)

  15. Neutron/gamma coupled library generation and gamma transport calculation with KARMA 1.2

    SciTech Connect

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


    KAERI has developed a lattice transport calculation code KARMA and its multi-group cross section library generation system. Recently, the multi-group cross section library generation system has included a gamma cross section generation capability and KARMA also has been improved to include a gamma transport calculation module. This paper addresses the multi-group gamma cross section generation capability for the KARMA 1.2 code and the preliminary test results of the KARMA 1.2 gamma transport calculations. The gamma transport calculation with KARMA 1.2 gives the gamma flux, gamma smeared power, and gamma energy deposition distributions. The results of the KARMA gamma calculations were compared with those of HELIOS and they showed that KARMA 1.2 gives reasonable gamma transport calculation results. (authors)

  16. 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


    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.


    SciTech Connect

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


    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.

  18. 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.


    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.

  19. Evaluation of the measurement geometries and data processing algorithms for industrial gamma tomography technology.


    Lee, N Y; Jung, S H; Kim, J B


    In this paper, we evaluated the measurement geometries and data processing algorithms for industrial gamma tomography technology. Several phantoms simulating industrial objects were tested in various conditions with the gamma-ray CT system developed in KAERI (Korea Atomic Energy Research Institute). Radiation was measured with lead shielded 24 1x1in Nal detectors. Regarding the parallel beam geometry, the EM algorithm showed the best resolution among the algebraic reconstruction technique (ART), simultaneous iterative reconstructive technique (SIRT) and expectation maximization (EM). However, the fan beam scanning was more time efficient than the parallel projection for the similar quality of reconstructed image. Future developments of the industrial gamma ray CT will be focused on a large-scale application which is more practical for a diagnosis in the petrochemical industry.

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

    SciTech Connect

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


    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.

  1. 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


    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.

  2. 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


    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


    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. 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.


    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

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


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


    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.

  6. 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


    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.

  7. 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


    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

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

    SciTech Connect

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


    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)

  9. Application of the Cold Crucible for Melting of UO{sub 2}/ZrO{sub 2} Mixtures

    SciTech Connect

    Hong, S.W.; Min, B.T.; Shin, Y.S.; Park, I.K.; Kim, J.H.; Song, J.H.; Kim, H.D.


    The melting and discharge technique of UO{sub 2}/ZrO{sub 2} mixtures using the cold crucible melting method that does not need a separate crucible such as tungsten one with high melting point is developed and applied to the KAERI FCI test called TROI. To discharge the melt from a cold crucible into a fuel-coolant interaction chamber after melting, a plug is specially designed using the concept for electro-magnetic field characteristics so as to as thin as possible the crust that is formed between the melt and plug. Its function keeps the melt in the crucible during melting period and provides the melt discharge path. About 8.5 kg melt is discharged from the cold crucible to the melt-water interaction chamber through the punched hole with 8 cm in diameter. The melt temperature is also measured and analyzed from observation of the melt surface. The power balance using the operating parameters such as current, voltage and coupling factor of R.F generator is analyzed. (authors)

  10. Evaluation of Blowdown and Condensation (B and C) Loop Behavior During Air Clearing Phase with a Prototypic Sparger Using the RELAP Code: CPT-3 Test

    SciTech Connect

    Jeung, J.S.; Ko, H.J.; Park, C.K.; Cho, S.; Song, C.H.


    KAERI has performed a series of blowdown tests to evaluate the performance of the prototypic sparger which will be used in the APR1400. Among the tests, the CPT-3 test is one of the fundamental blowdown tests and the objective of CPT-3 test was to determine the effect of air mass inside the piping on the IRWST boundary during an operation of SDVS (Safety Depressurization and Vent System). The test is conducted from the initial system pressure of 15.2 MPa and steam temperature of 343.2 deg. C. In the paper, the transient thermal-hydraulic behavior in the discharge line observed in the CPT-3 test is evaluated using the RELAP5/MOD3.1 code. The result shows that RELAP5/MOD3.1 can predict the blowdown behavior of the test facility properly and the code can be used as an evaluation tool for the verification of SDVS design of the APR 1400. (authors)

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

    SciTech Connect

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


    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.

  12. 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.


    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.

  13. Determination of the neutron fluence spectra in the neutron therapy room of KIRAMS.


    Kim, B H; Kim, J S; Kim, J L; Kim, Y S; Yang, T G; Lee, M Y


    High energy proton induced neutron fluence spectra were determined at the Korea Institute of Radiological and Medical Sciences (KIRAMS) using an extended Bonner Sphere (BS) set from the Korea Atomic Energy Research Institute (KAERI) in a series of measurements to quantify the neutron field. At the facility of the MC50 cyclotron of KIRAMS, two Be targets of different thicknesses, 1.0 and 10.5 mm, were bombarded by 35 and 45-MeV protons to produce six kinds of neutron fields, which were classified according to the measurement position and the use or no use of a beam collimator such as the gantry of the neutron therapy unit. In order to obtain a priori information to unfold the measured BS data the MCNPX code was used to calculate the neutron spectrum, and the influence of the surrounding materials for cooling the target assembly were also reviewed through this calculation. Some dosimetric quantities were determined by using the spectra determined in this measurement. Dose equivalent rates of these neutron fields ranged from 0.21 to 5.66 mSv h(-1)nA(-1) and the neutron yields for a thick Be target were 3.05 and 4.77% in the case of using a 35 and a 45-MeV proton, respectively.

  14. Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations

    NASA Astrophysics Data System (ADS)

    Kim, Do Heon; Gil, Choong-Sup; Chang, Jonghwa; Lee, Yong-Deok


    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69-group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including the new evaluations in the resonance region covering the thermal region, and the expected ENDF/B-VII including those in the upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows a maximum difference of 5.02% compared to ENDF/B-VI.7. However, the isotopic compositions of all the fission products calculated with the expected ENDF/B-VII show no differences when compared to ENDF/B-VI.7 for the thermal reactor benchmark cases.

  15. 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.


    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.

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

    SciTech Connect

    Schwarz, M.; Hardt, P. von der


    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.

  17. Response of six neutron survey meters in mixed fields of fast and thermal neutrons.


    Kim, S I; Kim, B H; Chang, I; Lee, J I; Kim, J L; Pradhan, A S


    Calibration neutron fields have been developed at KAERI (Korea Atomic Energy Research Institute) to study the responses of commonly used neutron survey meters in the presence of fast neutrons of energy around 10 MeV. The neutron fields were produced by using neutrons from the (241)Am-Be sources held in a graphite pile and a DT neutron generator. The spectral details and the ambient dose equivalent rates of the calibration fields were established, and the responses of six neutron survey meters were evaluated. Four single-moderator-based survey meters exhibited an under-responses ranging from ∼9 to 55 %. DINEUTRUN, commonly used in fields around nuclear reactors, exhibited an over-response by a factor of three in the thermal neutron field and an under-response of ∼85 % in the mixed fields. REM-500 (tissue-equivalent proportional counter) exhibited a response close to 1.0 in the fast neutron fields and an under-response of ∼50 % in the thermal neutron field.

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

    SciTech Connect

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


    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)

  19. Analysis of the Loss of Forced Reactor Coolant Flow Accident in SMART using RETRAN-03/INT

    SciTech Connect

    Kim, Tae-Wan; Suh, Kune-Yull; Lee, Un-Chul; Park, Goon-Cherl; Kim, Jae-Hak


    Small and medium integral type nuclear reactors are getting much attention for the peaceful use of nuclear energy in non-electric area such as district heating, seawater desalination and ship propulsion. An integral type nuclear co-generation reactor, SMART(System-integrated Modular Advanced ReacTor, 330 MWt), has been developed by KAERI (Korea Atomic Energy Research Institute) since 1996. In this study, the safety analysis for SMART using modified RETRAN-03 code whose name is RETRAN-03/INT is performed to examine the applicability of RETRAN-03/INT code. For the safety analysis of integral reactor with helical-coiled steam generators, RETRAN-03 code has been modified and verified using experimental results. New heat transfer coefficients are added for helical-coiled steam generator. And, the heat transfer model for steam generator is modified due to the different primary and secondary side heat flow from U-tube type steam generator. The loss of forced reactor coolant flow accident is selected for safety analysis in this study. Also it is considered as a single failure that one of three trains of passive residual heat removal system is failed. The results from MARS/SMR code and RETRAN-03/INT code are compared. (authors)

  20. Uranium and other trace elements' distribution in Korean granite: implications for the influence of iron oxides on uranium migration.


    Lee, Seung Yeop; Baik, Min Hoon


    To understand trace radionuclide (uranium) migration occurring in rocks, a granitic batholith located at the Korea Atomic Energy Research Institute (KAERI) site was selected and investigated. The rock samples obtained from this site were examined using mineralogical methods, including scanning electron microscopy (SEM) and electron probe microanalysis (EPMA). The changes in the distribution pattern of uranium (U) and small amounts of trace elements, and the mineralogical textures affected by weathering, were examined. Based on the element distribution analyses, it was found that Fe2+ released from fresh biotite is oxidized in short geological time, forming amorphous iron oxides, such as ferrihydrite, around silicate minerals. In that case, the amorphous ferrihydrite does not show distinct adsorption for U. However, as it gradually crystallizes to goethite or hematite, the most U-rich phases were found to be associated with the secondary iron oxides having granular forms. This evidence suggests that the geological subsurface environment is favorable for the crystallized iron oxides to keep their structures more stable for a long time as compared with the amorphous phases. There is a possibility that the long residence of U which is in contact with the stable crystalline phases of iron may finally lead to the partial sequestration of U in their structure. Consequently, it seems that Fe-oxide crystallization can be a dominating mechanism for U uptake and controls long-term U transport in granites with low U contents.

  1. Treatment of Radioactive Organic Wastes by an Electrochemical Oxidation

    SciTech Connect

    Kim, K.H.; Ryue, Y.G.; Kwak, K.K.; Hong, K.P.; Kim, D.H.


    A waste treatment system by using an electrochemical oxidation (MEO, Mediated Electrochemical Oxidation) was installed at KAERI (Korea Atomic Energy Research Institute) for the treatment of radioactive organic wastes, especially EDTA (Ethylene Diamine Tetraacetic Acid) generated during the decontamination activity of nuclear installations. A cerium and silver mediated electrochemical oxidation technique method has been developed as an alternative for an incineration process. An experiment to evaluate the applicability of the above two processes and to establish the conditions to operate the pilot-scale system has been carried out by changing the concentration of the catalyst and EDTA, the operational current density, the operating temperature, and the electrolyte concentration. As for the results, silver mediated oxidation was more effective in destructing the EDTA wastes than the cerium mediated oxidation process. For a constant volume of the EDTA wastes, the treatment time for the cerium-mediated oxidation was 9 hours and its conversion ratio of EDTA to water and CO{sub 2} was 90.2 % at 80 deg. C, 10 A, but the treatment time for the silver-mediated oxidation was 3 hours and its conversion ratio was 89.2 % at 30 deg. C, 10 A. (authors)

  2. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    SciTech Connect

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.


    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

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


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


    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.

  4. 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


    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.

  5. 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


    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.

  6. Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels.

    SciTech Connect

    Wolf, S. F.


    The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns.

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

    SciTech Connect

    Versey, Joshua R.


    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.

  8. Solenoid assembly with beam focusing and radiation shielding functions for the 9/6 MeV dual energy linac

    NASA Astrophysics Data System (ADS)

    Cha, Sungsu; Kim, Yujong; Ju, Jinsik; Joo, Youngwoo; Lee, Byeong-No; Lee, Soo Min; Kim, Jae Hyun; Buaphad, Pikad; Lee, Byung Cheol; Cha, Hyungki; Ha, Jang Ho; Park, Hyung Dal; Song, Ki Beak; Lee, Seung Hyun; Kim, Heesoo


    The Korea Atomic Energy Research Institute (KAERI) has been developing a Container Inspection System (CIS) by using a dual-energy (9/6 MeV) S-band (= 2856 MHz) electron linear accelerator. The key components of the CIS are the electron linear accelerator (including an electron gun, an accelerating structure, an RF power source, cooling chillers, vacuum pumps, magnet power supplies, and two solenoid magnets with beam focusing and shielding functions), a tungsten target for X-ray generation, an X-ray collimator, a detector array, and a container moving system. Generally, in accelerators, beam focusing is mainly done by solenoids operating in the region of a few MeV to keep the shape of transverse beam symmetrically round so as to reduce the loss of electrons, which increases the beam current and the beam power. In addition, a specially-designed component is needed to protect against the radiation due to the lost electrons. In this paper, we describe the design, fabrication, and optimization of two specially- designed solenoids with focusing and radiation shielding functions for a dual-energy S-band electron linear accelerator for a CIS.

  9. 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.


    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.

  10. 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


    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).

  11. Zone Freezing Study for Pyrochemical Process Waste Minimization

    SciTech Connect

    Ammon Williams


    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

  12. 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.


    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

  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


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

    SciTech Connect

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


    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)

  15. Conditioning of Waste LiCl Salt from Pyrochemical Process Using Zeolite A

    SciTech Connect

    Kim, J.G.; Lee, J.H.; Kim, E.H.; Ahn, D.H.; Kim, J.H.


    The electrolytic (LiCl-Li{sub 2}O) reduction process (Advanced spent fuel Conditioning Process; ACP) and the electrorefining process, which are being developed by the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as a LiCl salt and a LiCl-KCl eutectic salt, respectively. These waste salts must meet certain criteria for a disposal. A conditioning process composed of an immobilization and then a thermal treatment, for LiCl salt waste from the ACP has been developed using zeolite A. The immobilization of molten LiCl salt waste was conducted in a blender by mixing it with zeolite A at 923 K, producing a salt-loaded zeolite (SLZ). During the immobilization, the zeolite A was transformed to zeolite Li-A [Li{sub 2}Al{sub 2}Si{sub 2}O{sub 80}], with some minor phases such as a Li-type sodalite [Li{sub 8}Cl{sub 2}-Sod; Li{sub 8}(AlSiO{sub 4}){sub 6}Cl{sub 2}] and Nepheline for some zeolite-rich condition. In order to obtain a final ceramic waste form with Na-type sodalite [Na{sub 8}Cl{sub 2}-Sod; Na{sub 8}(AlSiO{sub 4}){sub 6}Cl{sub 2}], the very highly leach-resistant crystal phase, the SLZ with r (=LiCl/zeolite) < 0.3 should be treated in a high temperature furnace above 1173 K, which was independent from an addition of glass frit during a mixing. (authors)

  16. 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


    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

  17. First neutral beam injection experiments on KSTAR tokamak.


    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


    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.

  18. 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.


    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)

  19. Margin for In-Vessel Retention in the APR1400 - VESTA and SCDAP/RELAP5-3D Analyses

    SciTech Connect

    Joy Rempe; D. Knudson


    If cooling is inadequate 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 lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with such plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe pressurized water reactor (PWR) (AP600), which relied upon external reactor vessel cooling (ERVC) for in-vessel retention (IVR), resulted in the U.S. Nuclear Regulatory Commission (USNRC) approving the design without requiring certain conventional features common to existing light water reactors (LWRs). IVR of core melt is therefore a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced LWRs. However, it is not clear that currently proposed ERVC without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a three-year, United States (U.S.) -Korean International Nuclear Energy Research Initiative (INERI) project was initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) explored options, such as enhanced ERVC performance and an enhanced in-vessel core catcher (IVCC), that have the potential to ensure that IVR is feasible for higher power reactors.

  20. Spent fuel management status perspectives in Korea

    SciTech Connect

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


    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. 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


    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.


    SciTech Connect

    S.M. Frank


    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

  3. Release of boron and cesium or uranium from simulated borosilicate waste glasses through a compacted Ca-bentonite layer

    NASA Astrophysics Data System (ADS)

    Chun, K. S.; Kim, S. S.; Kang, C. H.


    The long-term release behavior of some elements from simulated borosilicate waste glasses (S-, K- and A-glass) in contact with a domestic compacted Ca-bentonite block and synthetic granitic groundwater at 80°C under argon atmosphere has been studied by dynamic leach tests since 1997 at KAERI. S- and K-glass differ mainly in their aluminum content, and A-glass contains 19.35 wt% UO 2 instead of fission product elements. Up to the present, the mass loss is almost the same as the normalized boron loss. This means that boron is an indicator on the dissolution of borosilicate waste glass. The leach rates of boron from K- and S-glasses after 861 days were approximately 3.1×10 -2 and 3.0×10 -2 g/ m2 day, respectively. However, the release rates of cesium through the bentonite block from K- and S-glasses were about 1/10th of the release rate of boron, which were almost the same around 2.5×10 -3 g/ m2 day. This may be due to their adsorption on the bentonite. The leach rate of boron from the A-glass was about 5.4×10 -2, but the leach rate of uranium from the A-glass specimen was quite low, below 4×10 -7 g/ m2 day. The low concentration of uranium in the leachates suggests that it hardly moves in a compacted bentonite block. By the EPMA, a yellowish uranium compound was deposited on the surface of the bentonite in contact with the A-glass specimen. The species of this phase should be identified to understand the release mechanism of uranium.

  4. 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


    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}

  5. Feasibility Study for Monitoring Actinide Elements in Process Materials Using FO-LIBS at Advanced spent fuel Conditioning Process Facility

    SciTech Connect

    Han, Bo-Young; Choi, Daewoong; Park, Se Hwan; Kim, Ho-Dong; Dae, Dongsun; Whitehouse, Andrew I.


    Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in the target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)


    SciTech Connect



    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.

  7. Discharge Characteristics of Large-Area High-Power RF Ion Source for Positive and Negative Neutral Beam Injectors

    NASA Astrophysics Data System (ADS)

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


    A large-area high-power radio-frequency (RF) driven ion source was developed for positive and negative neutral beam injectors at the Korea Atomic Energy Research Institute (KAERI). The RF ion source consists of a driver region, including a helical antenna and a discharge chamber, and an expansion region. 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 the stable and steady-state operations of high-power RF discharge. Plasma ignition of the ion source is initiated by the injection of argon-gas without a starter-filament heating, and the argon-gas is then slowly exchanged by the injection of hydrogen-gas to produce pure hydrogen plasmas. The uniformities of the plasma parameter, such as a plasma density and an electron temperature, are measured at the lowest area of the driver region using two RF-compensated electrostatic probes along the direction of the short-and long-dimensions of the driver region. The plasma parameters will be compared with those obtained at the lowest area of the expansion bucket to analyze the plasma expansion properties from the driver region to the expansion region. supported by the Ministry of Science, ICT and Future Planning of the Republic of Korea under the ITER Technology R&D Program, and National R&D Program Through the National Research Foundation of Korea (NRF) Funded by the Ministry of Science, ICT & Future Planning (NRF-2014M1A7A1A03045372)

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

    SciTech Connect

    Ehud Greenspan


    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.

  9. 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


    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)

  10. Relative biological efficiency for the induction of various gene mutations in normal and enriched with 10B Tradescantia cells by neutrons from 252Cf source.


    Cebulska-Wasilewska, A; Schneider, K; Kim, J K


    The effectiveness of neutrons from a Californium-252 source in the induction of various abnormalities in the Tradescantia clone 4430 stamen hair cells (Trad-SH assay) were studied. A special attention was paid to check whether any enhancement in effects is visible in the cells enriched with boron ions. Inflorescences, normal or pretreated with chemicals containing boron, were irradiated in the air with neutrons from a 252Cf source at KAERI, Taejon, Korea. To estimate the relative biological effectiveness (RBE) of the beam under the study, numbers of Tradescantia inflorescence without chemical pretreatment were irradiated with various doses of X-rays. The ranges of radiation doses used for neutrons were 0-1.0Gy and for X-rays 0-0.5Gy. Following the culturing according to standard procedures screening of gene and lethal mutations in somatic cells of stamen hairs was done in the extended period, between days 7 and 19 after exposures. Maximal RBE values for the induction of pink, colorless and lethal mutations were evaluated from comparison of the slopes in linear parts of the dose response curves obtained after irradiation with X-rays and californium source. The RBE(max) value or the induction of gene mutation was estimated as 7.2 comparing the value 5.6 in the studies reported earlier. The comparison of dose-response curves and its alteration, due to changes in the cells and plants environment during and after irradiation, explains the observed differences. Inflorescence pretreated with borax responded to neutrons differently depending on the biological end points. Although, for the induction of pink mutations no significant difference was observed, though, in the case of cell lethality, pretreated with boron ion plants have shoved a statistically significant increase of the RBE value from 5.5 to 34.7, and in the case of colorless mutations from 1.6 to 5.6.

  11. An Integration of the Restructured Melcor for the Midas Computer Code

    SciTech Connect

    Sunhee Park; Dong Ha Kim; Ko-Ryu Kim; Song-Won Cho


    The developmental need for a localized severe accident analysis code is on the rise. KAERI is developing a severe accident code called MIDAS, which is based on MELCOR. In order to develop the localized code (MIDAS) which simulates a severe accident in a nuclear power plant, the existing data structure is reconstructed for all the packages in MELCOR, which uses pointer variables for data transfer between the packages. During this process, new features in FORTRAN90 such as a dynamic allocation are used for an improved data saving and transferring method. Hence the readability, maintainability and portability of the MIDAS code have been enhanced. After the package-wise restructuring, the newly converted packages are integrated together. Depending on the data usage in the package, two types of packages can be defined: some use their own data within the package (let's call them independent packages) and the others share their data with other packages (dependent packages). For the independent packages, the integration process is simple to link the already converted packages together. That is, the package-wise structuring does not require further conversion of variables for the integration process. For the dependent packages, extra conversion is necessary to link them together. As the package-wise restructuring converts only the corresponding package's variables, other variables defined from other packages are not touched and remain as it is. These variables are to be converted into the new types of variables simultaneously as well as the main variables in the corresponding package. Then these dependent packages are ready for integration. In order to check whether the integration process is working well, the results from the integrated version are verified against the package-wise restructured results. Steady state runs and station blackout sequences are tested and the major variables are found to be the same each other. In order to verify the results, the integrated

  12. 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


    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

  13. Computational Fluid Dynamic simulations of pipe elbow flow.

    SciTech Connect

    Homicz, Gregory Francis


    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

  14. In-situ tracer tests and models developed to understand flow paths in a shear zone at the Grimsel Test Site, Switzerland

    NASA Astrophysics Data System (ADS)

    Blechschmidt, I.; Martin, A. J.


    how the results have been used to test and modify the hydraulic and conceptual models. *CFM partners are: BMWi / FZK-INE, Germany; JAEA, Japan; SKB / KTH, Sweden; KAERI, Korea; POSIVA, Finland; CRIEPI, Japan; NAGRA, Switzerland

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

    SciTech Connect

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


    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

  16. 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

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

    NASA Astrophysics Data System (ADS)

    Hwang, Seong Sik; Kim, Hong Pyo


    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

  18. Preface

    NASA Astrophysics Data System (ADS)

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


    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