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Maeda, Shigetaka; Ito, Chikara; Aoyama, Takafumi; Maeda, Yukimoto; Chatani, Keiji
Transactions of the American Nuclear Society, 103(1), p.581 - 582, 2010/11
The experimental fast reactor Joyo of the Japan Atomic Energy Agency is the first liquid sodium fast reactor in Japan. Thirty years of successful operation of Joyo has shown excellent safety and reliability, and has contributed much to the LMFBR development program. Many kinds of irradiation experience have been accumulated to develop the fuels and materials for the prototype reactor Monju and future fast reactors. Accumulated data have been registered with OECD/NEA database with expectation that these data will be widely used. Joyo is presently temporary shutdown because of periodical inspection including in-vessel inspection and repair. After restart, Joyo will play a key role for a wide variety of science and technology fields as fast neutron irradiation bed.
Negishi, Hitoshi; Chatani, Keiji; Tanigawa, Shingo
Genshiryoku Nenkan 2008, p.53 - 61, 2007/09
The Japanese government assessed the result of "Feasibility Study on Commercialized FR Cycle System 2nd Phase" and a major concept that is the combination of a sodium cooled FBR (oxide fuel), an advanced aqueous reprocessing and a simplified pelletizing was selected. From now on, the JAEA invests the development resource to a major concept intensively, and aims to put it to practical use by new project, "FaCT". In "Monju", the plant improvement construction have been working on schedule, and the various tests for restart are advanced. The R&D results in "Monju" will be applied to FBR development near future. "Joyo" has been operated over 70,000 hours, and has provided the fields to develop FBR fuel & material. In addition, the external utilization of "Joyo" is enhancing now. FBR development is activating worldwide. The international cooperation under GIF, GNEP and INPRO and the research collaboration with America/France are carrying.
Koi, Mamoru; Chatani, Keiji; Hasegawa, Makoto
Saikuru Kiko Giho, (18), p.94 - 96, 2003/00
The application of the fluoride volatility process in the reprocessing of fuel from the fast breeder reactor is regarded as one of the economical methods. Plutonium hexafluoride (PuF6), however, reacting with fluorine (F2) and plutonium dioxide (PuO2) as the raw material, is in an unstable condition and tends to remain as a solid compound in the process after decomposing into plutonium tetrafluoride (PuF4). Suitable conditions should be established for the practical use of this process. One of them is to enhance the stability of PuF6. The behaviour of plutonium fluorination and relevant chemical reactions were investigated by referring to sundry literature and by thermodynamic calculation. It was then compared with recent data from laboratory scale experiments for this paper. Results from the theoretical analysis agreed with experimental observation that PuF6 could be formed stably under a high temperature condition (approx.1000 K) with over supply of figher concentration of F2.
Chatani, Keiji; Yanagisawa, Tsutomu;
International Scientific-Practieal Confenence XXI; Nuclear Weapon Free Century, 0 Pages, 2001/00
None
Chatani, Keiji; ; Ito, Chikara; Setyadi*; ;
PNC TN9410 94-032, 76 Pages, 1993/12
Deposition density and gamma dose rate of the radioactive corrosion product(CP) have been measured along the primary sodium loop in Experimental Fast Reactor "JOYO" during every annual inspection in order to make clear the CP behavior and to verify the CP behavior analysis code "PSYCHE". As a result of the previous seven CP measurements, it is made clear that major CP nuclides deposited in the primary sodium loops are 
Mn on the cold leg (CL) piping (from the outlet of the Intermediate Heat Exchanger (IHX) to the inlet of the reactor vessel (R/V)) and 
Co on the hot leg (HL) piping (from the outlet of the R/V to the inlet of the IHX), and so on. In this study the CP behavior has been evaluated by using the measurement results during the 10th annual inspection. The results of this study are summarized as follows: (1)The distribution of the measured CP deposition density and gamma dose rate show the same tendency observed previously and no unusual phenomenon is observed. (2)The buildup of Mn and Co is seemed to reach saturation because the average CP deposition density and the averaged dose rate is the same value as measured in the 9th annual inspection. (a)Deposition density of Mn is 30kBq/cm
for HL piping, 60 kBq/cm at CL(1) piping (from the outlet of the IHX to the inlet of the pump) and 130 kBq/cm at CL(2) piping (from the outlet of the pump to the inlet of the R/V). Deposition density of Co is 9 kBq/cm at HL piping, 3kBq/cm at CL(1) piping and 9 kBq/cm at CL(2) piping. (b)Dose rate is 0.5mSV/h at HL piping, 0.6mSV/h at CL(1) piping and 1 mSV/h at CL(2) piping. (3)As a comparison between "PSYCHE" calculation(C) and measurement(E), C/E ratio is 0.9 to 1.5 for CP deposition density, and 1.6 for dose rate. The agreement between calculation and measurement is fairly good.
Isozaki, Kazunori; ; Ito, Hideaki; ; Chatani, Keiji; ;
PNC TN9520 93-006, 198 Pages, 1992/11
None
; Chatani, Keiji; ; ; ;
PNC TN9410 92-345, 166 Pages, 1992/10
An evaluation about the radioactive corrosion product (CP) behaviour in sodium cooling systems of a fast reactor is presented in this report, based on the obtained measurement results in the operating experience of JOYO. The objective of this work is to update the calculational model for predicting the release and deposition behaviour of CP in primary sodium cooling systems of a fast reactor. The evaluation results are as follows; (1)The main radionuclides of CPs transported to the out-of-reactor primary sodium loop are Mn and Co, and Mn is the most dominant. On the other hand, Co is the most dominant nuclide found in the liquid waste from spent fuel cleaning, which is produced by removal of activated CP deposits from surfaces of core sub-assemblies in sodium cleaning. (2)The deposition rate of Mn onto the hot-leg (HL) piping walls corresponds fairly with the saturation of radioactivity induced in core materials by activation, on the other hand, that onto the cold-leg (CL) piping walls has been being accelerated. The deposition rate of Co, due to the dependency of activation and release in a core, is strongly affected by the re-fuelling pattern and the oxygen concentration in sodium, and suggests the detouching process of deposits from wall surfaces. (3)Although Mn was transported and deposited preferentially in the HL of the primary cooling system in an early stage, the transport and deposition in the CL regions has overcomed that in the HL along operating time. Co was transported and deposited preferentially in the HL and the similar distribution pattern has been maintained thoroughout the operating periods. (4)The solution - precipitation model for CP behaviour in flowing sodium system was verified via the sensitivity test of model parameters and optimizing them on the above mentioned results, giving the measured to caluculated values of 1.36 or 1.03 for Mn or Co buildup, and 1.61 ...
; Chatani, Keiji; ; ;
PNC TN9410 92-224, 81 Pages, 1992/07
The radioactive corrosion product (CP) deposition density and gamma dose rate have been measured along the primary sodium circuits in Experimental Fast Reactor "JOYO" during every annual inspection and the CP behavior analysis code "PSYCHE" has been verified with measurement data in order to contribute the reduction of exposure dose of plant personal. The deposition density is measured by using a pure germanium detector system and determined by multiplying count rates by conversion factor. Gamma dose rate is measured with CaSO
thermoluminescence dosimeters (TLD). This report presents measurement results during the 9th annual inspection and the evaluation results for all data measured so far. The results on this study are summarized as follows: (1)Major CP nuclides deposited along the primary sodium circuits are Mn and Co. Mn is most dominant isotopes. Amounts of deposited Mn is about twenty times as much as those of Co. (2)Mn is deposited mainly on the cold leg pipings between the outlet of the intermediate heat exchanger (IHX) and the inlet of the reactor vessel. Co is deposited mainly on the hot leg pipings between the outlet of the reactor vessel and the inlet of IHX. (3)The buildup of Mn is saturated at 4
4.5 EFPY. The averaged dose rate of the pipings is saturated at about 1.5mSV/h. The dose rates of IHX and primary sodium pump are about 1.5 mSv/h and 2.1 mSv/h, respectively. The dose rate distributions around IHX and primary sodium pump show the peaks at the stagnant part of the flow and at the turbulence part. (4)Calculation by "PSYCHE" and measurement are compared. Calculation-to-measurement ratio is 1.2 for the CP deposition density and 1.5 for the dose rate. It can be said that the features of the CP behavior in the primary circuit of "JOYO" is made clear. The more effort will be required for the evaluation of CP behavior for subassemblies such as outer reflectors, clearness of ...
Chatani, Keiji; ; ; Masui, Tomohiko*; Nagai, Akinori; ;
PNC TN9410 92-186, 63 Pages, 1992/06
Dose rates around UGT (Upper Guide Tube) of CRDM (Control Rod Drive Mechanism) have been measured in Experimental Fast Reactor "JOYO" during the 9th periodical inspection in order to reflect the study on the shield thickness of UIS (Upper Internal Structure) cask, which has been planned to be used for a Large Fast Reactor. Absolute amount of radioactive corrosion products (CP) is evaluated by gamma spectra analysis for waste water from cleaned UGT. The results on this study are summarized as follows: (1)Measured dose rates distribution around UGT before and after clean-up show the same reduction. The affection of CP is not clearly observed for the dose rate distribution. (2)The relative values of dose rate, which are evaluated by considering the inside structure of UGT, show the attenuation of 10
from bottom to sodium level of UGT. The above relative distribution agrees well with that of measurement data using U-235 fission chamber, which was conducted at MK-I core start-up tests, except the stellite region. (3)As to the relative values of dose rate, calculation by "DOT3.5" and estimation by measured dose rate agree within factor 3 for the attenuation of 10. It is confirmed that the calculation can predict well the measurement. (4)Absolute amount of CP estimated by gamma spectra analysis and waste water analysis is 180 MBq. Co dominates 92 % of CP. This value agrees with the prediction by corrosion product behavior analysis code "PSYCHE" within factor 2.
; Chatani, Keiji; ;
PNC TN9450 92-001, 156 Pages, 1992/03
This report is intended to make it easier to apply the measured date obtained from the Axial Shield Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) in 1990 as part of a series of eight experiments planned for Japanese-American Shielding Program for Experimental Research (JASPER) program starting in 1986. The Axial Shield Experiment was planned to study the neutron attenuation characteristics of the axial shield, which is designed in the fuel assembly to reduce the neutron fluence in regions above the core. In order that the experimental neutron spectrum would be representative of the expected neutron spectra directly above the FBR core, the Tower Shielding Reacor (TSR) source spectrum was altered by a spectrum modifier, which was used in two previous experiments also. The modified spectrum entered the test section, which consisted of seven hexagonal shield assemblies surrounded by B,C and concrete. Three different axial shield designs were studies, Either B,C or stainless steel was used as a shielding material. Neutron measurements were made with various detectors behind the experimental configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-11839, "Measurements for JASPER Program Axial Shield Experiment"). Additional information reported by the PNC assignee is utilized also.
Chatani, Keiji; ; ; ; *; *; *
PNC TN9410 92-076, 348 Pages, 1992/03
JASPER (Japanese American Shielding Program of Experimental Researches) is the cooperative research program between PNC and US-DOE using TSF (Tower Shielding Facility) in ORNL (Oak Ridge National Laboratory) as the experiment facility. This report summarizes the works in FY'1991 as follows; Planning the experiment configuration for JASPER Program, Analyses of the JASPER Program experiment, Analyses of the former TSF experiment, and Development of the methods for FBR shielding analyses. (1)Analyses of the JASPER Program Experiment In FY'1991 Axial shield Experiment data were mainly analyzed, and some of In-vessel Fuel Storage (IVS) Experiment data were also analyzed. The Fast Reactor Shielding Analysis System developed by PNC has been applied to the analyses for JASPER Program experiments. (Axial Shield Experiment Analysis) Axial Shield Experiment was conducted from August 1990 through December 1990 as part of a continuing series of eight experimennts planned for the JASPER Program. The experiment serves not only to provide data for the verification of analysis system in calculating the neutron streaming in each design, but also to provide a basis for determining the shielding effectiveness of stainless steel (SS) and boron carbide (BC). four types of experimental configuration were used. The conclusions of the analyses are as follows: (a)For the spectrum modifier which provides a spectrum of neutron representative of those incident on the axial shield for the FBR core, the two-dimensional calculation showed good agreement with the experimental data. It was confirmed that the two-dimensional calculation could estimate the experimental data with almost the same accuracy as in the analyses for the Radial shield Attenuation and the Fission Gas Plenum Experiments. (b)For the homogeneous mockups, the two-dimensional ealculation could give the good agreement with the experimental data. (c)For the central blockage type mockups, in which the coolant flows ...
*; Chatani, Keiji*; *; Nakanishi, Seiji; *
PNC TN9410 87-182, 79 Pages, 1987/12
In order to optimize the fuel handling system of Large Scale-up Fast Breeder Reactor, the equipment design of the IVS(In-vessel Storage) type fuel handling system and of the EVS (EX-vessel Storage) type were conducted and compared with each other. As concerns IVS type, the equipments designed in Key Technological Design Study (II) was reduced by about 13% for elimination of the air cell and high density storage of spent fuel. With regard to EVS type, it was found that Holding Cylinder Sodium pot EVS type was practical and that it could store spent fuel with 20kW decay heat for each assembly at maximum. It can be considered that EVS type is coordinate with IVS type from a viewpoint of materials and advantageous to Large Scale-up Fast Breeder Reactor for (1)reduction of reactor vessel diameter, (2)elimination of the fuel handling process from sodium environment to water, and (3)realization of ex-vessel NIS.
; Chatani, Keiji; Ito, Kazuhiro; Suzuki, Soju;
Liquid Metal Systems; Material Behavior and Physical Chemistry in Liquid Metal Systems 2, ,
None
Aoyama, Takafumi; Ito, Chikara; Okawachi, Yasushi; Maeda, Shigetaka; Suzuki, Soju; Chatani, Keiji; Takeda, Toshikazu*
no journal, ,
no abstracts in English
Aoyama, Takafumi; Maeda, Yukimoto; Chatani, Keiji
no journal, ,
In the field of structure material development, developing high irradiation resistant material and making design formula by considering irradiation damage are necessary. In Joyo, a lot of irradiation test for developing these materials were conducted by high level irradiation technology and flexible operation with its high level fast neutron flux. And there are several post irradiation examination facilities around Joyo and they make irradiation center. This report is to describe irradiation capabilities and roles of Joyo in the planning session of material section.
Sekine, Takashi; Kitamura, Ryoichi; Maeda, Yukimoto; Chatani, Keiji
no journal, ,
Since in-vessel observation for a sodium fast reactor has to be conducted under severe conditions that include high temperatures (approximately 200
C) and high radiation doses (approximately 200 Gy/h). In addition, since the primary sodium coolant has to be always kept in the reactor vessel to remove the decay heat of fuel subassemblies. Therefore an in-vessel observation equipment has to be designed to not only stand the severe conditions but also be capable of being inserted into the sealed reactor vessel through the holes built into the rotating-plug. In Joyo, in-vessel inspections and repair technologies are developed and applied to observe the upper core structure and bent irradiation test subassembly. The detail design of the Joyo restoration work is now pushed forward based on the in-vessel observation results.
Kitamura, Ryoichi; Sekine, Takashi; Maeda, Yukimoto; Chatani, Keiji
no journal, ,
no abstracts in English
