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JAEA Reports

Evaluation of neutron fluence on JOYO core structure components

Ishida, Koichi; Maeda, Shigetaka; Saikawa, Takuya*; Masui, Tomohiko*

JNC TN9400 2002-005, 68 Pages, 2002/03

JNC-TN9400-2002-005.pdf:2.14MB

It is essential to evaluate the radiation damage of core structure materials used for core support plate and reactor vessel to maintain the safe operation of nuclear reactor plant. Therefore, surveillance tests for the irradiated specimen have been conducted in the experimental fast reactor JOYO to assure the integrity and to evaluate the life time. Neutron fluence and related spectral information are key palameters in evaluation of irradiation effects on the mechanical properties. They are usually predicted based on the calculation using the DORT two-dimensional transport code. In order to evaluate the calculation accuracy, the surveillance irradiation rigs (SVIRs) with dosimeter sets and gradient-monitor to monitor neutron fluences and temperatures were loaded several positions of the JOYO MK-II core. They were irradiated between 34$$^{th}$$ and 35$$^{th}$$ cycle. Based on the verification, the JOYO neutron field was precisely characterized and the calculated neutron flux at the positions of irradiated specimen and those of the core structure components need to be evaluated were corrected based on the experiments. As a result of this study, the following items are concluded: (1)The maximum fast neutron fluence (E$$>$$ 0.1Mev) on surveillance test specimen is determined as 2.07$$times$$10$$^{22}$$n/cm$$^{2}$$ at 9th row of the core. (2)The neutron fluences at the positions of surveillance test specimen were higher than those of the core structure components. (3)For the core support plate which seems to be most critical for JOYO life time, the fast neutron fluence at present is 9.38$$times$$10$$^{20}$$n/cm$$^{2}$$ and will reach 2.31$$times$$10$$^{21}$$ n/cm$$^{2}$$ at the end of life. The fast neutron fluence of reactor vessel is 3.12$$times$$10$$^{19}$$n/cm$$^{2}$$ at present and will reach 4.83$$times$$10$$^{19}$$n/cm$$^{2}$$ at the end of life.

JAEA Reports

JOYO impurity concentrations of radioactive nuclide data in the primary system (MK-II core)

; Saikawa, Takuya*; Masui, Tomohiko*; Arima, Toshihiro*

JNC TN9410 2001-014, 26 Pages, 2001/03

JNC-TN9410-2001-014.pdf:0.56MB

The experimental fast reactor "JOYO" served as the MK-II irradiation test bed core for testing fuel and material for FBR development for 18 years from 1982 to 2000. "JOYO" has no fuel failure since the initial criticality. Impurity concentrations of fission products data were accumulated in the primary argon gas and primary sodium during the MK-II core operation in order to obtain background value. 352 samples of primaly argon gas and the online gamma-ray monitor determined the fission products concentration data in the primary argon gas. In order to demonstrate the performance of the cold trap pre-filter, the cold trap pre-filter function confirmation tests were carried out in 1995 during 10$$^{th}$$ annual inspection. The $$^{137}$$Cs concentration data in the primary sodium were determined by 10 samples of primary sodium. The in core tag gas release tests were carried out during 29th cycle to 31st cycle. The online gamma-ray monitor determined the activation tag gas concentration data in the primary argon gas These fission products concentration data, the cold trap pre-filter function confirmation tests data and in core tag gas release tests data were compiled, which were recorded on CD-ROM for user convenience.

JAEA Reports

Improvement of material dosimetry for irradiation test in JOYO

Ito, Chikara; Aoyama, Takafumi; Masui, Tomohiko*; Saikawa, Takuya*

JNC TN9400 99-029, 26 Pages, 1999/03

JNC-TN9400-99-029.pdf:0.77MB

The material dosimetry using the multiple foil activation method has been carried out in order to assure the accuracy and reliability of neutron fluences for the irradiation tests in the experimental fast reactor "JOYO". In this study, the neutron fluences were calculated by the JOYO core management code system "MAGI" for the subassemblies which were irradiated at the positions around control rods or reflector boundary in the JOYO Mk-II core. Improvement of neutron fluence was evaluated when the "MAGI" calculation was corrected with the dosimetry results. The difference of the neutron spectrum adjustment was also investigated between ENDF/B-V and JENDL-3 dosimetry files. The major results obtained are summarized as follows; (1)The reaction rates of the dosimeters calculated by the adjusted neutron spectrum agreed well with the measured values, and its error was reduced. (2)The neutron spectrum adjusted using JENDL-3 dosimetry file was significantly improved than that by ENDF/B-V in the energy range of 10$$sim$$100 keV, because of the less error of the neutron capture cross sections of Co and Sc. (3)It showed that the correction rate of the "MAGI" calculation by the dosimetry results ranged 10$$sim$$30% for the subassemblies of the JOYO irradiation test.

JAEA Reports

Neutron intensity of spent fast reactor fuel

Takamatsu, Misao; Aoyama, Takafumi; Masui, Tomohiko*

PNC TN9410 98-011, 46 Pages, 1997/11

PNC-TN9410-98-011.pdf:1.53MB

Neutron intensity of spent fuel is important not only for the shielding design and dose evaluation of the reprocessing plant and the transportation of the mixed oxide (MOX) fuel, but also for the core management, because it contains more minor actinides than that of LWR fuel. In order to obtain the experimental data and to improve the accuracy of burnup calculation, the neutron counting rate from a spent fuel subassembly of the JOYO MK-II core with a burnup of 62,5ooMWd/t and cooling time of 5.2 years was measured in the spent fuel storage pond at JOYO. The measured neutron counting rate was then converted to the neutron intensity using the detector response which was obtained by the Monte Carlo calculation code "MCNP-4A". And the neutron intensity was compared with the calculated value based on the 3-D diffusion theory with 7 energy groups using the JOYO core management code system "MAGI". The major results obtained in this study are summarized as follows; (1) The measured neutron intensity per fuel subassembly was about 2.7$$times$$10$$^{6}$$ n/s and is about 3 times as much as that of fresh (unirradiated) fuel. (2) The average C/E value of neutron intensity was about 1.07. (3)It was found that the axial neutron intensity didn't simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of $$^{244}$$Cm.

JAEA Reports

Measurement and evaluation of dose rates for upper guide tube of control rod drive mechanism in experimental fast reactor "JOYO"

Chatani, Keiji; ; ; Masui, Tomohiko*; Nagai, Akinori; ;

PNC TN9410 92-186, 63 Pages, 1992/06

PNC-TN9410-92-186.pdf:1.64MB

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$$^{-4}$$ 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$$^{-4}$$. 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. $$^{60}$$Co dominates 92 % of CP. This value agrees with the prediction by corrosion product behavior analysis code "PSYCHE" within factor 2.

Oral presentation

Research study to advance irradiation field characterization method of Joyo MK-III core, 2; Evaluation of neutron irradiation condition by mean of neutron dosimetry

Maeda, Shigetaka; Ito, Chikara; Aoyama, Takafumi; Saikawa, Takuya*; Masui, Tomohiko*

no journal, , 

In 2003, Joyo MK-III core was upgraded to increase the irradiation testing capability. This paper describes the details of distributions of neutron flux and reaction rate in the MK-III core that was measured by characterization tests during the first two operating cycles. The calculation accuracy of the core management codes HESTIA, TORT and MCNP, was also evaluated by the measured data. The calculated fission rates of $$^{235}$$U by HESTIA agreed well with the measured one within approximately 4% in the fuel region. MCNP could simulate within 6% in the central non-fuel irradiation test subassembly and the radial reflector region, while large discrepancies were obtained in TORT results. Hence, the precise geometry model was effective in evaluating the neutron spectrum and the flux at such locations.

Oral presentation

Inspection and repair techniques in reactor vessel of sodium cooled fast reactor, 9-7; $$gamma$$ ray dose rate evaluation of upper core structure

Yamamoto, Takahiro; Ito, Keisuke; Ito, Chikara; Maeda, Shigetaka; Ito, Hideaki; Sekine, Takashi; Masui, Tomohiko*

no journal, , 

no abstracts in English

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