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Tobita, Minoru*; Konda, Miki; Omori, Takeshi*; Nabatame, Tsutomu*; Onizawa, Takashi*; Kurosawa, Katsuaki*; Haraga, Tomoko; Aono, Ryuji; Mitsukai, Akina; Tsuchida, Daiki; et al.
JAEA-Data/Code 2022-007, 40 Pages, 2022/11
JAEA-Data-Code-2022-007.pdf:1.99MB
Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete, ash, ceramic and brick samples generated from JRR-3, JRR4 and JRTF facilities. In this report, we summarized the radioactivity concentrations of 24 radionuclides (
H, 
C, 
Cl, 
Ca, 
Co, 
Ni, 
Sr, 
Nb, 
Tc, 
Ag, 
I, 
Cs, 
Ba, 
Eu, 
Eu, 
Ho, 
U, 
U, Pu, 
Pu, 
Pu, 
Am, 
Am, 
Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2020-2021.
Hamamoto, Shimpei; Ishitsuka, Etsuo; Nakagawa, Shigeaki; Goto, Minoru; Matsuura, Hideaki*; Katayama, Kazunari*; Otsuka, Teppei*; Tobita, Kenji*
Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), 5 Pages, 2021/10
Impurity concentrations of hydrogen and hydride in the coolant were investigated in detail for the HTTR, a block type high-temperature gas reactor owned by Japan. As a result, it was found that CH
was 1/10 of H
concentration, which was under the conventional detection limit. If the ratio of H to CH in the coolant is the same as the ratio of HT to CH
T, the CHT has a larger dose conversion factor, and this compositional ratio is an important finding for the optimal dose evaluation. Further investigation of the origin of CH suggested that CH was produced as a result of a thermal equilibrium reaction rather than being released as an impurity from the core.
Tsuchida, Daiki; Haraga, Tomoko; Tobita, Minoru*; Omori, Hiroyuki*; Omori, Takeshi*; Murakami, Hideaki*; Mitsukai, Akina; Aono, Ryuji; Ishimori, Kenichiro; Kameo, Yutaka
JAEA-Data/Code 2020-022, 34 Pages, 2021/03
JAEA-Data-Code-2020-022.pdf:1.74MB
Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete samples generated from JRR-3 and JPDR. In this report, we summarized the radioactivity concentrations of 22 radionuclides(H, C, Cl, Ca, Co, Ni, Sr, Nb, Ag, Ba, Cs, Eu, Eu, Ho, U, U, Pu, 
Pu, Am, Am, Cm) which were obtained from radiochemical analysis of the samples.
Kobayashi, Jun; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki
JAEA-Research 2012-014, 40 Pages, 2012/07
JAEA-Research-2012-014.pdf:7.8MB
As the temperatures difference between the control rod channels and the core fuel subassemblies is around 100 
C centigrade, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of upper internal structure (UIS) in JSFR. Then, a water experiment was conducted using an 1/3 scale 60 sector model. As a result, thermal striping phenomena in the region between the fuel subassembly outlet and the bottom of the UIS were grasped. The modified geometry of the UIS bottom and the handling head of the primary CR channel was created so as to suppress the cold jets from the CR channels. The comparison of measured temperature fluctuations around the CR channels revealed that the modified geometry was effective to decrease the temperature fluctuation intensity and amplitude in the sensitive frequency band to the stress conversion.
Ezure, Toshiki; Miyake, Yasuhiro*; Tobita, Akira; Kimura, Nobuyuki; Kamide, Hideki
JAEA-Research 2012-005, 56 Pages, 2012/05
JAEA-Research-2012-005.pdf:10.06MB
In the design of JSFR, a two-loop cooling system and a compact reactor vessel are employed to achieve the economical improvement. However, these innovative designs lead to the increase of coolant velocity. As the results, strong vortices at the H/L intake may cause the cavitation (vortex cavitation). Therefore, the evaluation of occurrence behavior of vortex cavitation is the important issue for the structural integrity of reactor. In the present study, fundamental water experiments were performed in the cylindrical tank geometry. The water temperature was varied from 10C to 80C to clarify the influence of kinematic viscosity, 
. The occurrences behaviors of vortex cavitation were evaluated quantitatively by visualization measurement and image analysis. As the results, it was clarified that there was little dependence on under the small conditions, while it was relatively obvious under the large conditions.
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
Nuclear Engineering and Design, 241(11), p.4409 - 4419, 2011/11
Times Cited Count：55 Percentile：96.64(Nuclear Science & Technology)In the design of Japan Sodium-cooled Fast Reactor (JSFR), the flow-induced vibration (FIV) has been studied for the large-diameter hot-leg pipe with a short-elbow. The FIV will have a excitation source which is caused by the pressure fluctuation in the pipe. In this study, water experiments with two types of elbows with different curvature ratios were conducted in order to investigate the interaction between flow separation and the secondary flow due to the elbow curvature. The experiments were conducted with the short-elbow and the long-elbow under Re = 1.8E5 and 5.4E5 conditions.
Kobayashi, Jun; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki; Watanabe, Osamu*; Oyama, Kazuhiro*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Design study of an advanced loop-type sodium-cooled fast reactor, JSFR, has been carried out in a frame work of Fast Reactor Cycle Technology Development Project (FaCT) in Japan. As the temperature differences among the control rod channels, blanket assemblies and the core fuel assemblies are 100C centigrade in the maximum, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). In this investigation, a water experiment was conducted using a 1/3 scale 60 sector model of the core and reactor upper plenum. Characteristics of temperature fluctuations near the cold fluid outlets were obtained and it was confirmed that several countermeasures can reduce temperature fluctuations at the bottom of UIS.
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09
In the design of Japan Sodium-cooled Fast Reactor (JSFR), the flow-induced vibration (FIV) has been studied for the large-diameter hot-leg pipe with a short-elbow. The FIV will be a phenomenon caused by the pressure fluctuation in the pipe. In this study, for the purpose of clarification of FIV mechanism, the velocity and pressure fluctuations in the elbow pipe were measured. It was found that the pressure fluctuation on the wall with elbow was closely related with the movement of separation region formed near the elbow outlet.
Kimura, Nobuyuki; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki; Nagasawa, Kazuyoshi*
JAEA-Research 2010-065, 191 Pages, 2011/03
JAEA-Research-2010-065.pdf:14.57MB
Thermal stratification water experiments using a 1/10th scale model were carried out for an advanced loop type sodium cooled reactor. Experimental parameters were core outlet velocity, temperature difference during scram, and height of the plug which infill the hole at the dipped plates for setup of a fuel handling machine. It was found that the height and the rising speed of stratification interface depended on the Richardson number. Furthermore the temperature gradient of the stratification interface depended on the Peclet number.
Ezure, Toshiki; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11
A fundamental water experiment was performed in a cylindrical tank geometry to investigate the influences of fluid viscosity on the cavitation due to sub-surface vortex (vortex cavitation). In order to clarify the influence of fluid viscosity, the fluid temperature was varied from 10 C to 80 C to change the kinetic viscosity () of fluid from 1.3
10
to 3.710
m
/s. The occurrences of vortex cavitation were detected by image analysis on digital images of vortex cavitation captured by a digital CMOS camera. Then, the occurrences of vortex cavitation were evaluated from the relation between the yield fraction curves of vortex cavitation and the cavitation factor under several different conditions. The experimental results showed that the influence of was obvious under the large conditions. However, the influence became smaller according to the decrease of .
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 9 Pages, 2010/11
In the design of Japan Sodium-cooled Fast Reactor (JSFR), the flow-induced vibration (FIV) arising from the piping geometry and operating condition may occur in the short-elbow pipe. In this study, water experiments using two types of 1/8 scaled elbows with different curvature ratio, r/D = 1.0 and 1.5, were conducted in order to investigate the mechanism of flow fluctuation due to the elbow curvature. The measurements of the velocity fields and pressure fluctuation revealed that the periodic pressure fluctuation of St = 0.56 occurred near the separation region which was formed constantly on the wall of inside.
Oto, Tsutomu; Kimura, Tadashi; Miyauchi, Masaru; Nemoto, Nobuaki; Tobita, Kenji; Fukasaku, Akitomi; Takahashi, Kunihiro
JAEA-Review 2010-017, 21 Pages, 2010/07
JAEA-Review-2010-017.pdf:7.68MB
An integrity investigation of a Canal Expanded Joint was carried out as one of the integrity investigation of the JMTR reactor building related facilities and components, before the repair or replacement work of the JMTR related facilities that had begun in FY2007. The Canal Expanded Joint will be used for long-term after the JMTR restart. In the integrity investigation, the visual inspection, the performance test (Surface observations, Durometer hardness test) were investigated respectively and the integrity of the Canal Expanded Joint was confirmed. In order to use the Canal Expanded Joint continuously for long-term, it is important for maintaining the integrity of the Canal Expanded Joint by the periodical maintenance and the repairing work including that has been conducted up to now.
Miyauchi, Masaru; Kimura, Tadashi; Oto, Tsutomu; Nemoto, Nobuaki; Tobita, Kenji; Fukasaku, Akitomi; Takahashi, Kunihiro
JAEA-Review 2010-008, 106 Pages, 2010/06
JAEA-Review-2010-008.pdf:8.89MB
An integrity investigation was carried out for the JMTR Concrete Structure (Vent Stack, Trench, Canal Building, Filter Bank), which was the concrete structure and would be used for the long-term after JMTR restart, before the repair or replacement work of the JMTR related facilities that had begun in FY2007. In the integrity investigation, the concrete surface deterioration, the rebound number (nondestructive strength test), compressive strength using drilled concrete core test piece, the static modules of elasticity, the carbonation depth, the reinforced bar corrosion and the chloride ion content were investigated respectively and the integrity of concrete was confirmed. After the investigation, repair works such as re-painting of the Vent Stack and Trench were carried out from the viewpoint of prevention of flaking off, floating of the painting and the thinning due to the investigation results. In order to use the JMTR concrete structure continuously for long-term, it is important for maintaining the integrity of a concrete structure by the periodical maintenance and the repairing work including the building outer-wall surface painting that has been conducted up to now.
Kimura, Tadashi; Oto, Tsutomu; Miyauchi, Masaru; Nemoto, Nobuaki; Tobita, Kenji; Fukasaku, Akitomi; Takahashi, Kunihiro
JAEA-Review 2010-001, 27 Pages, 2010/03
JAEA-Review-2010-001.pdf:18.81MB
In order to investigate an integrity of the UCL (Utility Cooling Line) elevated water tank to be used for a long term after the JMTR restart, investigation on the base of the UCL elevated tank, especially for the part which had a significant aged effect, was carried out before the refurbishment work of the JMTR related facilities which had begun in FY2007. In the integrity investigation, it was confirmed that some part of the base bolts had significant aged effects, there were no evidence of crack and false indication in the welding region though thinning were observed in some parts of the base plate and the surrounding plate. After the investigation, repair works such as re-painting of the UCL elevated water tank were carried out from the viewpoint of prevention of flaking off, floating of the painting and the thinning due to corrosion based on the investigation results. In order to maintain the integrity of the UCL elevated tank, the periodical maintenance and the repair works of the base of the UCL elevated tank are important for continuous use of it in future.
Kimura, Tadashi; Oto, Tsutomu; Miyauchi, Masaru; Nemoto, Nobuaki; Tobita, Kenji; Fukasaku, Akitomi; Takahashi, Kunihiro
JAEA-Review 2009-054, 73 Pages, 2010/03
JAEA-Review-2009-054.pdf:6.79MB
An integrity investigation was carried out for the JMTR reactor building, which was the concrete structure and would be used for the long-term after JMTR restart, before the repair or replacement work of the JMTR related facilities that had begun in FY2007. In the integrity investigation, the concrete surface deterioration, the rebound number (nondestructive strength test), compressive strength using drilled concrete core test piece, the static modules of elasticity, the carbonation depth, the reinforced bar corrosion and the chloride ion content were investigated respectively and the integrity of concrete was confirmed. In order to use the JMTR reactor building continuously for long-term, it is important for maintaining the integrity of a concrete structure by the periodical maintenance and the repairing work including the building outer-wall surface painting that has been conducted up to now.
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
JAEA-Research 2009-051, 30 Pages, 2010/02
JAEA-Research-2009-051.pdf:6.22MB
The main cooling system of Japan Sodium-cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9 m/s in the primary hot leg pipe with large-diameter. The maximum Reynolds number in the piping reaches 4.210
. The hot leg pipe having an elbow with curvature ratio of r/D = 1.0, so-called "short elbow", which enables a compact reactor vessel. Under such a system condition, the flow-induced vibration (FIV) is concerned at the short elbow. It is significant to obtain the knowledge of the fluctuation intensity and spectra of velocity and pressure fluctuations in order to grasp the mechanism of the FIV. In this study, water experiments were conducted. Two types of 1/8 scaled elbows with different curvature ratio, r/D = 1.0, 1.5, were used to investigate the influence of curvature on velocity fluctuation at the elbow. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow at the elbow outlet and separation flow at the inner wall of elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow.
Kimura, Nobuyuki; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki; Miyake, Yasuhiro*
JAEA-Research 2009-026, 160 Pages, 2009/10
JAEA-Research-2009-026.pdf:15.97MB
Thermal stratification after a scram is one of main thermal loads of a reactor vessel in sodium cooled fast reactor. Water experiments using an 1/10th scaled model were carried out for an advanced loop type sodium cooled reactor. The reactor vessel is highly compact and has an upper inner structure (UIS), which has a slit in radial direction for fuel handling. The jet from core outlet goes through the UIS slit and makes asymmetric and locally high velocity field in the reactor upper plenum. Steep temperature distribution across the stratification interface and temperature fluctuations were found near the UIS slit. It was revealed that they were resulted from the impingement of the jet through the slit at the interface. Parameter experiments showed the characteristics of thermal stratification interface.
Ezure, Toshiki; Kimura, Nobuyuki; Tobita, Akira; Miyakoshi, Hiroyuki; Kamide, Hideki
JAEA-Research 2009-021, 44 Pages, 2009/09
JAEA-Research-2009-021.pdf:22.4MB
An compact sodium-cooled fast reactor has been investigated in the Fast Reactor Cycle Technology Development Project. The compact-sizing may cause the cover gas entrainment (GE) at the free surface of reactor vessel. The prevention of gas entrainment is one of the significant thermal hydraulic issues. In the previous study, a water experiment in the 1/1.8 scale partial model was performed to check the adequacy of reactor design. From the result, it has been assured GE did not occur in the rated condition of reactor. However, there are the specific operating conditions with the low sodium level due to the thermal contraction concerning the cold startup or shutdown of reactor, where the sodium temperature is low. The objective of this study is to evaluate the onset condition of gas entrainment in those condition. The onset condition of GE was clarified in low liquid level. From the result, it was assured gas entrainment did not occur with the doubled D/P geometry.
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 14 Pages, 2009/09
The main cooling system of Japan Sodium-Cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9 m/s in the primary hot leg pipe having a short elbow with large-diameter. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned. The evaluation of the structural integrity of pipings in JSFR will need to grasp the mechanism of FIV at the elbow. In this study, two types of 1/8 scaled elbows with different curvature ratio were used to investigate the influence of curvature on velocity fluctuation. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow and separation flow in the elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow.
Ebisawa, Hiroyuki; Hanakawa, Hiroki; Asano, Norikazu; Kusunoki, Hidehiko; Yanai, Tomohiro; Sato, Shinichi; Miyauchi, Masaru; Oto, Tsutomu; Kimura, Tadashi; Kawamata, Takanori; et al.
JAEA-Technology 2009-030, 165 Pages, 2009/07
JAEA-Technology-2009-030.pdf:69.18MB
The condition of facilities and machinery used continuously were investigated before the renewal work of JMTR on FY 2007. The subjects of investigation were reactor building, primary cooling system tanks, secondary cooling system piping and tower, emergency generator and so on. As the result, it was confirmed that some facilities and machinery were necessary to repair and others were used continuously for long term by maintaining on the long-term maintenance plan. JMTR is planed to renew by the result of this investigation.
