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JAEA Reports

Study on the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute (FY2020)

Asakura, Kazuki; Shimomura, Yusuke; Donomae, Yasushi; Abe, Kazuyuki; Kitamura, Ryoichi; Miyakoshi, Hiroyuki; Takamatsu, Misao; Sakamoto, Naoki; Isozaki, Ryosuke; Onishi, Takashi; et al.

JAEA-Review 2021-020, 42 Pages, 2021/10

JAEA-Review-2021-020.pdf:2.95MB

The disposal of radioactive waste from the research facility need to calculate from the radioactivity concentration that based on variously nuclear fuels and materials. In Japan Atomic Energy Agency Oarai Research and Development Institute, the study on considering disposal is being advanced among the facilities which generate radioactive waste as well as the facilities which process radioactive waste. This report summarizes a study result in FY2020 about the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute.

JAEA Reports

Study on the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute (FY2019)

Asakura, Kazuki; Shimomura, Yusuke; Donomae, Yasushi; Abe, Kazuyuki; Kitamura, Ryoichi

JAEA-Review 2020-015, 66 Pages, 2020/09

JAEA-Review-2020-015.pdf:4.27MB

The disposal of radioactive waste from the research facility need to calculated from the radioactivity concentration that based on variously nuclear fuels and materials. In Japan Atomic Energy Oarai Research and Development Institute, the study on considering disposal is being advanced among the facilities which generate radioactive waste as well as the facilities which process radioactive waste. This report summarizes a study result in FY2019 about the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute.

JAEA Reports

Impact assessment of the forest fires on Oarai Research and Development Center Waste Treatment Facility

Shimomura, Yusuke; Hanari, Akira*; Sato, Isamu*; Kitamura, Ryoichi

JAEA-Technology 2015-062, 47 Pages, 2016/03

JAEA-Technology-2015-062.pdf:1.85MB

In response to new standards for regulating waste management facilities, it was carried out impact assessment of forest fires on the waste management facilities existed in Oarai Research and Development Center of Japan Atomic Energy Agency. At first, a fire spread scenario of forest fires was assumed. The intensity of forest fires was evaluated from field surveys, forest fire evaluation models and so on. As models of forest fire intensity evaluation, Rothermel Model and Canadian Forest Fire Behavior Prediction (FBP) System were used. Impact assessment of radiant heat to the facilities was carried out, and temperature change of outer walls for the assumed forest fires was estimated. The outer wall temperature of facilities was estimated around 160$$^{circ}$$C at the maximum, it was revealed that it doesn't reach allowable temperature limit. Consequently, it doesn't influence the strength of concrete. In addition, a probability of fire breach was estimated to be about 20%. This report illustrates an example of evaluation of forest fires for the new regulatory standards through impact assessment of the forest fires on the waste management facilities.

JAEA Reports

Corroborative tests for Oarai Waste Reduction Treatment Facility using the in-can type high frequency induction heating method

Sakauchi, Hitoshi; Sato, Isamu*; Donomae, Yasushi; Kitamura, Ryoichi

JAEA-Technology 2015-059, 352 Pages, 2016/03

JAEA-Technology-2015-059.pdf:51.53MB

OWTF (Oarai Waste Reduction Treatment Facility) is constructed for volume reduction processing and stabilization treatment of $$alpha$$ solid waste, which was generated from hot facilities in Oarai Research and Develop Center of Japan Atomic Energy Agency, using in-can type high frequency induction heating by remote control. This report describes corroborative tests, in which incinerating and melting performance for OWTF is confirmed with a full-scale testing furnace. We have been carrying out the tests of incinerating and melting treatment with some kinds of simulated wastes, such as enclosure form of radioactive wastes, material and articles.

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Observation technical development in a reactor vessel of the fast reactor

Imaizumi, Kazuyuki; Saito, Takakazu; Tobita, Shigeharu; Nagai, Akinori; Kitamura, Ryoichi; Okazaki, Yoshihiro

JAEA-Technology 2012-027, 49 Pages, 2012/08

JAEA-Technology-2012-027.pdf:7.07MB

In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. In order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident occurred in Joyo, the following observation systems were specifically developed for Joyo. And the following two observations were conducted. (1) Simple overhead observation using a standard video camera for the top of the sub-assemblies and the in-vessel storage rack (2) Narrow space observation using remote handling device equipped with radiation-resistant fiberscope for the bottom face of the upper core structure. As a result, the observations under the actual reactor environment were successfully made even in the narrow space in the reactor vessel and the results provided useful information on incident investigations and planning of restoration work.

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Development of a high radiation resistance fiberscope (Joint research)

Naito, Hiroyuki; Itagaki, Wataru; Okazaki, Yoshihiro; Imaizumi, Kazuyuki; Ito, Chikara; Nagai, Akinori; Kitamura, Ryoichi; Shamoto, Naoki*; Takeshima, Yoshiyuki*

JAEA-Technology 2012-009, 100 Pages, 2012/05

JAEA-Technology-2012-009.pdf:9.89MB

The radiation characteristics of image fiber and light guide fiber were evaluated to develop a high radiation resistant fiberscope for the fast reactor in-vessel observation. It is known that a pure silica core fiber has a high radiation resistance and radiation resistance is influenced with impurities in silica. Moreover it is necessary to change the clad material of the light guide fiber because that of the current light guide fiber is acrylate, which is weak against radiation. Hence the improved fibers consist of a pure silica core with 1,000 ppm OH and fluorine-doped silica clad. As a result of a $$gamma$$ irradiation test, we confirm that OH inhibits the generation of the precursor by $$gamma$$ irradiation. About the clad material, we confirmed that the transmission loss of the fluorine-doped silica clad fiber is smaller than that of the acrylate clad fiber. About the mechanical strength of a fiber, we confirmed that there is no weakening the strength of the fiber and no exfoliation of the coating from the glass. In this study, we discovered the fiber which consists of a pure silica core with 1,000 ppm OH and fluorine-doped silica clad has a high radiation resistance and it is possible to observe using this fiber under the 200 $$^{circ}$$C after 5$$times$$10$$^{5}$$ Gy irradiation.

JAEA Reports

Plan and reports of coupled irradiation (JRR-3 and JOYO of research reactors) and hot facilities work (WASTEF, JMTR-HL, MMF and FMF); R&D project on irradiation damage management technology for structural materials of long-life nuclear plant

Matsui, Yoshinori; Takahashi, Hiroyuki; Yamamoto, Masaya; Nakata, Masahito; Yoshitake, Tsunemitsu; Abe, Kazuyuki; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ishikawa, Kazuyoshi; Kikuchi, Taiji; et al.

JAEA-Technology 2009-072, 144 Pages, 2010/03

JAEA-Technology-2009-072.pdf:45.01MB

"R&D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant" was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of "Evaluation of Irradiation Damage Indicator" in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research-and-Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency.

Journal Articles

In-pile creep rupture properties of ODS ferritic steel claddings

Kaito, Takeji; Otsuka, Satoshi; Inoue, Masaki; Asayama, Tai; Uwaba, Tomoyuki; Mizuta, Shunji; Ukai, Shigeharu*; Furukawa, Tomohiro; Ito, Chikara; Kagota, Eiichi; et al.

Journal of Nuclear Materials, 386-388, p.294 - 298, 2009/04

 Times Cited Count:29 Percentile:89.01(Materials Science, Multidisciplinary)

In order to examine irradiation effect on creep rupture strength of Oxide Dispersion Strengthened (ODS) steel claddings, the in-pile creep rupture test was conducted using Material Testing Rig with Temperature Control (MARICO)-2 in the experimental fast reactor JOYO. Fourteen creep rupture events were successfully detected by the temperature change in each capsule and the $$gamma$$-ray spectrometry of the cover gas. Time to creep ruptures of six ODS steel specimens were identified by means of Laser Resonance Ionization Mass Spectrometry (RIMS), and no irradiation effect on creep rupture strength was confirmed within the irradiation condition in the MARICO-2 test.

Journal Articles

Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

Soga, Tomonori; Sekine, Takashi; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

Journal of Power and Energy Systems (Internet), 2(2), p.692 - 702, 2008/00

The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using Joyo. Two irradiation experiments were conducted in the MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing americium, MOX fuel containing americium and neptunium, and reference MOX fuel. The first test was conducted with high linear heat rates of 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and two test pins were replaced with dummy pins. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24 hours in August 2006 to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP. Post irradiation examination of these pins to confirm the irradiation behavior of MA-MOX fuel is underway.

Journal Articles

In-pile creep rupture experiment of ODS cladding materials in the experimental fast reactor Joyo

Ito, Chikara; Kagota, Eiichi; Ishida, Koichi; Kitamura, Ryoichi; Aoyama, Takafumi

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 9 Pages, 2007/04

no abstracts in English

Journal Articles

Research and development of minor actinide-containing fuel and target in a future integrated closed cycle system

Osaka, Masahiko; Serizawa, Hiroyuki; Kato, Masato; Nakajima, Kunihisa; Tachi, Yoshiaki; Kitamura, Ryoichi; Miwa, Shuhei; Iwai, Takashi; Tanaka, Kenya; Inoue, Masaki; et al.

Journal of Nuclear Science and Technology, 44(3), p.309 - 316, 2007/03

 Times Cited Count:25 Percentile:85.69(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Development of techniques for minor actinides transmutation using fast reactor; Irradiation tests for Am and Np containing fuel in experimental fast reactor JOYO

Soga, Tomonori; Sekine, Takashi; Takamatsu, Misao; Kitamura, Ryoichi; Aoyama, Takafumi

UTNL-R-0453, p.13_1 - 13_8, 2006/03

no abstracts in English

Journal Articles

Development of minor actinide containing fuel/target for the use in a future integrated system of fast reactor and accelerator driven system

Osaka, Masahiko; Serizawa, Hiroyuki*; Kato, Masato; Inoue, Masaki; NAKAJIMA, Kunihisa*; Tachi, Yoshiaki; Kitamura, Ryoichi; Oki, Shigeo; Miwa, Shuhei; Iwai, Takashi*; et al.

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

Development of minor actinide containing fuel/target, i.e., (Pu,Am)O$$_{2}$$-MgO, (Pu,Np)O$$_{2}$$-MgO, (U,Pu,Np)O$$_{2}$$, (U,Pu,Np)N and (Pu,Np,Zr)N, for the use in a future integrated system of fast reactor and accelerator driven system is underway as a collaborative work between JAERI and JNC. The present statuses of fabrication test and property measurements are given. Irradiation test in the experimental fast reactor JOYO is also mentioned.

JAEA Reports

Periodic safety review of the experimental fast reactor JOYO; Review of the activity for safety

Maeda, Yukimoto; Kashimura, Yoichi; Suzuki, Toshiaki; Isozaki, Kazunori; Hoshiba, Hideaki; Kitamura, Ryoichi; Nakano, Tomoyuki; Takamatsu, Misao; Sekine, Takashi

JNC TN9440 2005-001, 540 Pages, 2005/02

JNC-TN9440-2005-001.pdf:8.35MB

Periodic safety review (Review of the activity for safety) which consisted of "Comprehensive evaluation of operation experience" and "Incorporation of the latest technical knowledge" was carried out up to January 2005.

JAEA Reports

lmprovement of linear heat rate calculation for fast reactor MOX fueI using Monte Carlo code "MCNP"

; Kitamura, Ryoichi; Aoyama, Takafumi

JNC TN9400 2000-071, 36 Pages, 2000/07

JNC-TN9400-2000-071.pdf:1.27MB

A three dimensional, continuous energy Monte Carlo code "MCNP" was applied to accurately evaluate the linear heat rate of mixed oxide (MOX) fuel used in a fast reactor. The test fuel pins to be analyzed were irradiated at the core center position in the experimental fast reactor JOYO MK-II core as the power-to-melt test, which was performed in order to optimize the thermal analysis method employed in the MOX fuel pin design. ln the calculation, the heterogeneous structure of the irradiation test subassembly (B5D-2) which accommodated test fuel pins inside was modeled precisely. The neutron flux distribution within the subassembly was calculated using MCNP with the neutron source which was based on the diffusion theory by the JOYO MK-II core management code system "MAGI" The linear heat rates of fuel pins were then obtained by multiplying the fission rate and fission energy of individual fissile nuclide. The gamma heating was calculated by MAGl considering the delayed fission gamma and it was included in the neutron heating obtained by MCNP. The accuracy of MCNP was verified by the post irradiation examination. using the fission product yields ($$_{148}$$Nd) measured for the melted part of test fuel pins, the linear heat rates were evaluated and compared with the MCNP results. The average ratios of MCNP calculations to the experimental values were 0.955$$pm$$ 0.020, indicating MCNP can precisely calculate the linear heat rates. By correcting the MCNP calculation results with the C/E values obtained above, the linear heat rates of test fuel pins were evaluated to be 620 - 685 W/cm at the core center height.

JAEA Reports

Design of uninstrumented irradiation subassembly type-D (UNIS-D)

; Miyakawa, Shunichi; Mitsugi, Takeshi; Kitamura, Ryoichi

JNC TN9410 99-010, 350 Pages, 1999/06

JNC-TN9410-99-010.pdf:11.62MB

In the needs of the fuel irradiation test in "Joyo" MK-III core, there have been required that the irradiation of high performance fuel at high liner heat rate to high burn-up range, or the irradiation of advanced fuel such as MA fuel and Vipac fuel. In order to carry out these irradiation tests, newly designed irradiation subassembly is required with special features of; (1)Capability of the re-assembling after post-irradiation examination, even if the number of fuel in the identical irradiation condition decreases because of intermediate inspection. (2)Enhanced flexibility of the irradiation temperature setting ( in the present, UNIS-B's has 6 cases on the maximum). (3)Sufficient flexibility for the coolant flow distribution in the subassembly by extending variety of the flow rate setting. UNIS-D is a fuel irradiation subassembly which has been developed from above viewpoints. It is a compartment loading type irradiation subassembly that is able to load maximum of 18 compartments. Two types of compartments $$gamma$$ -type and $$delta$$ -typc arc prepared for UNIS-D. Thc sufficient consideration has also been made on the rc-assembling. A $$gamma$$ -type is the same compartment as the existing UNIS-B's and a $$delta$$ -type is the newly designed one for UNIS-D. Three to five fuel pins are loaded into a $$gamma$$ -type compartment and only one pin is loadcd into a $$delta$$ -type compartment. It is possible to carry out the irradiation test in a maximum of 18 test temperature conditions within a subassembly, since it has the sufficient flexibility for the coolant flow distribution. As for the development of UNIS-D, we have finished the detailed structure design and the design verification by the water flow examination, which confirmed that the UNIS-D exceeded its required performances in use and that its structure design was satisfactory.

JAEA Reports

Improvement of evaluation of irradiation test condition in JOYO

; Kitamura, Ryoichi; Aoyama, Takafumi

JNC TN9400 99-017, 26 Pages, 1999/03

JNC-TN9400-99-017.pdf:0.75MB

In order to improve the calculation accuracy of the irradiation test conditions in the experimental fast reactor JOYO, a three dimensional, continuous energy Monte Carlo code "MCNP" was employed. The fission product yields (148Nd) in the instrumented test assembly (INTA-2) which was loaded between the driver fuel subassemblies and reflectors were evaluated in this study. The results from the "MCNP" code were compared with the experimental values. In the calculation of the "MCNP" code, the neutron source distribution data which was calculated with the JOYO core management code system "MAGI" was used to evaluate the neutron flux and spectrum inside the INTA-2 subassembly. The $$^{148}$$Nd yields were evaluated with the calculated fission reactions and the fission yield data from JNDC-V2. The calculated $$^{148}$$Nd yields werc compared with the experimental values which were obtained from post irradiation examination results. The ratios of the "MCNP" values to experimental values of $$^{148}$$Nd yields are about 0.99$$sim$$1.00 of all pins. The "MCNP" code calculation values were within the range of the experimental error. These results indicate that the irradiation condition of the subassembly which was loaded next to the reflector can be evaluated accurately with the "MCNP" code.

Oral presentation

Upgrade of irradiation techniques in the experimental fast reactor "JOYO", 1; Traversable sample irradiation rig

Takamatsu, Misao; Tobita, Shigeharu; Sekine, Takashi; Kitamura, Ryoichi; Aoyama, Takafumi

no journal, , 

no abstracts in English

Oral presentation

Development of techniques for minor actinides transmutation using fast reactor, 1; Irradiation test for MA containing fuel in experimental fast reactor JOYO

Soga, Tomonori; Kitamura, Ryoichi; Abe, Kazuyuki; Koyama, Shinichi; Kato, Masato

no journal, , 

no abstracts in English

Oral presentation

Development of irradiation technology with temperature control

Kagota, Eiichi; Noguchi, Koichi; Kitamura, Ryoichi; Abe, Kazuyuki

no journal, , 

no abstracts in English

38 (Records 1-20 displayed on this page)