Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 58

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Chlorination of UO$$_{2}$$ and (U,Zr)O$$_{2}$$ solid solution using MoCl$$_{5}$$

Sato, Takumi; Shibata, Hiroki; Hayashi, Hirokazu; Takano, Masahide; Kurata, Masaki

Journal of Nuclear Science and Technology, 52(10), p.1253 - 1258, 2015/10

 Times Cited Count:3 Percentile:65.3(Nuclear Science & Technology)

In order to explore the applicability of the chlorination by MoCl$$_{5}$$ as a potential pretreatment technique for waste treatment of fuel debris by pyrochemical methods, chlorination experiments of UO$$_{2}$$ and (U$$_{0.5}$$Zr$$_{0.5}$$)O$$_{2}$$ simulated fuel debris were carried out in two steps: the first one is a chlorination reaction by homogeneous heating, the second one is a volatilization of molybdenum by-product by heating under temperature gradient condition. Most of UO$$_{2}$$ and (U$$_{0.5}$$Zr$$_{0.5}$$)O$$_{2}$$ powder were converted to UCl$$_{4}$$ or UCl$$_{4}$$ and ZrCl$$_{4}$$ mixture at 573 K, respectively. In the case of (U$$_{0.5}$$Zr$$_{0.5}$$)O$$_{2}$$sintered particle, most of sample was converted to the chlorides because the products evaporated and be separated from sample surface at 773 K, while only the surface of the sample disk was converted to the chlorides at 573 and 673 K. Most of molybdenum by-product and ZrCl$$_{4}$$ were separated from UCl$$_{4}$$ by volatilization at 573 K.

JAEA Reports

Essentials of neutron multiplicity counting mathematics; An Example of U-Pu mixed dioxide

Hosoma, Takashi

JAEA-Research 2015-009, 162 Pages, 2015/08

JAEA-Research-2015-009.pdf:22.3MB

Neutron coincidence counting assay systems have been developed in the last two decades. Objects would extend to high-mass uranium-plutonium dioxide containing other spontaneous fission nuclei, so essentials of neutron multiplicity counting were reconsidered and expanded: (a) Formulae of multiplicity distribution were algebraically derived up to septuplet using a probability generating function; (b) Leakage multiplication was evaluated not by Monte Carlo method but by an average length from an arbitrary point inside a sample to an arbitrary point on its surface and a probability of induced fission within the length; (c) Mechanism of coincidence counting was associated with a couple of different time axes in Poisson process, and consequently a pair of close-to-coincident neutrons from the process was derived. For the formulae, new expressions using combination were wrote down. For spectrum and mean free path, actually treated uranium-plutonium dioxide was selected as an example.

JAEA Reports

Report on the fuel treatment facility operation

Kokusen, Junya; Seki, Masakazu; Abe, Masayuki; Nakazaki, Masato; Kida, Takashi; Umeda, Miki; Kihara, Takehiro; Sugikawa, Susumu

JAERI-Tech 2005-004, 53 Pages, 2005/03

JAERI-Tech-2005-004.pdf:5.92MB

This report presents operating records of dissolution of uranium dioxide and concentration of uranyl nitrate solution and acid removal, which have been performed from 1994 through 2003, for the purpose of feeding 10% and 6% enriched uranyl nitrate solution fuel to Static Experimental Critical Facility(STACY) and Transient Experimental Critical Facility(TRACY) in Nuclear Fuel Safety Engineering Facility(NUCEF).

Journal Articles

An Investigation into dissolution rate of spent nuclear fuel in aqueous reprocessing

Mineo, Hideaki; Isogai, Hikaru; Morita, Yasuji; Uchiyama, Gunzo*

Journal of Nuclear Science and Technology, 41(2), p.126 - 134, 2004/02

 Times Cited Count:5 Percentile:62.46(Nuclear Science & Technology)

A simple equation was proposed for the dissolution rate of spent LWR fuel, of which the change in the dissolution area was estimated by taking into account of the area of the cracks occurring due to thermal shrinkage of the pellets during irradiation. The applicability of proposed equation was examined using LWR fuel dissolution test results in the present study as well as the results obtained by other workers. The equation showed good agreements with the dissolution test results obtained from spent fuel pellets and pulverized spent fuel. It was indicated that the proposed equation was simple and would be useful for the prediction of dissolution of spent LWR fuels. However, the initial effective dissolution area, the parameter of the equation, was found to depend on the temperature, which could not be explained by the proposed equation. Further studies on the role of other factors affecting dissolution rate, such as nitrous acid, in the dissolution of spent fuel was required.

Journal Articles

Investigation of vaporization behavior of hyper-stoichiometric uranium dioxide by Knudsen effusion mass spectrometry

Nakajima, Kunihisa; Arai, Yasuo

Journal of Nuclear Materials, 317(2-3), p.243 - 251, 2003/05

 Times Cited Count:3 Percentile:73(Materials Science, Multidisciplinary)

The Knudsen effusion mass-spectrometric measurement of pure UO$$_{2+x}$$(s) are carried out at 1673, 1773 and 1873K to evaluate $$Delta$$G$$_{f}$$$$^{circ}$$(UO$$_{3}$$,g) as well as to measure the partial pressures of UO$$_{3}$$(g) and O$$_{2}$$(g) over UO$$_{2+x}$$(s) as function of the O/U ratio. It was found that the partial pressures of O$$_{2}$$(g) over UO$$_{2+x}$$(s) almost agree with the experimental data reported in the past and the values derived from the empirical equation given by Nakamura and Fujino. Further, it was found that the values of $$Delta$$G$$_{f}$$$$^{circ}$$(UO$$_{3}$$,g) obtained in this study are in good agreement with the recommended values.

Journal Articles

Behavior of simulated spent fuel in subcritical water

Mineo, Hideaki; Suzuki, Tadashi; Morita, Yasuji

Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.334 - 338, 2003/00

Behavior of spent nuclear fuel in subcritical water was investigated to look at the feasibility of fission-products (FPs) separation without organic solvent. The study employed unirradiated UO$$_{2}$$ particles simulating spent fuel burned up to 45,000MWdt$$^{-1}$$, which includes FP elements in oxide form: Sr, Zr, Mo, Ru, Rh, Pd, Ag, Ba, La, Ce, Pr, Nd and Sm. Also, alloy particles consisted of Mo, Ru, Rh and Pd were prepared to simulate the metallic phase of FP. 12.728 g of the fuel and 52 mg of the alloy were placed in a 10 ml pressure vessel, where subcritical water was fed. The temperature was 523, 573, 623 and 663K, while the pressure was kept at 29MPa. Dissolved fraction decreased with elevating temperature. It was found that more than 5% of Ba, Mo and Pr were respectively dissolved. The dissolved fraction of Sr and Rh were about 1%, and about 0.3% for Zr. La, Ce, Nd and Sm, indicated almost the same result as U, which was about 0.1%. It was suggested that the subcritical water could separate portion of FP. Further study would be carried out with smaller-sized fuel.

Journal Articles

Heat capacity measurements on unirradiated and irradiated fuel pellets

Amaya, M.*; Une, Katsumi*; Minato, Kazuo

Journal of Nuclear Materials, 294(1-2), p.1 - 7, 2001/04

 Times Cited Count:12 Percentile:33.08(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Oxygen potential and defect structure of the solid solution, Mg-Gd-UO$$_{2}$$

Fujino, Takeo*; Sato, Nobuaki*; Yamada, Kota*; Okazaki, Manabu*; Fukuda, Kosaku; Serizawa, Hiroyuki; Shiratori, Tetsuo*

Journal of Nuclear Materials, 289(3), p.270 - 280, 2001/03

 Times Cited Count:2 Percentile:79.29(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Thermal conductivities of irradiated UO$$_{2}$$ and (U,Gd)O$$_{2}$$

Minato, Kazuo; Shiratori, Tetsuo; Serizawa, Hiroyuki; Hayashi, Kimio; Une, Katsumi*; Nogita, Kazuhiro*; Hirai, Mutsumi*; Amaya, M.*

Journal of Nuclear Materials, 288(1), p.57 - 65, 2001/01

 Times Cited Count:19 Percentile:20.1(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Effects of grain size and PCI restraint on the rim structure formation of UO$$_{2}$$ fuels

Une, Katsumi*; Nogita, Kazuhiro*; Suzaka, Yojiro*; Hayashi, Kimio; Ito, Kunio*; Eto, Yoshinori*

International Topical Meeting on Light Water Reactor Fuel Performance, 2, p.775 - 785, 2000/00

no abstracts in English

Journal Articles

Radiation damage in UO$$_{2}$$ bombarded with iodine and nickel ions at energies around 100 MeV

Hayashi, Kimio; ; Fukuda, Kosaku

Advances in Science and Technology, 24, p.439 - 446, 1999/00

no abstracts in English

Journal Articles

Nonuniformity effect on reactivity of fuel in slurry

; *

Nuclear Technology, 122(3), p.265 - 275, 1998/00

 Times Cited Count:1 Percentile:84.27(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Radiation damage of UO$$_{2}$$ by high-energy heavy ions

; ; Fukuda, Kosaku

Journal of Nuclear Materials, 248, p.191 - 195, 1997/00

 Times Cited Count:16 Percentile:22.47(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Non-uniformity effect on reactivity of fuel in slurry

Okuno, Hiroshi; *

PHYSOR 96: Int. Conf. on the Physics of Reactors, 4, p.L74 - L82, 1996/00

no abstracts in English

Journal Articles

Solid-aqueous phase equilibria of uranium(VI) under ambient conditions

G.Meinrath*; ; Kimura, Takaumi; Yoshida, Zenko

Radiochimica Acta, 75(3), p.159 - 167, 1996/00

no abstracts in English

JAEA Reports

Calculations for reflector savings of steel and stainless steel

*; Okuno, Hiroshi

JAERI-Research 94-047, 39 Pages, 1994/12

JAERI-Research-94-047.pdf:0.89MB

no abstracts in English

JAEA Reports

Analyses with MCNP code of the criticality experiment of pellet-solution coexisting low-enriched uranium fuel

*; Arakawa, Takuya*; Okuno, Hiroshi

JAERI-Data/Code 94-018, 36 Pages, 1994/12

JAERI-Data-Code-94-018.pdf:1.09MB

no abstracts in English

JAEA Reports

58 (Records 1-20 displayed on this page)