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JAEA Reports

Behavior of carbon-14 in the Tokai reprocessing plant

; ; ;

JNC-TN8410 2001-021, 33 Pages, 2001/09

JNC-TN8410-2001-021.pdf:4.37MB

Carbon-14 released from the nuclear facilities is an important radionuclide for the safety assessment, because it tends to accumulate in environment through food chain and has as a significant impact to personal dose. Carbon-14 has been monitored routinely as one of the main gaseous radionuclides exhausted from the Tokai Reprocessing Plant (TRP) since OCtober of 1991. Furthermore, behavior of carbon-14 in TRP has been investigated through the reprocessing operation and the literature survey. This report describes the result of investigation about the behavior of carbon-14 in TRP as followings. (1)Only a very small amount of carbon-14 in the fuel was liberated into the shear off-gas and most of it was liberated into the dissolver of-gass. Part of the carbon-14 was trapped at the caustic scrubber installed in the of-gas treatment process, and untrapped carbon-14 was released into the environment from the main stack. Amount of carbon-14 released from the main stack was about 4.1$$sim$$6.5GBq every ton of uranium reprocessed. (2)Carbon-14 trapped at the caustic scrubbers installed in the dissolver off-gas and in the vessel off-gas treatment process is transferred to the low active waste vessel. Amount of carbon-14 transferred to the low active waste vessel was about 5.4$$sim$$ 9.6GBq every ton of uranium reprocessed. (3)The total amount of carbon-14 input to TRP was summed up to about 11.9$$sim$$15.5 GBq every ton of uranium reprocessed considering the released amount from the main stack and the trapped amount in the off-gas treatment devices. The amount of nitrogen impurity in the initial fuel was calculated about 15$$sim$$22ppm of uranium metal based on the measured carbon-14. (4)The solution in the low active waste vesselis concentrated at the evaporator.Most of the carbon-14 in the solution was transferred into concentrated solution. (5)Tokai vitrification Demonstration Facility (TVF) started to operate in 1994. Since then, carbon-14 has been measured in the ...

JAEA Reports

None

*

JNC-TN1440 2000-007, 115 Pages, 2000/08

JNC-TN1440-2000-007.pdf:4.45MB

no abstracts in English

JAEA Reports

None

Oda, Yoshihiro; Funasaka, Hideyuki; *; Obara, Kenji*; *

JNC-TY8400 2000-002, 47 Pages, 2000/03

JNC-TY8400-2000-002.pdf:2.53MB

no abstracts in English

JAEA Reports

The development of mass balance estimation code; The development and the analyzed example with object type code(I)

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JNC-TN9400 2000-034, 48 Pages, 2000/03

JNC-TN9400-2000-034.pdf:1.56MB

The study and the development to put FBR (Fast Breeder Reactor) to practical use have been doing. So many kinds of technologies are investigated to construct nuclear fuel recycle received to the society. The most important aim of reprocessing has been to extract U and Pu from spent fuels effectively, but, now, the demands for reprocessing are many kinds on nuclear fuel recycle system's construction. These need to be accepted sufficiently. The system that consists of electrolysis, extraction, with molten salt and melting metal, volatilization and condensation using the difference of vapor pressure is suggested, because, differently from LWR (Light Water Reactor), FBR can use the low decontamination factor's fuel. When the engineering scale plant is designed, the dry reprocessing has unsolved problems(ex. process flow) because of less demonstrative scale plants of the dry reprocessing than ones of the wet reprocessing. So the analysis and the estimation of mass balance that is most fundamental in the dry reprocessing system's design need to keep up with the system's alteration (to add new processes etc.) flexibly. This study aim is to develop the mass balance estimation code of dry reprocessing that satisfies the demand mentioned above.

JAEA Reports

None

*; Fujiwara, Masayuki*

JNC-TJ8430 2000-001, 55 Pages, 2000/03

JNC-TJ8430-2000-001.pdf:4.82MB

no abstracts in English

JAEA Reports

None

*; *; *

JNC-TJ8420 2000-003, 99 Pages, 2000/03

JNC-TJ8420-2000-003.pdf:5.47MB

no abstracts in English

JAEA Reports

Study on solubility of transuranium elements, II

Moriyama, Hirotake*

JNC-TJ8400 2000-050, 47 Pages, 2000/03

JNC-TJ8400-2000-050.pdf:1.49MB

In support of the safety assessment of geologic disposal of high levcl radioactive wastes, the solubility of transuranium elements was studied. The solubility of PuO$$_{2}$$$$cdot$$xH$$_{2}$$O was measured undcr a reducing condition, and the solubility product K$$^{0}_{sp}$$ and the stability constant $$beta$$$$_{4}$$ of Pu(OH)$$_{4}$$ were obtained. The obtained K$$^{0}_{sp}$$ value was found to be much smaller than that predicted by Rai et al. from its dependence on ionic radius. Also, the solubility of PuO$$_{3}$$3 $$cdot$$ xH$$_{2}$$O was measured under an oxidizing condition and the solubility product K$$^{0}_{sp}$$ was obtained. In the analysis of hydrolysis constants of actinide ions, it was found that the systematic trend of the hydrolysis constants was well explained by the hard sphere model considering the effective charges of actinide ions.

JAEA Reports

Analysis by fracture network modelling

WILLIAM S.DERSHO*; *

JNC-TJ8440 2000-001, 408 Pages, 2000/02

JNC-TJ8440-2000-001.pdf:21.62MB

None

JAEA Reports

Sorption studies of plutonium on geological materials - year 2

J. A. BERRY*; M. BROWNSWORD*; D. J. ILETT*; Linklater, C. M.*; Mason, C.*; TWEED, C. J.*

JNC-TJ8400 2000-060, 60 Pages, 2000/02

JNC-TJ8400-2000-060.pdf:2.95MB

Batch sorption experiments have been carried out to investigate the sorption behaviour of plutonium onto basalt and sandstone from the appropriate rock-equilibrated waters under different redox eonditions. Redox Potentials in solution were controlled by the addition of two reducing agents and one oxidising agent. Thermodynamic chemical modelling was undertaken to interpret the results. The sorption models were based on iron oxide. They adequately reproduced the data for sorption of plutonium onto sandstone, but tended to underpredict sorption onto basalt.

JAEA Reports

Influence of naturally-occurring heterogeneous complex-forming materials on the migration behavior of actinides in the geosphere (III)

Tochiyama, Osamu*

JNC-TJ8400 2000-044, 53 Pages, 2000/02

JNC-TJ8400-2000-044.pdf:1.41MB

To estimate the polyelectrolyte effect and the effect of the heterogeneous composition of humic acids, the complex formation constants of Eu(III) and Ca(II) with Aldrich humic acid and polyacrylic acid were obtained, for Eu(10$$^{-8}$$ to 10$$^{-5}$$ M) by solvent extraction with TTA and TBP in xylene, for Ca (10$$^{-10}$$M) with TTA and TOPO in cyclohexane and for Ca(10$$^{-4}$$M) by using ion-selective electrode. By defining the apparent formation as $$beta_{alpha}$$ = [MR$$_{m}$$]/([M][R]), where [R] denotes the concentration of dissociated functional group, [M] and [MR$$_{m}$$] denote the concentration of free and bound metal ion and pcH is defined as-log[H], the values of log$$beta_{alpha}$$ have been obtained at pcH 4.8 - 5.5 in 0.1 - 1.0M NaClO$$_{4}$$ and NaCl. Log$$beta_{alpha}$$ of Eu-humate varied from 5.0 to 9.3 and that of Ca-humate from 2.0 to 3.4..For both humate and polyacrylate, log$$beta_{alpha}$$ increased with pcH or with the degree of dissociation. The increase in the ionic strength O.1 to 1.0 M decreased the log$$beta_{alpha}$$, the decrease in log$$beta_{alpha}$$ of Eu(III)-humate is 1.6, that of Eu(III), polyacrylate 0.7, that of Ca(II)-humate 1.9 and that of Ca(II)-polyacrylate 1.2. While the increase in the metal ion produced no effect on log$$beta_{alpha}$$ of polyacrylate, log$$beta_{alpha}$$ of humate decreased. Depending on the concentration of Eu(III), the coexistence of Ca(II) reduced log $$beta_{alpha}$$ of humate by 0 to 0.8. The dependence of log$$beta_{alpha}$$ of humate on the metal ion concentration suggests the coexistence of strong and weak binding sites in the hmnic acid.

JAEA Reports

None

PNC-TN1000 98-004, 21 Pages, 1998/07

PNC-TN1000-98-004.pdf:0.86MB

no abstracts in English

JAEA Reports

None

PNC-TJ1632 98-001, 112 Pages, 1998/03

PNC-TJ1632-98-001.pdf:2.55MB

no abstracts in English

JAEA Reports

None

; ; ; ; *; ;

PNC-TN8410 97-192, 60 Pages, 1997/09

PNC-TN8410-97-192.pdf:6.04MB

None

JAEA Reports

None

*

PNC-TJ8211 97-002, 145 Pages, 1997/03

PNC-TJ8211-97-002.pdf:8.95MB

no abstracts in English

JAEA Reports

None

Nojiri, Ichiro; *

PNC-TN8410 96-398, 91 Pages, 1996/12

PNC-TN8410-96-398.pdf:6.09MB

None

JAEA Reports

None

Ohara, Hiroshi*; *

PNC-TJ8164 96-009, 261 Pages, 1996/09

PNC-TJ8164-96-009.pdf:13.68MB

no abstracts in English

JAEA Reports

None

Kato, Masato; ;

PNC-TN8410 96-247, 97 Pages, 1996/08

PNC-TN8410-96-247.pdf:40.8MB

None

JAEA Reports

Investigation of pyrometallurgical partitioning and extracting technology of irradiated fuel

Yumoto, Ryozo*; Yokochi, Yoji*; Koizumi, Masumichi*; Seki, Sadao*

PNC-TJ9409 96-002, 93 Pages, 1996/03

PNC-TJ9409-96-002.pdf:2.64MB

The state of development of pyrometallurgical partitioning and extracting technology of irradiated fuel is investigated. Also in case of perfoming the test at O-arai engineering center, the contents of the test, equipments, structure and arragement of cells that equipments are installed, are studied. The purpose of the test is to confirm the realization of the process and behavior of FP and TRU elements, and off-gass that cannot be made dear by cold test. In this study it is assumed that $$sim$$100g monju fuel (94,000MWd/t B.U, cooled for 550 days) per batch is treated. Four processes are picked up except for pin sectioning and powdering, as important subjects. They are as follows. (1)reduction of oxide fuel (2)electrorefining (3)cathode processing (4)extraction of TRU elements. And the outline of the test, blocked flow chart and the outline of equipment are clarified. And the outline of chart is drawn. Moreover, the specification of analitical equipments which are necessary to analyze the product is shown. From spent chloride, TRU and a part of FP elements are extracted and they are recycled for electrorefining and so on. The salt-waste including residual FP elements is kept in a receptacle after being absorbed into Zeorite and soldified. As the disposition of these tests, modified test cell in the existing FMF, modified concrete cell in AGF, new cell at B2F in the existing FMF and new cell at second auxiliary room in FMF extension are studied. As result of considering the disposition for equipment, the difficulty of reconstucting new cell including of equipments, method of mentenance, and equipments of ventilazion (Ar circumstance) including of management of off gas, and the plan of disposition, it is concluded that constructing iron cell into the second auxiliary room of FMF extension is best, because it is easy to construct safely, and the occurance of radioactive waste and the influence to other tests is little, and it is possible to examine more efficiently.

JAEA Reports

None

*

PNC-TJ8625 96-001, 28 Pages, 1996/03

PNC-TJ8625-96-001.pdf:0.93MB

no abstracts in English

JAEA Reports

A Study of direct disposal technology in the world (2)

*; *

PNC-TJ9222 95-002, 111 Pages, 1995/03

PNC-TJ9222-95-002.pdf:3.33MB

There are two methods of handling the spent fuel generated from the light water reactor; they are (l)direct disposal and (2)reprocessing-plutonium recycling. At present, Japan is following the line of "Reprocessing-Plutonium Recycling," but in the rest of the world, the movement for reviewing the Plutonium recycling is spreading, and in the future, the world opinion and pressure from overseas countries will increase against this method. Under these circumstances, Japan must compare the two methods to clarify the meaning of plutonium recycling. In this investigation, an overseas document by which the spent fuel had directly examined disposal was investigated. And, the content of those documents was arranged. The case of which directly disposed in Japan was set and the basic specification and the cost were evaluated. As a result of the investigation, the disposal cost became 54,900,000 yen/tU in the case with our country. This evaluation value is about 25% higher than Sweden and Finland where the cost is the highest in an overseas case. In cost items, the ratio which the article expense occupies is high. Moreover, the cost of construction and the close of underground facilities occupies the entire half for our country. This investigation is an evaluation based on in the case of the evaluation the current. Therefore, I want you to note going as for a technical detailed examination. However, the guess of the cost when directly disposing in Japan pounded. Moreover, the nuclear material control of the spent fuel is not evaluated. I want you to note cannot the comparison for that with the disposal of the glass solidification body.

64 (Records 1-20 displayed on this page)