Release of boron and cesium or uranium from simulated borosilicate waste glasses through a compacted Ca-bentonite layer

Chun, Kwan Sik; Kim, S.S; Kang, Chul Hyung

doi:10.1016/s0022-3115(01)00612-2

Cited by 13 publications

(7 citation statements)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The uranium concentration in the bentonite surface that was in contact with the spent fuel increased a little as shown in Table 3. However, the apparently yellow-colored uranium silicate compound layer reported, [15] which was formed at the interface between a bentonite block and the simulated borosilicate waste glass a 20% of uranium from the 6-months dynamic leach test, was not found in this study.…”

Section: Resultscontrasting

confidence: 64%

Comparison of the Amount of Nuclides Released from the Spent Fuel in Contact with and without a Compacted Bentonite Block

Kim

Choi

Kang

et al. 2006

Journal of Environmental Science and Health, Part A

View full text Add to dashboard Cite

A spent LWR fuel specimen between Ca-bentonite blocks was leached in a simulated bentonite-saturated granitic water (SBGW) for 165 days and its results were compared with those of a specimen leached without a bentonite block. The amounts of Cs, Sb, Sr, Eu, Am, U and Pu released from a 4.3 mm thickness of a fuel pellet with a 50,400 MWD/MTU burn-up in the SBGW without a bentonite block were 2.2, 0.25, 0.15, 0.02, 0.01, 0.01 and approximately 5 x 10(-4)% of their inventories, respectively. However, the amounts of nuclides released from the specimen between the 1.4 Mg/m(3) bentonite blocks were decreased by three times at least. Moreover, the concentrations of the nuclides in the leachate were very low because most of them were retained in the bentonite blocks.

show abstract

Section: Resultscontrasting

confidence: 64%

Comparison of the Amount of Nuclides Released from the Spent Fuel in Contact with and without a Compacted Bentonite Block

Kim

Choi

Kang

et al. 2006

Journal of Environmental Science and Health, Part A

View full text Add to dashboard Cite

show abstract

“…Alkali borosilicate glasses have potential application in the nuclear industry as suitable materials for the immobilization of high-activity nuclear waste [1,2]. Structural characterization of these glasses is essential for understanding glass durability.…”

Section: Introductionmentioning

confidence: 99%

Structure study of multi-component borosilicate glasses from high-Q neutron diffraction measurement and RMC modeling

Fábián

Sváb

Proffen

et al. 2008

Journal of Non-Crystalline Solids

View full text Add to dashboard Cite

“…Multi-component alkali borosilicate glasses are generally accepted as suitable materials for the immobilization of high-level atomic wastes [1]. Structural characterization of these host glasses is essential for understanding of glass durability.…”

Section: Introductionmentioning

confidence: 99%

Network structure of multi-component sodium borosilicate glasses by neutron diffraction

Fábian

Sváb

Mészáros

et al. 2007

Journal of Non-Crystalline Solids

View full text Add to dashboard Cite

Release of boron and cesium or uranium from simulated borosilicate waste glasses through a compacted Ca-bentonite layer

Cited by 13 publications

References 2 publications

Comparison of the Amount of Nuclides Released from the Spent Fuel in Contact with and without a Compacted Bentonite Block

Comparison of the Amount of Nuclides Released from the Spent Fuel in Contact with and without a Compacted Bentonite Block

Structure study of multi-component borosilicate glasses from high-Q neutron diffraction measurement and RMC modeling

Network structure of multi-component sodium borosilicate glasses by neutron diffraction

Contact Info

Product

Resources

About