Corrosion tests with uranium- and plutonium-loaded ceramic waste forms.

Morss, Lester R.; Johnson, S. G.; Ebert, W. L.; DiSanto, T.; Frank, S. M.; Holly, J. L.; Kropf, A. Jeremy; Mertz, Carol J.; O’Holleran, T. P.; Richmann, Michael K.; Sinkler, W.; Tsai, Yifen; Warren, A. R.; Noy, M.

doi:10.2172/810461

Cited by 5 publications

(16 citation statements)

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“…IV-24b suggests that the two modes seen in light scattering results correspond to small (U,Pu)O 2 crystallites and the larger aluminosilicates. The maximum peak height of the smaller size distribution corresponds to a particle size of 8 μm and the maximum of the larger size distribution corresponds to a particle size of about 120 μm (Morss et al, 2002a;Morss et al, 2002b). The two size fractions correspond well with the sequential filtration analysis of the test solutions, the results of which were shown in Fig.…”

Section: Ivb5 Release Of U and Pu During Cwf Degradationsupporting

confidence: 61%

“…A series of tests was conducted with HIP CWF and PC CWF materials made with salts containing amounts of U and Pu that span the likely range of waste form compositions to study the release behaviors of these radionuclides as the CWF corrodes Morss et al, 2002a;Morss et al, 2002b). Four different U,Pu-loaded HIP CWF materials were made using zeolites with residual water contents of 0.12 or 3.5 mass % and salts doped with either 5.25 mass % U + 1.76 mass % Pu or with 1.76 mass % U + 5.30 mass % Pu.…”

Section: Ivb5 Release Of U and Pu During Cwf Degradationmentioning

confidence: 99%

“…Four different U,Pu-loaded HIP CWF materials were made using zeolites with residual water contents of 0.12 or 3.5 mass % and salts doped with either 5.25 mass % U + 1.76 mass % Pu or with 1.76 mass % U + 5.30 mass % Pu. These resulted in two CWF materials with 0.421 mass % U + 0.141 mass % Pu (CWF 237m and CWF 239m) and two with 0.141 mass % U + 0.426 mass % Pu (CWF 238m and CWF 240m) (Morss et al, 2002b, Table 20B). The letter "m" following the CWF identifier indicates that a mixture of crushed material from several products was used in the PCTs.…”

Section: Ivb5 Release Of U and Pu During Cwf Degradationmentioning

confidence: 99%

“…The near-edge spectrum is shown in Fig. IV-25 superimposed with the spectrum of a PuO 2 reference sample (Morss et al, 2002a;Morss et al, 2002b).…”

Section: Ivb5 Release Of U and Pu During Cwf Degradationmentioning

confidence: 99%

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Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Ebert¹

2006

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Section: Ivb5 Release Of U and Pu During Cwf Degradationsupporting

confidence: 61%

Section: Ivb5 Release Of U and Pu During Cwf Degradationmentioning

confidence: 99%

Section: Ivb5 Release Of U and Pu During Cwf Degradationmentioning

confidence: 99%

“…The near-edge spectrum is shown in Fig. IV-25 superimposed with the spectrum of a PuO 2 reference sample (Morss et al, 2002a;Morss et al, 2002b).…”

Section: Ivb5 Release Of U and Pu During Cwf Degradationmentioning

confidence: 99%

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Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Ebert¹

2006

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“…When the eutectic salt containing PuCl 3 is contacted with zeolite, even very dry zeolite used to produce the CWF, residual H 2 O in the zeolite structure reacts strongly with Pu to form PuO 2 . This phenomenon has been observed in past studies on the behavior of the CWF [28]. The top pattern in Fig.…”

Section: X-ray Analysissupporting

confidence: 77%

Plutonium-238 alpha-decay damage study of the ceramic waste form.

Frank

Barber

Cummings

et al. 2006

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Ceramic Solidification of Salt-containing Waste from Pyrochemical Reprocessing of Spent Nuclear Fuel

Chai¹,

Shi²

2019

Journal of Inorganic Materials

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For the future advanced nuclear fuel cycle system, pyrochemical technology based on molten salt electrolysis is generally considered to be one of the most promising and reliable reprocessing technologies. The salt-containing waste generated in each step of the pyrochemical process needs to be converted into a ceramic waste form, which can be stably disposed in a long term manner in deep geological repository. This is of pivotal importance for the scale-up and industrialization of molten salt electrolysis based pyrochemical processing. In this review, the current research progresses of ceramic solidification technology in main nuclear energy countries with respect to salt-containing wastes were summarized and reviewed, with emphasis on ceramic solidifications of salt-containing wastes from electro-reduction process in LiCl-based salt and electro-refining process in LiCl-KCl salt. In addition, future perspectives in this field are also given.

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Corrosion tests with uranium- and plutonium-loaded ceramic waste forms.

Cited by 5 publications

References 0 publications

Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Plutonium-238 alpha-decay damage study of the ceramic waste form.

Ceramic Solidification of Salt-containing Waste from Pyrochemical Reprocessing of Spent Nuclear Fuel

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