JAEA-Review 2020-006, 261 Pages, 2020/09
A literature review was conducted on the increase in surface area of vitrified products of HLW due to the fracturing caused by cooling during glass pouring process and by mechanical impact, from the perspective of a parameter of the radionuclide release model in the performance assessment of geological disposal system studied overseas. The review was focused on the value of surface area increase factor set as a parameter in the model, the experimental work to evaluate an increase in surface area, and how the parameters on surface area were determined based on the experimental results. The surface area obtained from the experiments executed in Japan was also discussed in comparison with the overseas studies. On the basis of the investigation, the effects of various conditions on the surface area were studied, such as a diameter of vitrified product, cooling condition during and after the glass pouring, impact on vitrified products during their handling, environment after the closure of disposal facility, and others. The causes of fracturing are associated with the phenomena or events in the waste management process such as production, transport, storage, and disposal. The surface area increase factors set in the nuclide release model of the glass and their bases were reviewed. In addition, the measured values and the experimental methods for surface increase factors published so far were compared. Accordingly, the methods for measuring surface area as the bases were identified for these factors set in the models. The causes of fracturing and features of these factors were studied with respect to the relation with the waste management process. The results from the review and assessment can contribute to the expanding the knowledge for the conservative and realistic application of these factors to performance assessment, and to the developing and upgrading of safety case as a consequence.
Simonnet, M.; Barr, N.*; Drot, R.*; Le Naour, C.*; Sladkov, V.*; Delpech, S.*
Radiochimica Acta, 107(4), p.289 - 297, 2019/04
Wan, T.; Saito, Shigeru
Metals, 8(8), p.627_1 - 627_22, 2018/08
Sawaguchi, Takuma; Tsukada, Manabu; Yamaguchi, Tetsuji; Mukai, Masayuki
Clay Minerals, 51(2), p.267 - 278, 2016/05
The dependences of the dissolution rate of compacted montmorillonite on activity of OH (a-) and temperature (T) were investigated. The dissolution rate of montmorillonite () in compacted pure montmorillonite, which was formulized as = 10 (a-) e, was higher than that in the compacted sand-bentonite mixtures: = 3500 (a-) e. The difference can be explained by considering the decrease in a- in the mixtures accompanied by dissolution of accessory minerals such as quartz and chalcedony. The dissolution rate model developed for pure montmorillonite is expected to be applied to bentonite mixtures if quantification of the decrease in a- is achieved somehow.
Kitamura, Akira; Takase, Hiroyasu*
Journal of Nuclear Science and Technology, 53(1), p.1 - 18, 2016/01
Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Focusing especially on the effects of -radiation in safety assessment, this study has reviewed research into the effects of -radiation on the spent nuclear fuel, canisters and outside canisters.
Kitamura, Akira; Takase, Hiroyasu*; Metcalfe, R.*; Penfold, J.*
Journal of Nuclear Science and Technology, 53(1), p.19 - 33, 2016/01
Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of -radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.
Yamaguchi, Tetsuji; Sakamoto, Yoshifumi; Akai, Masanobu; Takazawa, Mayumi; Iida, Yoshihisa; Tanaka, Tadao; Nakayama, Shinichi
Physics and Chemistry of the Earth, 32(1-7), p.298 - 310, 2007/00
Dissolution rate of montmorillonite, diffusivity of hydroxide ion and permeability coefficient in compacted sand-bentonite mixtures were experimentally determined and formulated. A coupled mass-transport/chemical-reaction code was developed to predict variation in permeability of engineered bentonite barrier with alkaline fluid by using the formulae.
Haraga, Tomoko; Kameo, Yutaka; Nakashima, Mikio
Bunseki Kagaku, 55(1), p.51 - 54, 2006/01
A relatively large quantity of sample solutions have to be prepared for radiochemical analysis of solidified products yielded by plasma melting treatment of non-metallic radioactive wastes. In order to dissolve the solidified products sample rapidly, dissolution method with microwave heating devices was applied. In a conventional method only by external heating with various mixtures of acids (HNO, HF, HClO and HSO), a 0.1 g amount of the sample was dissolved with difficulty. However, applying the microwave assisted dissolution method, a 1 g amount of the sample was completely dissolved in a shorter time. Thereby the time for dissolution procedures was shortened less than a one-tenth. The present dissolution method was successfully applied to the blast furnace slag as a reference material to determine main elements with good precision.
Asakura, Toshihide; Kim, S.-Y.; Morita, Yasuji; Ozawa, Masaki*
Journal of Nuclear and Radiochemical Sciences, 6(3), p.267 - 269, 2005/12
An electrolytic extraction (EE) method, i.e. electro-reductive deposition, of Tc from nitric acid aqueous solution was studied for future reprocessing. After 30 min of constant potential electrolysis by carbon electrode at -0.3 V vs. SSE (Standard Silver Electrode), Tc concentration in 3 M nitric acid decreased to 93 % of the initial value, which corresponds to 7 % of deposition. With co-existence of Pd, the value reached to 15 % of deposition equivalent by electrolysis at 0.0 V vs. SSE for 60 min. An acceleration effect of Pd on Tc deposition (promoter effect) was suggested. The concentration, however, increased to the initial value after further electrolysis and competing re-dissolution of deposited Tc was also suggested. In cyclic voltammetry measurements, it was found that the deposit from Tc-Pd-Ru-Rh solution dissolved easier than that from Pd-Ru-Rh did. In electrolyzed Tc solution, an absorption peak at 482 nm was found. It can be attributed to the complex with nitrite anion, and the complex formation is proposed as one possible mechanism of Tc re-dissolution.
Kurata, Yuji; Futakawa, Masatoshi; Saito, Shigeru
Journal of Nuclear Materials, 343(1-3), p.333 - 340, 2005/08
In order to study effects of temperature and alloying elements on corrosion behavior in liquid Pb-Bi which will be used for an accelerator driven system (ADS), corrosion tests of various steels were conducted under static liquid Pb-Bi condition. The tests were performed in oxygen-saturated liquid Pb-Bi at 450C and 550C for 3000h. Oxide films were formed during corrosion at 450C and 550C. Corrosion depth of steels decreased at 450C with increasing Cr content of steels. Austenitic stainless steels containing Ni didn't exhibit appreciable dissolution of Ni and Cr at 450C. The thick ferrite layer produced by dissolution of Ni and Cr was found in JPCA and type 316ss at 550C. For this reason the corrosion depth of austenitic stainless steels, JPCA and type 316ss became large. A Si-added austenitic stainless steel showed good corrosion resistance at 550C because a protective oxide film formed on the steel prevented dissolution of Ni and Cr into liquid Pb-Bi.
Kokusen, Junya; Seki, Masakazu; Abe, Masayuki; Nakazaki, Masato; Kida, Takashi; Umeda, Miki; Kihara, Takehiro; Sugikawa, Susumu
JAERI-Tech 2005-004, 53 Pages, 2005/03
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).
Shirai, Osamu; Kato, Tetsuya*; Iwai, Takashi; Arai, Yasuo; Yamashita, Toshiyuki
Journal of Physics and Chemistry of Solids, 66(2-4), p.456 - 460, 2005/02
Electrochemical behaviors of PuN and (U, Pu)N in the LiCl-KCl eutectic melt containing UCl and PuCl at 773 K were investigated by cyclic voltammetry. The electrochemical dissolution of PuN and (U, Pu)N began nearly at -1.0 V vs. the Ag/AgCl reference electrode. The rest potentials of PuN and (U, Pu)N were observed at about 0.15 V more negative potential than that of UN since the equilibrium potential of UN is about 0.15 V more positive than that of PuN. In the cyclic voltammogram measured by using (U, Pu)N as the working electrode, a steep rise of the positive current was observed at more positive potential than -0.4 V in analogy with the cyclic voltammogram measured by using UN as the working electrode. In addition, there were two anodic current waves in the voltammogram with (U, Pu)N, though the wave form was not clear. This indicates that UN and PuN would be dissolved independently irrespective of formation of the solid solution, (U, Pu)N.
Mineo, Hideaki; Isogai, Hikaru; Morita, Yasuji; Uchiyama, Gunzo*
Journal of Nuclear Science and Technology, 41(2), p.126 - 134, 2004/02
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.
Kurata, Yuji; Kikuchi, Kenji; Saito, Shigeru; Futakawa, Masatoshi; Sasa, Toshinobu
FZKA-6876, p.190 - 198, 2003/12
A report at MEGAPIE(Megawatt Pilot Experiment) Technical R & D Meeting is collected into an FZK report. According to the static corrosion tests, Al surface-treated layer produced by the gas diffusion method exhabited corrosion resistance to liquid Pb-Bi, while Al surface-treated layer produced by the melt dipping method suffered a severe corrosion attack. Furtheremore, it was found that a thick ferrite layer was formed in the surface of austenitic stainless steel at 550C. Dissolution at high temperature parts, precipitation of Fe-Cr alloy and deposition of PbO at low temperature parts occured in the first loop corrosion test. These caused plugging of the narrow passage in electro-magnetic pumu(EMP) system. Adoption of filters and a wide passage in an EMP system, and the use of an inner-polished tube specimen brought about a good effect.
Matsumoto, Shiro*; Uchiyama, Gunzo; Ozawa, Masaki*; Kobayashi, Y.*; Shirato, K.*
Radiochemistry, 45(3), p.219 - 224, 2003/05
no abstracts in English
Mineo, Hideaki; Goto, Minoru; Iizuka, Masaru*; Fujisaki, Susumu; Hagiya, Hiromichi*; Uchiyama, Gunzo
Separation Science and Technology, 38(9), p.1981 - 2001, 2003/05
no abstracts in English
Umeda, Miki; Nakazaki, Masato; Kida, Takashi; Sato, Kenji; Kato, Tadahito; Kihara, Takehiro; Sugikawa, Susumu
JAERI-Tech 2003-024, 23 Pages, 2003/03
MOX dissolution with silver mediated electrolytic oxidation method is planned for the preparation of plutonium nitrate solution to be used for criticality safety experiments at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). Silver mediated electrolytic oxidation method uses the strong oxidisation ability of Ag(II) ion. This method is thought to be effective for the dissolution of MOX, which is difficult to be dissolved with nitric acid.In this paper, the results of experiments on dissolution with 100 g of MOX are described. It was confirmed by the results that the MOX powder to be used at NUCEF was completely dissolved by silver mediated electrolytic oxidation method and that Pu(VI) ion in the obtained solution was reduced to tetravalent by means of NO purging.
Kida, Takashi; Umeda, Miki; Sugikawa, Susumu
JAERI-Data/Code 2003-001, 29 Pages, 2003/03
MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in NUCEF. A simulation code for the MOX dissolution has been developed for the operating support. In this report an outline of the simulation code is proposed and a comparison with the experimental data and a parameter study on the MOX dissolution rate are described.The principle of this code is based on Zundelevich's model for PuO dissolution using Ag. The influence of nitrous acid on the material balance of Ag and the surface area of MOX powder on the basis of particle size distribution are taken into consideration in this model. A comparison with experimental data was carried out to confirm a validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using the appropriate MOX dissolution rate constant. The parameters affecting the dissolution rate were studied, it was found that MOX particle size was major governing factor on the dissolution rate.
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 particles simulating spent fuel burned up to 45,000MWdt, 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.
Ogawa, Hiroaki*; Kiuchi, Kiyoshi
JAERI-Research 2002-037, 48 Pages, 2002/12
The difference in hydrogen permeation among candidate cladding materials such as 25Cr-35Ni stainless steel, Nb liner and reference materials such as 18Cr-8Ni SS, and Zr of Zircaloy base metal were evaluated by low energy plasma permeation simulated to hydrogen excited by heavy neutron irradiation. RF excitation source was arranged for the experimental apparatus in cooperating with temperature and bias control. Comparing with the thermodynamic gas driven permeation (GDP) in the same hydrogen pressure, the hydrogen permeation rate by the plasma driven permeation (PDP) was markedly accelerated at low to medium temperature range. The temperature dependency showed a knick at around 530K due to hydrogen-defect interactions. Comparing with Zr, Nb showed the high hydrogen solubility without the degradation by hydrate formation that is required to a getter material. The difference in PDP among candidates was analyzed with a new dissolution model for hydrogen.