Studies of Ceramic Fuels with the Use of Fission Gas Release Loop, (I)

Fission gas release from a UO²-graphite mixture was studied during irradiation with the use of the Fission Gas Release Loop in the JRR-3 reactor. The release rates of fission krypton and xenon increased proportionally with neutron flux (6×10¹⁰–6×10¹² n/cm²·sec) and exponentially with temperature (400°–1,000°C). A burst of fission gas was observed when the specimen was abruptly heated to a higher temperature. These results can be explained by a mechanism whereby fission gas is trapped in defects created in graphite by fission fragments and released through annealing of the defects.     

Studies of Ceramic Fuels with the Use of Fission Gas Release Loop, (II)

The γ-ray spectra of fission gases released from UO₂-graphite pellets under neutron irradiation were measured. With and without separating fission gases into xenon and krypton, 25 kinds of γ-ray were observed and assigned to nine nuclides, ^85mKr, ⁸⁷Kr, ⁸⁸Kr(⁸⁸Rb), ¹³³Xe, ¹³⁵Xe, ^135mXe and ¹³⁸Xe (¹³⁸Cs). A value of 15 min is proposed for the half-life of ¹³⁸Xe, based on analysis of the decay curves. Discussion is given on problems related to determination of the release rate of each fission product through measurement of the height of each peak in the γ-ray spectrum.     

Studies on Ceramic Fuels with the Use of Fission Gas Release Loop, (III)

Flux monitoring in the specimen chamber of FGRL was carried out by directly and continuously reading with a γ-ray spectrometer the content of ¹⁶N in the primary cooling water produced by ¹⁶O(n,p)¹⁶N reaction within the cooling jacket of FGRL. 16N generation is not influenced by γ-ray build-up in the reactor nor by the temperature in the specimen chamber. The detector can be simply set outside the pipe through which the cooling water flows, because the high energy γ-rays (6.13 MeV) emitted during the ¹⁶N decay easily penetrate the pipe wall.     

Effects of Temperature Cycling and Heating Rate on Fission Gas Release of BWR Fuels

The effects of temperature cycling and heating rate on the release behavior of ⁸⁵Kr have been studied for U0₂ pellets irradiated in a commercial BWR during 3 and 4 cycles (burn-up: 23 and 28GWd/t), by using a post irradiation annealing technique. In addition, characteristics of intergranular bubbles in base-irradiated and annealed specimens (burn-up: 6~28GWd/t) have been examined by SEM fractography.

No significant difference in the release of ⁸⁵Kr was observed between the cyclic heating from 700 to 1,400°C and isothermal heating at 1,400°C. The maximum release rate of ⁸⁵Kr during heating up to 1,800°C became lower with decreasing heating rate in the range of 0.03–10°C/s, while its cumulative fractional releases were about 20~30%, almost independent of heating rate. The fractional coverage of the grain face area occupied by intergranular bubbles saturated around 40~50 for the specimens annealed at 1,600-1,800°C, independent of specimen burn-up and heating conditions (temperature, heating rate and duration). A relationship between intergranular bubble concentration N_g per unit area of grain face and average bubble diameter d_g was expressed as N_g∝d_{g 2.1}     

In-Situ Measurement of Fission Product Plateout in In-Pile Gas Loop Using Portable Ge(Li) Gamma-Ray Spectrometer

A nondestructive and quantitative method of measuring fission products deposited on inside surface of the primary duct has been developed. A portable Ge(Li) detector with a lead collimator was used for the external scanning of γ-rays emitted from the fission product plateout in the duct.

Upon termination of the first irradiation cycle for the coated particle fuels in the high temperature in-pile gas loop OGL-1 installed in JMTR (Japan Material Testing Reactor), in-situ measurements of the plateout were carried out by the above method at different points in the primary system.

Identified nuclides in the measured γ-ray spectra were ¹³¹I, ¹³³I, ⁹⁵Zr, ⁹⁵Nb, ¹³⁷Cs, ^110mAg, ⁹⁹Mo, ¹⁴⁰La, ⁶⁰Co, ⁵⁸Co, ⁵⁴Mn, ⁵¹Cr, ⁶⁵Zn, ⁵⁹Fe, ¹²²Sb, ¹²⁴Sb, ¹⁸²Ta, ¹⁸¹Hf, ²²Na and ⁴⁶Sc. The plateout densities of these nuclides were obtained using conversion coefficients determined by detector calibration and calculation, which were from 10^-5 to 10^-2 μCi/cm².

It was shown clearly by the experiment that the present measuring method using an in-situ γ-ray spectrometer is useful in the studies of fission product plateout in the primary cooling system of a high temperature gas cooled reactor.     

Design and Operational Features of the Low Temperature Fissiochemical Loop

A low temperature in-pile loop for the irradiation of chemical reactants, especially by fission fragments, has been designed, constructed and operated successfully. The main features of the loop are briefly described: the capsule for sample irradiation, the design and performance of the helium cooling system and safety considerations. Gaseous chemical reactants charged in an instrumented double capsule at pressures up to 20 kg/cm², can be irradiated in a nuclear reactor at any temperature between + 20° and – 190°C, for periods up to 5hr. The thermal performance of the helium cooling system was in good agreement with the design calculations. Irradiations of ethylene and other gases could be conducted at an absorbed fission fragment dose rate of about 20 Mrad/hr, which is several times larger than the background reactor radiation. Thus, chemical reaction studies by fission fragments can be safely conducted at. ambient to low temperatures in this loop.     

The Application of an Advanced Ion Exchange Process to Reprocessing Spent Nuclear Fuels, (I)

In order to develop an advanced ion exchange process for the reprocessing of spent nuclear fuels, a novel anion exchanger, AR-01 with the resin embedded in porous silica beads and benzimidazoles as functional groups has been manufactured. Adsorption behavior of various fission product elements (FPs) and uranium in nitric acid medium were investigated experimentally using this anion exchanger. Separation performance of FPs from U(VI) in simulated spent fuel solutions was demonstrated by column chromatography utilizing dilute HNO₃ and thiourea as eluents.

Most FPs such as Cs(I), Sr(II), Mo(VI), Rh(III) and trivalent rare earths showed negligibly slight adsorption and could be separated from U(VI) satisfactorily. Cerium(IV) was strongly adsorbed, but was gradually reduced to non-adsorptive Ce(III) by the anion exchanger. Zirconium(IV) presented weak adsorption and its a part mixed with U(VI) in the column experiments. Ruthenium(III) exhibited quite strong adsorption in a broad HNO₃ concentration range as the form of anionic nitrosylnitrato-complexes, its most amount mixed with U(VI). Palladium(II) showed significantly strong adsorption probably due to complexes formation with the anion exchanger. The adsorbed Pd(II) was effectively eluted out by thiourea and separated from U(VI) and other FPs completely.     

Low-Temperature Partitioning of Plutonium with High U(VI)-Loading for Fuel Reprocessing

Measurement of the distribution ratios of Pu(IV), U(VI) and HNO₃ at low temperatures and its treatment with DIST code revealed that a high U (VI)-loading of 30% TBP in n-dodecane splits Pu(IV) down to the aqueous phase more strongly than do at 25°C. Based on these findings, flowsheet conditions to separate Pu(IV) from U(VI) were investigated with EXTRA.M code including the distribution equations obtained above. And tentative flowsheets for non-reductive Pu-splitting process at a temperature of 5°C were proposed for fuel reprocessing mainly based on the effects of U (VI)-loading in the solvent and temperature on distribution ratios of Pu(IV) and U(VI). Distribution ratios of the fission products, Zr, Nb, Ru and Ce were also measured to assess their decontamination from U or Pu products in the above process. Finally behavior of Np, in the proposed partitioning process was discussed by analysis with EXTRA. M code and a redox reaction model.