Fission xenon release from lightly irradiated (Th,U)O2 powders

The release of ¹³³Xe from (Th,U)O₂ was studied at a low fission density by using a post-irradiation technique. The uranium concentrations of the specimens ranged from 0.15 to 20 mol% U0₂. Heating curves of the release gave almost the same pattern in shape, while the total release, which was the combination of an in-pile release and a post-irradiation release up to 1000°C, increased with uranium concentration except for the nominally pure ThO₂. Effects of preparation conditions of specimens such as atmosphere, temperature and stoichiometry were also studied and found to be minor. Possible release mechanisms were discussed.     

Fission Gas Release from Nuclear Fuels, (II)

The release behavior of fission gases in U-metal, UO₂ and uranium carbides, irradiated at a relatively low temperature (below 100°C) to low dosage, was studied by out-of-pile experiments.

It was found that fission gas (¹³³Xe) released from a specimen by fission fragment recoil is mostly captured in the wall of the irradiating capsule or in the capsule support material.

The amount of fission gas released into the void space of the capsule is proportional to the surface area and to the fuel burn-up, and is controlled by a knock-out release mechanism. The number of U atoms considered to take part in the knock-out mechanism by evaporation or displacement due to the intrusion of a recoil fission fragment, is estimated to be 1.4×10⁵~2.7×10⁵ atoms for U-metal and 5×10⁴~10×10⁴ atoms for UO₂ and uranium carbides.     

Release Process of Fission Xenon Produced by β-Decay of Precursor Iodine in Natural Graphite at 1,000°C

In order to study the release process of fission xenon produced by decay of the precursor iodine in natural graphite powder, post-irradiation experiments were carried out at 1,000°C soon after 20 min irradiation with UO₂ powder. In an experiment with interruption (2 hr heating—10 day cooling—10 hr heating), it was found that the decay of ¹³³I to ¹³³Xe caused a fraction of the ¹³³Xe produced to be released rapidly, while the remaining fraction was released gradually. This was followed by a continuous heating experiment, during which the release rates of ¹³³Xe, ¹³⁵Xe and ^135mXe were measured, and in which production and release of xenon occurred simultaneously in a manner similar to conditions prevailing under actual irradiation experiments. The rates of nuclide release were explained by assuming the same release probability for each xenon nuclide as that for ¹³³Xe in the first experiment. The origin of the initial rapid release by decay of iodine to xenon was considered mainly to be the change of chemical interaction between the defects and the atoms trapped therein. The release behavior of xenon after its production was explained by assuming that the activation energies for the release from the defects were lowered by about 10kcal/mol by the decay.     

Diffusion of Fission-Produced Xenon in Uranium Monocarbide

The diffusion of Xe atoms in UC at temperature from 800° to 1,400°C was studied on samples irradiated to various doses up to 2.9×10¹⁸ nvt. The apparent activation energy of the diffusion was found to vary with fluence, which corresponded to the extent of neutron or fission fragment damage in the specimen.

The diffusion was observed to be enhanced at exposures below 10¹⁷ nvt. At the higher doses, on the other hand, trapping of the gas atoms by defect clusters became evident, which inhibited the gas release.

When the irradiated samples were once heated up to 1,400°C, the activation energy of Xe diffusion through them showed a constant value of 83±5 kcal/mol, which was independent of the neutron fluence.

The scattered values of activation energy reported in literature are explained from the present results as resulting from differences in the activation energy of Xe diffusion, which depends on the extent of radiation damage retained in the UC.     

Numerical Solution Method for Prediction of Heat Transfer in Tube Bundles of Large-Size Intermediate Heat Exchangers in FBR Plants

A study was made of uranium contamination in (a) the coating layers of TRISO particles (a-1) before compacting and (a-2) separated from once-compacted fuel heat-treated at 1,400 or 1,800°C, and (b) in the matrix material of the same compacts. The contamination in the pyrocarbon layers of the coating was determined, after mechanically separating the coating layers, by a procedure of neutron activation, burn-off and ¹³³Xe trapping. For the silicon carbide coating layer, the fragments of coating left from the above procedure were fused into alkaline melt, and the ¹³³Xe released at each heating step was trapped. For the matrix material, the fuel compacts were deconsolidated electrolytically or mechanically, followed by activation analysis. The results of the foregoing measurements proved the uranium contamination in pyrocarbon and silicon carbide coating to be at most of the order of 10^?4 in reference to uranium content in kernel, while the corresponding value for particles sampled from fuel compacts heat-treated at 1,800°C were appreciably higher. The corresponding values found for the matrix material were of the order of 10^?5.     

Transient Behavior of Minor Isotopes in Cascade for Uranium Enrichment

A numerical method is proposed for analyzing the transient state of a three-component cascade. The method is applied to an examination of the behavior of minor isotopes of uranium in an enrichment plant.

It is shown that the transient behavior of ²³⁴U in cascade differs significantly from that of ²³⁵U. It is indicated that in order to apply MIST (” Minor Isotope Data as a Safeguards Technique” proposed by Blumkin et al.) effectively, it would be necessary to examine the behavior of the isotopic ratio (²³⁵U/²³⁴U) at various transient states as well as at steady state.

The concentration of ²³⁶U does not necessarily distribute monotonically from the top toward the bottom of the cascade for the production of highly enriched uranium. It is shown how the characteristic distribution of ²³⁶U establishes itself gradually with lapse of time.     

Analysis of the Escape Behavior of Xenon Captured in Graphite by Nuclear Recoil, (II)

The escape of ¹³³Xe introduced into natural graphite by recoil from UO₂ particle was studied at temperatures ranging from 200° to 1,450°C and the resulting data correlated with the duration of isothermal post-irradiation heating. The escape data were used to verify the proposed escape mechanism, which postulates three successive processes—release from trapping site, diffusion of the released xenon, and desorption from surface.

The concept of initial activation energy spectrum, which determines the trapping states of xenon atoms before any annealing begins in terms of the activation energy for the release, was introduced to express the release behavior mathematically. The mathematical treatment for both the release and the diffusion processes permits expression of the whole escape process with a single equation, assuming no resistance to xenon transport in the desorption process. The validity of this equation was substantiated by observed data. It was made clear that rapid sweep of the spectrum is the cause of the initial burst stage in which the escape rate is dependent upon diffusion of the released xenon, and that the slow sweep of the spectrum causes the ensuing slackened process in which the release from the trapping sites constitutes the limiting factor for the rate of escape.     

The mechanisms of in-pile fission gas release from UO2

In-pile release mechanisms of fission gas from UO₂ at low temperatures were studied. The release of ¹³³Xe, ¹³⁵Xe, ¹³⁸Xe, ^85mKr, ⁸⁸Kr and ⁸⁷Kr from a sintered UO₂ pellet was measured at temperatures ranging from 250 to 930°C using a graphite specimen holder. The release from the holder, in which a fraction of fission gas was recoil-implanted, was subtracted to obtain the net release from the UO₂ pellet. Knock-out release from the UO₂ was measured directly, and it was found that it was not the main release mechanism, at least not for short-lived nuclides. A ‘pseudo-recoil’ release model is proposed to explain the low temperature release under irradiation. In the model, some of the defects produced by fission fragments act as short-lived carriers for fission gas.     

Dancoff Factor in Arrayed-Type Cladded Fuel Lattice

The initial-stage sintering mechanism of hyperstoichiometric urania prepared by sol-gel process was determined in relation to temperature during constant rate heating (CRH). The urania powder used in this experiment was prepared by crushing in Ar atmosphere the micro- spheric gel of UO₂ obtained by sol-gel process, and reducing the resulting powder by heating in H₂ for 1 hr at 500°C. The results obtained from densification measurements indicated that the initial-stage sintering proceeded in two phases governed by different shrinkage mechanisms, as follows.

1. The sintering up to 675°C would be due to a mechanism such as rearrangement of grains and/or plastic flow.

2. Sintering from 750° to 800°C was interpreted as being controlled by uranium volume diffusion.

The estimated diffusion coefficient D = 1.42×10^?6 exp(-52,500/RT) cm²/sec. This value agreed in order of magnitude with the uranium diffusion coefficients measured by other workers for hyperstoichiometric urania.     

Experimental Study on Fluoride Volatility Process for Thorium Fuels

Selective removal of uranium from (Th/U)O₂ by fluorination with fluorine was studied experimentally. The fluorination was performed both in a small boat and in a 2 inch inner diameter fluid-bed reactor.

Fuel particles tend to agglomerate in the reactor due to the large amount of reaction heat and the comparatively low melting point of ThF⁴. The fluorinated fuels produced in the fluid-bed reactor were found to be partially agglomerated. Fractional retention of uranium was smaller in the agglomerated parts than in the un-agglomerated, and smaller in the outer layers of the cakes than in the core.

On the other hand, it was also established beyond doubt from the results of the small boat fluorination experiment that heavy agglomeration inhibits the volatilization of uranium in the form of UF₆. Inhibition of the violet exothermic reaction by lowering the fluorine pressure in the early stage of fluorination was found to be a very effective method of obtaining high uranium recovery. It was demonstrated that more than 99% of the uranium could be volatilized within 4 to 5 hr at a temperature of 580°C.

The experimental results on the effects of temperature, particle size and fluorine concentration are presented. The variations of reaction rate observed in the course of fluorination are also discussed.     

Post-irradiation studies on knock-out and pseudo-recoil releases of fission products from fissioning UO2

By using post-irradiation techniques, in-pile releases of ¹³³Xe, ^85mKr, ⁸⁸Kr, ⁸⁷Kr and ¹³⁸Xe from UO₂ fissioning at low temperatures below about 200° C are studied: these are analyzed into a time-dependent knock-out and time-independent pseudo-recoil releases. For the latter, a “self knock-out” mechanism is proposed: when a fission fragment loses thoroughly its energy near the UO₂ surface and stops there, it will knock out the surface substances and accordingly the fragment (i.e. the fission product) will be released. The effective thickness of the layer where the self knock-out occurs is found to be ~7Å. As for the knock-out release, the following is estimated from its dependence on various factors: the knock-out release of fission products occurs from the surface layer with the effective thickness of ~20Å: the shape of UO₂ matrix knocked out by one fission fragment passing through the surface is equivalent to a cylinder ~32Å diameter by ~27Å thick, (i.e. the knock-out coefficient for UO₂ is ~660 uranium atoms per knock-out event). On the basis of the above estimations, the conclusions derived from the past in-pile studies of fission gas releases are evaluated.     

Neutron damage and fission product gas release in the initial burst region of uranium monocarbides

Thermal neutron damage and fission product gas (¹³³ Xe) release in a burst region of uranium monocarbides were studied. After neutron irradiation, the electrical resistivity was measured from room temperature to 800° C. Three recovery stages were revealed in the resistivity of UC irradiated to $\text{4.0 × 10}{}^{16}$ nvt. The lattice parameter of UC with the same irradiation also showed three stages of recovery up to 1050°C. The initial burst of Xe from UC was studied in a dose range between $\text{1.6 × 10}{}^{15}$ and $\text{2.9 × 10}{}^{18}$ nvt. The burst occurred in three steps for lightly irradiated specimens, while there were two steps of the burst in heavily irradiated specimens. The activation energies for each burst step were calculated. From the results obtained here, we concluded that the burst was correlated with the recovery of damage in the neutron-irradiated UC.     

A study of the generation of 232U in UO2 and MOX fuels

To clarify the generation pathway of ²³²U, an important nuclide for dose evaluation at various stages in the reuse of uranium, concentrations of ²³²U generated through various pathways were evaluated for UO₂ and mixed oxide (MOX) fuels. Burnup calculation was conducted with ORIGEN2.2 code adopting ORLIBJ40 library, a set of cross-section libraries based on JENDL-4.0. It was found that differences in ²³²U concentrations in UO₂ and MOX fuels mainly arise from differences in the initial compositions of ²³⁴U, ²³⁵U, and ²³⁶U. It was also found that the contribution of plutonium and americium isotopes in MOX fuels is small compared with that of uranium isotopes. The results clarified that the capture cross sections of ²³⁰Th, ²³¹Pa, ²³⁵U, and ²³⁶U, as well as the (n,2n) cross sections of ²³⁷Np and ²³⁸U, have a large effect on the generation of ²³²U. Additional investigation showed that ²³²U concentration is strongly affected not only by time after irradiation but also by time before irradiation.     

Diffusion and Evaporation of Fission Products in Coated Fuel Particles

Fractional releases of ¹³³Xe, ¹⁴⁰Ba and ⁸⁹Sr from slightly-irradiated pyrolytic-carbon-coated and SiC-coated particles were measured over a temperature range of 1,200°–1,750°C. The results are analyzed mathematically in order to obtain the diffusion and evaporation coefficients relevant to PyC and SiC. The resulting expressions for the coefficient of diffusion in PyC are 2.9x10-⁷ exp(-61x10³/RT) for ¹³³Xe and 4.7x10-² exp(51x10³/RT) for ¹⁴⁰Ba. For the coefficients of evaporation of ¹⁴⁰Ba from PyC, the expression is 3.5x10³ exp(-42x10³ /RT). As for SiC, the diffusion and evaporation coefficients of these nuclides are given for a temperature of 1,750°C. A high diffusivity path for the diffusion of ¹⁴⁰Ba is postulated to explain the difference in diffusion behavior between ¹³³Xe and ¹⁴⁰Ba in PyC.     

Absolute Fission-Rate Distributions in Lithium and Hybrid Fusion Blanket Assemblies, (II)

The absolute fission rates of ²³⁵U, ²³⁷Np, ²³⁸U and ²³²Th were measured in four types of spherical blanket assemblies containing lithium and/or natural uranium and/or graphite. The results of measurement are compared with those of one-dimensional transport calculations employing 100-group neutron cross-sections obtained from the ENDF/B-IV data file. It is shown that the ratios between calculated and experimental values of ²³²Th, ²³⁸U and ²³⁷Np fission rates decrease with distance from the assembly center, where D-T neutrons are generated. An overestimation of about 50% is observed in the calculated ²³⁵U fission rate for the graphite reflector region.

One of the main sources of the disagreement proves to lie in the inability of the codes adopted for generating the multi-group cross-section to take account of the angular distributions of the secondary neutrons emanating from nonelastic reactions. The results of the analysis indicate that the method of calculation currently employed in fusion reactor neutronics overestimates the reflection of neutrons and underestimates the penetration of fast neutrons when a graphite reflector is used.     

Production of Fission Product Xenon

Radioactive Xe isotopes were produced by in-pile melting of irradiated uranyl nitrate hexali yd rate crystals. An aluminum pipe closed at the end was placed adjacent to the reactor core of the TTR-1 (swimming pool type reactor), and uranyl nitrate hexahydrate crystals in a stainless steel capsule were irradiated therein at a neutron flux of 2×10¹¹/cm²-sec. After in-pile cooling for 10–15hr, the capsule was heated to melt the crystals. The rare gas activity released from the crystals was led out of the reactor by helium sweep and collected in a charcoal trap cooled by liquid nitrogen. More than 99% of the rare gas activity was recovered. The composition of the activity at the time of separation, while depending on cooling time and target irradiation history, was in the present case ¹³⁵Xe(–80%), ¹³³Xe(–10%) and small amounts of Kr isotopes. The remaining uranyl nitrate, crystallizes again upon removal of heat, and can serve for succeeding irradiations. About 30 mCi of radioactive Xe isotopes were produced by 4 hr irradiation using 5 g of 13% enriched U. Radioactive Xe produced in this way is being used for medical purposes.     

Production of 40 TBq Tritium Using Neutron-Irradiated 6Li-Al Alloy

A facility was reconstructed for producing tritium in 40 TBq per batch. Gaseous tritium was extracted from neutron-irradiated ⁶Li-Al alloy targets by heating at 700°C under vacuum and collected in uranium. The recovery yield of tritium was about 100% and the isotopic purity of the product was about 95%. Through the production run, no leakage of tritium from the facility was observed.     

Atmospheric 85Kr and 133Xe activity concentrations at locations across Japan following the Fukushima Dai-ichi Nuclear Power Plant accident

Atmospheric ⁸⁵Kr and ¹³³Xe activity concentrations were determined from weekly air samples collected at Sapporo, Akita and Chiba, Japan, throughout 2011. The results demonstrated that the Fukushima Dai-ichi Nuclear Power Plant accident in early March 2011 resulted in high ¹³³Xe activity concentrations as well as elevated levels of ⁸⁵Kr activity; there was a striking increase in the concentrations of both isotopes over the week running from 14 to 22 March as the radioactive plume released from the plant was captured. At Chiba, following the accident, the ⁸⁵Kr activity concentration increased from 1.38 to 17.7 Bq/m³, while the ¹³³Xe levels increased from below the minimum detectable concentration (MDC ≤ 1.9 × 10^?3 Bq/m³) to 1.3 × 10³ Bq/m³. Conversely, at Sapporo and Akita, high ⁸⁵Kr activity concentrations were not observed, due to differences in air transportation mechanisms based on wind directions. Duplicate samples were collected at Chiba to allow the simultaneous analyses of ⁸⁵Kr and ¹³³Xe at the Japan Chemical Analysis Center and the Bundesamt für Strahlenschutz in Germany and the results were in good agreement. The external effective radiation doses resulting from ⁸⁵Kr and ¹³³Xe releases following the accident were estimated to be approximately 7.0 × 10^?3 and 1.3 μSv, respectively, based on the activity concentrations of these nuclides from March to June in 2011 at Chiba.     

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

In-pile release of fission gas from sintered UC pellets in the presence of 8–230 ppm of water vapor in the He sweep gas was measured over the temperature range of 160°–1,000°C. A very complex release behavior was observed and the mechanisms of release were deduced from the manner in which the release depended on the decay constant. It was established that the release of short-lived fission gases during irradiation was controlled mainly by pseudo-recoil, while chemical reaction between UC and water vapor, as well as knock-out, appeared to contribute much more significantly in the case of the longer-lived fission gases. The release of fission gas after reactor shutdown was shown to be governed by the UC-H₂O reaction. The ratio of the release due to this reaction in reference to the total release was found to be dependent not only on the concentration of the water vapor but also on the amount present of the accumulated reaction products. Also, a discussion is given on the inordinately high release of ^135mXe observed at 600°C.     

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}