后处理厂释放气体中稳定惰性气体同位素在核保障监督中的应用研究

用Origen2．1计算模式对压水堆元件中Kr,Xe相关同位素与燃耗的关系进行了计算,并估算了后处理厂烟囱释放气体中Kr,Xe各稳定同位素的来源,丰度和原子浓度.^82Kr,^129Xe可用作环境样品中惰性气体同位素的天然本底;裂片^83Kr／^86Kr．^84Kr／^86Kr、^131Xe／^134Xe和^132Xe／^134Xe的丰度比值,可用于指示乏燃料燃耗,进而估算正在被分离的钚同位素组成,并有可能对后处理厂实行保障监督。     

On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation.     

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the “rim effect.” High burn-up phenomena in WWER-440 UO₂ fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity.In this paper, the radial burn-up and fissile products distributions of WWER-440 UO₂ fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.     

Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8–54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.     

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.     

Analytical investigations of irradiated inert matrix fuel

Yttria stabilised zirconia (YSZ) inert matrix fuel (IMF) fabricated at PSI and irradiated 3 years in the Halden Material Test Reactor (HBWR) since 2000, has been examined by Electron Probe Microanalysis (EPMA) and Secondary Ion Mass Spectroscopy (SIMS) after irradiation and compared with data gained for the unirradiated material. The examined pellet cross-section was estimated to have an equivalent burn-up of 22 MW d kg⁻¹. EPMA measurements demonstrate that the burn-up was rather flat over more than the half pellet radius. A Pu consumption of about 2.5 wt% has been measured with a higher rate in the fuel border zone. The high fuel temperature is responsible for a certain homogenisation of the mineral phases in the fuel centre region whereas the border zone has remained rather with an as-fabricated phase distribution. The central part was also characterised by a dense porosity distribution as well as a temperature and relocation driven depletion of the volatile fission products Xe and Cs. In addition, SIMS has been realised on the same specimen in order to determine the semi-quantitative distribution of different isotopes in the pellet.     

Molecular dynamics simulation of the pressure-volume-temperature data of xenon for a nuclear fuel

The exact equation of state for the fission gas is necessary for the accurate prediction of the fission gas behavior in a nuclear fuel. However, certain kinds of extrapolating data are used to construct and verify the equations of state for the fission gas because experimental data are very limited at high temperatures and pressures that are encountered in the nuclear fuel. To fill the lack of experimental data for the fission gas, the behavior of Xe gas atoms was investigated by molecular dynamics simulation assuming an exponential-six potential. The molecular dynamics simulation produced reasonable pressure-volume-temperature data for Xe and the simulation results were compared with existing equations of state for Xe.     

Estimation of the Grain Boundary Gas Inventory in MIMAS/AUC MOX Fuel and Consistency with REP-Na Test Results

In the frame of its research activities on fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” performed the REP-Na program in the CABRI reactor devoted to the study of Reactivity Initiated Accidents. Focused on high burn-up UO2 and MOX fuel behaviour, twelve tests (8 UO2 and 4 MOX) were realized from 1993 to 2000. In all these tests, the influence of grain boundary gas was evidenced and it appeared necessary to perform some estimation of its inventory in irradiated fuel. Such evaluations are presented for the MOX MIMAS/AUC fuel, based on two different approaches: “experimental” and “theoretical.” The fission gas amount located at the grain boundaries increases with burn-up in correlation with the production, but also with the initial Pu enrichment, as soon as the agglomerates have reached the full restructuring threshold for the High Burn-up Structure. The consistency with the REP-Na test results is checked, showing that a significant cladding deformation is needed, clearly higher than for UO2 fuel in order to release all the grain boundary gas in RIA. Furthermore, to the fission gas effect, adds the helium's occluded in the irradiated fuel whose amount increases with burn-up, Pu enrichment and ²⁴¹Pu and ²⁴¹Am initial content.     

Indirect assessment of 135Cs activity in the ventilation system of the Ignalina NPP RBMK-1500 reactor

The accumulation of long-lived ¹³⁵Cs isotope in the ventilation system components of the Ignalina NPP Unit 2 was investigated by spectrometric measurements and mathematical modeling. Volumetric activities of fission noble gas and other short-lived isotopes (⁴¹Ar, ^85mKr, ⁸⁷Kr, ⁸⁸Kr, ⁸⁸Rb, ¹³³Xe, ^133mXe, ¹³⁵Xe, ¹³⁸Xe, ¹³⁸Cs) have been measured by gamma spectrometric technique. Modeling of radionuclide transport in the ventilation system provides possibility of determining essential transport parameters: effective gas flow, mean gas retention time, deposition rate of aerosols. Estimated parameters were used for indirect evaluation of difficult to measure ¹³⁵Cs isotope activity in the ventilation system components: a delay chamber and aerosol filters. The results show that the major part of ¹³⁵Cs activity is accumulated in aerosol filters, whereas the total surface activity of the delay chamber is considerably lower. Specific activities of the ventilation system components of the Ignalina NPP Unit 2 are below the clearance levels for ¹³⁵Cs.     

Extension of the TRANSURANUS burn-up model

The validation range of the model in the TRANSURANUS fuel performance code for calculating the radial power density and burn-up in UO₂ fuel has been extended from 64 MWd/kgHM up to 102 MWd/kgHM, thereby improving also its precision. In addition, the first verification of calculations with post-irradiation examination data is reported for LWR-MOX fuel with a rod average burn-up up to 45 MWd/kgHM. The extension covers the inclusion of new isotopes in order to account for the production of ²³⁸Pu. The corresponding one-group cross-sections used in the equations rely on results obtained with ALEPH, a new Monte Carlo burn-up code. The experimental verification is based on electron probe microanalysis (EPMA) and on secondary ion mass spectrometry (SIMS) as well as radiochemical data of fuel irradiated in commercial power plants. The deviations are quantified in terms of frequency distributions of the relative errors. The relative errors on the burn-up distributions in both fuel types remain below 12%, corresponding to the experimental scatter.     

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.     

放射性惰性气体分离与分离材料研究进展

放射性惰性气体氙(¹³³Xe)、氪(⁸⁵Kr)与氩(³⁷Ar)是重要的气体裂变产物,主要产生于核电站反应堆、地下核试验、乏燃料后处理等人类核活动中。放射性惰性气体的快速高效分离、分析与检测在核军控核查、核环境监控、核燃料循环等领域中均有重要意义。利用固体多孔吸附材料在室温环境下从复杂环境气氛中选择性地将目标放射性惰性气体高效吸附分离出来是目前最简单与高效的方法。近些年发展的金属有机框架材料、多孔有机框架材料、多孔有机聚合物等新型多孔材料在惰性气体Xe与Kr的分离上已经展现出优异的性能与良好的应用前景。本文系统性地综述了放射性惰性气体(Xe、Kr、Ar)分离与分离材料的研究进展,并对未来研究趋势进行了展望。     

Gas phase in experimental fuel elements with compact uranium dioxide,irradiated in the SM-2 reactor

Conclusions A complex procedure has been developed for the study of gas release from nuclear fuel, including reactor measurements and post-reactor determination of the amount and composition of the gas medium in the fuel elements at room and elevated temperatures. In fuel elements with compact uranium dioxide (density 10.0–10.43 g/cm³), in addition to gaseous fission products and the helium introduced, Ar, H₂, O₂, CO, CO₂, and N₂ are present, and after irradiation their quantity exceeds the initial quantity, measured for unirradiated fuel elements, by a factor of several.The yield of Xe and Kr under the can of the fuel elements during irradiation of uranium dioxide in the SM-2 reactor amounts to 30–50%, but the measured ratio of Xe/Kr exceeds the calculated ratio by a factor of 1.2, because of the reaction¹³⁵Xe(n, )¹³⁶Xe. The content in the fuel of adsorbed helium is equal to 0.004 n.cm³/g UO₂. The data obtained can be used for physics and technological calculations, and also for refining the procedure for the determination of gas release.Translated from Atomnaya Énergiya, Vol. 57, No. 2, pp. 91–95, August, 1984.     

The evolution of gas from uranium dioxide

Conclusions We have examined the basic laws governing the character of gas evolution and the behavior of uranium dioxide during the operation of a fuel element prior to thorough burn-up. The behavior of the uranium dioxide and the evolution of gases depend directly on the distributions of energy evolution and temperature with respect to radius in the fuel element, and on the process of structure formation and burn-up in the different zones of the fuel. The evolution of gases from uranium dioxide during irradiation can be estimated on the assumption that all the gaseous fission products evolve from the columnar crystal zone and that evolution of gases from the equiaxial grain zone takes place by a mechanism of thermally activated diffusion.Sorption of gaseous fission products at the surface of the fuel can lead to errors in determining the quantity of gaseous fission products evolved from the uranium dioxide during postreactor determination by the fuel-can puncture method.Translated from Atomnaya Énergiya, Vol. 40, No. 5, pp. 390–395, May, 1976.     

Crack formation,crack healing and porosity redistribution during irradiation of UO2 and (U,Pu) O2

The irradiation-induced void volume redistribution in the fuel was analysed. The radial crack volume and porosity distributions, the central radii and the radial gap width were measured after irradiation and compared with the calculated values. Short-time (He-loop experiments in the FR2 reactor), medium-time (bundle irradiation in the BR2 reactor) and long-time (trefoil-irradiation in the DFR reactor) irradiated fuel pins were examined. The model of pore migration, used in the computer code SATURN-la, is based on the evaporation-condensation mechanism. Measured swelling rates were extrapolated to higher temperatures and used. The crack volume distribution was calculated on the basis of a multifractured fuel model. One can conclude from the comparison between calculated and measured void volume distributions that several mechanisms redistribute void volume. These are crack formation, crack healing, migration of sinter pores and fission gas bubbles, gas swelling, evaporation-condensation phenomena in the region of the central void, irradiation-induced sintering and increase in diameter of the cladding.     

Fuel burn-up in nuclear reactors

The paper describes the method of calculating fuel burn-up in nuclear reactors, taking into account the capture and multiplication of neutrons while slowing down. In the calculations, account is taken of the burn-up of U²³⁵ and the build-up and burn-up of Np²³⁹, Pu²³⁹, Pu²⁴⁰, Pu²⁴¹ and of the fission fragments.     

Performance evaluation of a resonance ionization mass spectrometer developed for the FFDL system of fast reactors

To prevent a fuel failure event from becoming a serious radiation accident, sodium-cooled fast reactors are equipped with a system for failed fuel detection and location (FFDL). The FFDL instrument employed in the prototype fast breeder reactor Monju is based on the gas tagging method, in which precise and accurate measurements of krypton and xenon isotope ratios (⁷⁸Kr/⁸⁰Kr, ⁸²Kr/⁸⁰Kr and ¹²⁶Xe/¹²⁹Xe) must be performed in a short time. Burnup measurements also contribute to accurate determination of ⁸²Kr/⁸⁰Kr. We have developed a highly sensitive resonance ionization mass spectrometer for the isotopic analyses, which uses resonance ionization of Kr and Xe atoms by a pulsed laser at wavelengths of 216.7 and 249.6 nm, respectively. In evaluating the performance of our spectrometer, we find that systematic errors caused by isotope shifts can be reduced to negligible levels, and that statistical errors of 3% at a nuclide concentration of 7 ppt can be achieved with a single measurement time of about 40 minutes for each Kr and Xe isotope ratio. This means that, within one hour, about 200 fuel assemblies can be individually identified with a probability of 99%, verifying the applicability of our spectrometer to the FFDL system of fast reactors.     

Modeling of fission gas swelling in LWR UO2 fuel under irradiation

An understanding of the behavior of fission gas in uranium dioxide (UO₂) fuel is necessary for the prediction of the performance of fuel rods under irradiation. A mechanistic model for matrix swelling by the fission gas in LWR UO₂ fuel is presented. The model takes into account intragranular and intergranular fission gas bubbles behavior as a function of irradiation time, temperature, fission rate and burn-up. The intragranular bubbles are assumed to be nucleated along the track of fission fragments, which play the dual role of creator and destroyer of intragranular bubbles. The intergranular bubble nuclei is produced until such time that a gas atom is more likely to be captured by an existing nucleus than to meet another gas atom and form a new nucleus. The capability of this model was validated by a comparison with the measured data of fission gas behavior such as intragranular bubble size, bubble density and total fuel swelling. It was found that the calculated intragranular bubble size and density are in reasonable agreement with the measured results in a broad range of average fuel burn-ups 6–83 GW d/tU. Especially, the model correctly predicts the fuel swelling up to a burn-up of about 70 GW d/tU.     

低温下活性炭吸附分离Kr和Xe

研究了低温下活性炭吸附分离Kr和Xe的方法。Kr和Xe混合气在-78 ℃活性炭吸附柱上进行富集,根据Kr、Xe在活性炭柱上脱附条件的差异实现了Kr和Xe的分离。结果表明,Kr和Xe的回收率均大于90%,Kr样品中Xe的去污系数达10⁴以上,Xe样品中Kr的去污系数达10³以上。     

复杂混合气体组分在低温活性炭表面的吸附特性

为实现大体积气体中微量放射性气体Kr、Xe同位素的测量,须将混合气体进行浓集并将目标气体吸附于10 mL左右的活性炭源盒中。本实验对混合气体中各组分在活性炭分离柱上的吸附性能进行研究,建立了通过去除其他杂质气体、浓集大体积气体制备放射性Kr和Xe活度源的方法。根据反应堆流出气体和核爆可能生成的气体组分,配制了模拟气体,使用活化的4A分子筛对其中的水和CO₂进行模拟去除,获得了流程中去除水和CO₂的实验条件;选择5个低温点(273、264、255、246、238 K),在低温活性炭柱上对H₂、CO、CH₄、Kr和Xe的吸附特性进行研究,测定了各气体在不同温度下的吸附穿透曲线。结果表明,室温下4A分子筛对水和CO₂有较好的吸附效果。低温下,H₂、CO不易在活性炭表面吸附;CH₄、Kr吸附性质相似;Xe吸附能力较强。低温下难以去除的CH₄可在高温下氧化去除。因此,可根据混合气体中各组分性质的不同实现杂质气体的去除和目标气体Kr、Xe的回收测量。