Determination of Li-6/Li-7 Isotopic Ratio by Means of Thermogravimetric Analysis for Lithium Sulphate Monohydrate

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r _o=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO₂ fuel pellet) at pellet average burnup of 1430 MWd/kgHM.     

Failure of high burnup fuels under reactivity-initiated accident conditions

Pulse irradiation experiments of high burnup light-water-reactor fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO₂ rod at a burnup of 69 GW d/t failed due to pellet-cladding mechanical interaction (PCMI) in the test LS-1. The fuel enthalpy at which fuel failure occurred was comparable to those for PWR-UO₂ rods of 71 to 77 GW d/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuel rods suggested that the morphology of hydride precipitation, which depends on the cladding texture, affects the fuel failure limit. The tests BZ-1 and BZ-2 with PWR-MOX rods of 48 and 59 GW d/t, respectively, also resulted in PCMI failure. The fuel enthalpies at failure were consistent with a tendency formed by the previous test results with UO₂ fuel rods, if the failure enthalpy is plotted as a function of the cladding outer oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states including oxidation and hydride precipitation, and the same failure limit is applicable to UO₂ and MOX fuels below 59 GW d/t.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Agency (JAEA). The paper presents recent results obtained from the NSRR power burst experiments with high burnup fuels, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Results from the recent four experiments on high burnup (about 60 to 78 MWd/kgU) PWR UO₂ rods with advanced cladding alloys showed that the fuel rods with improved corrosion resistance have larger safety margin against the PCMI failure than conventional Zircaloy-4 rods. The tests also suggested that the smaller inventory of inter-granular gas in the pellets with the large grain could reduce the fission gas release during the RIA transient; and high burnup structure in pellet periphery (so-called rim structure) does not have strong effect on reduction of the failure threshold because the PCMI load is produced primarily by solid thermal expansion.     

破损燃料棒二次氢化行为观察与分析

燃料棒在堆内运行时,由于初次破口会导致包壳发生二次氢化现象,二次氢化是导致燃料棒发生严重破损的重要因素。针对实际工况下的破损燃料棒,在中国原子能科学研究院燃料与材料检验设施（303热室）上开展了相关辐照后检验,并采用热室金相手段,对燃料棒二次氢化行为进行了观察分析。结果表明：二次氢化破口有明显的氢化肿胀现象;氢化物分阶段从内壁扩散到外壁,并形成“日爆”现象;二次氢化部位芯块温度明显升高,并会导致芯块气孔迁移、芯块晶粒长大、柱状晶生长等现象发生;相比未破损棒,破损棒二次氢化部位水侧氧化膜厚度有增加现象,但仍处于正常范围内。     

Behavior of Irradiated ATR/MOX Fuel under Reactivity Initiated Accident Conditions

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20MWd/kgHM were conducted at the NSRR in Japan Atomic Energy Research Institute to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Relatively large radial deformation of the fuel rods due to pellet-cladding mechanical interaction occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet.     

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.     

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO₂ LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded “laser-flash” device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO₂ of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.     

Effect of hydride orientation on fracture toughness of Zircaloy-4 cladding

Hydrogen embrittlement is one of the major degradation mechanisms for high burnup fuel cladding during reactor service and spent fuel dry storage, which is related to the hydrogen concentration, morphology and orientation of zirconium hydrides. In this work, the J-integral values for X-specimens with different hydride orientations are measured to evaluate the fracture toughness of Zircaloy-4 (Zry-4) cladding. The toughness values for Zry-4 cladding with various percentages of radial hydrides are much smaller than those with circumferential hydrides only in the same hydrogen content level at 25 °C. The fractograghic features reveal that the crack path is influenced by the orientation of zirconium hydride. Moreover, the fracture toughness measurements for X-specimens at 300 °C are not sensitive to a variation in hydride orientation but to hydrogen concentration.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

Experimental Observations on Fuel Pellet Performance at High Burnup

The Japanese and Spanish nuclear industries have conducted joint experimental programmes since early 1990's to address fuel performance issues such as fuel volume change and fission gas release. These efforts have produced large amount of valuable information on in-reactor performance of fuel materials representing current and potential future fuel designs. A large number of thoroughly characterised fuel rods composed of different materials have been irradiated in the Spanish PWR Vandellós II for up to five irradiation cycles achieving rod average burnup of up to 75 MWd/kgU.

This paper looks into the fuel pellet performance at high burnup only based on the extensive PIE programme both on-site and in hot-cells carried out over this fuel and other related data on similar fuel rods thus supporting and enriching the conclusions.     

Fast flux irradiation tests performed at high temperature

The first gas-cooled fast breeder reactor (GCFR) fast flux irradiation experiment [F-1(X094)] consists of seven fuel rods clad in 20% cold-worked 316 stainless steel. The rods are individually encapsuled, with sodium filling the gaps within the capsule walls. The rods are fueled with (15% Pu, 85% U)O₂ and have depleted UO₂ lower and upper axial blankets and charcoal to trap volatile fission products. The cladding i.d. temperature range covered by these rods is 570–760°C (1055–1400°F).The in-reactor performance of the fuel rods in the F-1 high-temperature experiment, which achieved a burnup of 121 MWd/kg (13.0 at.%) on the lead rod, is described. All rods in the experiment have remained intact. The results of interim examinations [at 25 and 50 MWd/kg (2.7 and 5.4 at.%)] of fuel and fission product behavior and transport and comparisons of observed results with LIFE-III code predictions are described.The F-3 experiment, which consists of ten encapsulated GCFR fuel rods with surface-roughened (ribbed) cladding, shares a nineteen capsule subassembly with Argonne National Laboratory. Temperatures are controlled over the range 675°C (1250°F) to 750°C (1380°F). Irradiation is in the core region of the EBR-II and thus permits achievement of a higher fluence-to-burnup ratio than that obtained in the F-1 experiment.Preliminary results of a planned interim examination at an exposure of 46 MWd/kg (4.9 at.%) burnup and a fluence of 5.2 × 10²² n/cm² show that cladding failures occurred in nine of the ten rods. Preliminary indications are that the failures are due to defects in the sodium bond between the fuel rod and the capsule.The tests completed and currently under way have been scoping in nature, and irradiation in EBR-II of GCFR prototypical fuel (pressure equalized) rods with ribbed cladding is required to provide the information needed for reactor design on effects of exposure to high fluence and burnup and on design reliability for a statistically significant number of rods. The design and the operating conditions for the F-5 experiment being prepared for this purpose are described.     

Radiological aspects and behavior of spent fuel considering long-term storage

For spent nuclear fuel management in Germany, the concept of dry interim storage in dual purpose casks before direct disposal is applied. Current operation licenses for storage facilities have been granted for a storage time of 40 years. Due to the current delay in site selection, an extension of the storage time seems inevitable. In consideration of this issue, GRS performed burnup calculations, thermal and mechanical analyses as well as particle transport and shielding calculations for UO₂ and MOX fuels stored in a cask to investigate long-term behavior of the spent fuel related parameters and the radiological consequences. It is shown that at the beginning of the dry storage period, cladding hoop stress levels sufficient to cause hydride reorientation could be present in fuel rods with a burnup higher than 55 GWd/tHM. The long-term behavior of the cladding temperatures indicates the possibility of reaching the ductile-to-brittle transition temperature during extended storage scenarios. Surface dose rates are 3 times higher when a cask is partially loaded with 4 MOX fuel assemblies. Due to radioactive decay, long-term storage will have a positive impact on the radiological environment around the cask.     

RANNS Code Analysis on the Local Mechanical Conditions of Cladding of High Burnup Fuel Rods under PCMI in RIA-Simulated Experiments in NSRR

The RANNS code analyzes behavior of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the RIA-simulated experiments in the Nuclear Safety Research Reactor (NSRR), OI-10 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. The pre-accident, or End-of-Life conditions of the rod were predicted by the fuel performance code FEAMXI-6. In the calculations by the two-dimensional model of RANNS, the plastic strain increases at the cladding ridges during PCMI were compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.     

Post Irradiation Examination of High Burnup PWR Fuel

Post irradiation examination (PIE) of a high burnt lead fuel assembly, which was irradiated to demonstrate fuel integrity at high burnup, was performed before the start of the full batch loading of high burnup fuel of 48GWd/t maximum fuel assembly burnup.

The lead fuel assembly was 17×17 B-type PWR fuel which was supplied by Nuclear Fuel Industries, Ltd. (NFI) and achieved the maximum burnup of 45 GWd/t after 4 cycles of irradiation in Ohi Unit 1 of the Kansai Electric Power Co. Inc. (Kansai).

Twelve fuel rods extracted from the lead fuel assembly at the reactor site were examined at the hot-cell facility of Japan Atomic Energy Research Institute (JAERI) in Tokai-mura.

Visually, the fuel rods appeared to be in good conditions, but some small spallings were observed at the second span from the top where oxide film was relatively thicker than other spans. Even in this span, the maximum oxide film thickness was less than 50 μm Fission gas release rate was less than 1%, which caused only a small increase in fuel rod internal pressure. Mechanical properties of the fuel cladding were evaluated by tensile tests.

These PIE results were within the range of other PIE data previously obtained from domestic and foreign PWR fuel rods. The data confirmed that the integrity of B-type fuel would be maintained at least up to 48 GWd/t.     

Uranium-zirconium hydride fuel properties

Properties of the two-phase hydride U_0.3ZrH_1.6 pertinent to performance as a nuclear fuel for LWRs are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of 4 decades ago and from the more restricted data base prepared for the TRIGA research reactors some 3 decades back. Transport, mechanical, thermal and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to ∼650 °C (the design limit recommended by the fuel developer is 750 °C). To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal instead of helium. Because the former has a thermal conductivity ∼100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens the possibility of initial gap sizes large enough to significantly delay the onset of pellet-cladding mechanical interaction (PCMI). The large fission-product swelling rate of hydride fuel (3× that of oxide fuel) requires an initial radial fuel-cladding gap of ∼300 m if PCMI is to be avoided. The liquid-metal bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel is the source of phenomena during operation that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrH_x, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1.6 to ∼1.45 at the center and ∼1.70 at the periphery. Because the density of the hydride decreases with increasing H/Zr ratio, the result of H redistribution is to subject the interior of the pellet to a tensile stress while the outside of the pellet is placed in compression. The resulting stress at the pellet periphery is sufficient to overcome the tensile stress due to thermal expansion in the temperature gradient and to prevent radial cracking that is a characteristic of oxide fuel. Several mechanisms for reduction of the H/Zr ratio during irradiation are identified. The first is transfer of impurity oxygen in the fuel from Zr to rare-earth oxide fission products. The second is the formation of metal hydrides by these same fission products. The third is by loss to the plenum as H₂.The review of the fabrication method for the hydride fuel suggests that its production, even on a large scale, may be significantly higher than the cost of oxide fuel fabrication.     

Pressurized heavy water reactor fuel behaviour in power ramp conditions

In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.     

Failure Propensities of Pressurized Zircaloy Tube Containing Iodine and Other Chemical Species

Abstract

Failure propensities of Zircaloy-4 cladding tube internally pressurized with Ar gas containing iodine and iodine plus each of other chemical species were examined at 360°C, to study the effect of corrosive fission products (FPs) on the integrity of spent nuclear fuel rods during dry storage, and also to assess the capability of preventing the spent fuel rod degradation.

The iodine stress corrosion cracking (SCC) of Zircaloy tube occurred in the long time/low stress exposure tests at stresses much lower than the conventional “threshold stress”, with considerably large strains at failure. The addition of cesium to iodine perfectly suppressed the SCC. It is inferred from these results that the degradation of spent fuel rods induced by corrosive fission products is unlikely during dry storage. Even if iodine alone should take effect, a proper strain limit could prevent spent fuel rods from incurring iodine induced effects because of considerably large strains necessary for iodine SCC of Zircaloy tube at low stresses.     

Effect of Fuel Pellet Eccentricity with Cladding on Fuel Rod Thermal Behavior under Reactivity Initiated Accident Condition

Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g?UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum temperature difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g?UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA conditions.