Instant Release Fraction Research Articles

Molybdenum release from high burnup spent nuclear fuel at alkaline and hyperalkaline pH

This work presents experimental data and modelling of the release of Mo from high-burnup spent nuclear fuel (63 MWd/kgU) at two different pH values, 8.4 and 13.2 in air. The release of Mo from SF to the solution is around two orders of magnitude higher at pH = 13.2 than at pH = 8.4. The high Mo release at high pH would indicate that Mo would not be congruently released with uranium and would have an important contribution to the Instant Release Fraction, with a value of 5.3%.Parallel experiments with pure non irradiated Mo(s) and XPS determinations indicated that the faster dissolution at pH = 13.2 could be the consequence of the higher releases from metallic Mo in the fuel through a surface complexation mechanism promoted by the OH− and the oxidation of the metal to Mo(VI) via the formation of intermediate Mo(IV) and Mo(V) species.

Review of Instant Release Fractions of Long-lived Radionuclides in CANDU and PWR Spent Nuclear Fuels Under the Geological Disposal Conditions

Review of Instant Release Fractions of Long-lived Radionuclides in CANDU and PWR Spent Nuclear Fuels Under the Geological Disposal Conditions

Caesium and iodine release from spent mixed oxide fuels under repository relevant conditions: Initial leaching results

Autoclave leaching experiments are conducted on three well-characterised, irradiated, and cladded mixed oxide fuel-rod segments with burnups ranging from 29 GWd/tHM to 52 GWd/tHM to investigate the instant release fraction of fission gases and long-lived fission products and to assess the long-term fuel matrix corrosion. The segments are exposed to bicarbonate solutions as reference groundwater at neutral pH and a synthetic young cementitious water at pH 13.5 under reducing atmosphere (4 vol% H2 in Ar at 40 bar pressure), since 2018. The initial leaching results for the fission products caesium and iodine as representative elements of the instant release fraction were found to depend on the leachate composition as well as on the fuel burnup.Graphical abstract

Development methodology on determination of instant release fractions for generic safety assessment for direct disposal of spent nuclear fuel

Development methodology on determination of instant release fractions for generic safety assessment for direct disposal of spent nuclear fuel

Contribution of phases segregated from the UO2 matrix to the release of radionuclides from spent nuclear fuel and duration of the Instant Release Fraction (IRF)

During the dissolution of the spent nuclear fuel (SNF) some radionuclides are released to the solution simultaneously from different sources in the fuel. This is of particular importance to some radionuclides that contribute to the Instant Release Fraction (IRF), which govern the initial radiation dose during the dissolution of the SNF. In this work a model that is able to discriminate between the different contributions responsible for the total concentration of a radionuclide in solution was developed. The model permits to establish that uranium and radionuclides that dissolved congruently with the UO2 matrix came from two sources as a function of time: oxidized phases on the surface of the SNF including fines and the matrix itself. Other radionuclides such as Ru and Rh were released from metallic precipitates with dissolution rates lower than the matrix dissolution rate.In the case of radionuclides that were expected to contribute to the IRF, this work showed that Cs, Rb and Sr had initial release rates higher than uranium because a fraction of such radionuclides were segregated from the matrix during the irradiation. Actually it was calculated that the fraction of those radionuclides in the grain boundaries in a BWR SNF powder sample from the center part of a pellet (burnup 42 MWd/kgU) was 2.1%, 0.9%, and 0.6% for Cs, Rb, and Sr, respectively. In addition, the model permitted to calculate the duration of their contribution to the IRF, matrix dissolution governed the release of such radionuclides after 137 days, 75 days, and 164 days for Cs, Rb, and Sr, respectively (at these times, the contribution of the release from grain boundaries was lower than the 0.1%).

Instant release fractions for 14C, 60Co, and 125Sb from irradiated Zircaloy oxide film

AbstractThe oxide films formed on spent fuel claddings are regarded as a potential source of the instantaneous release of radionuclides, such as 14C, after waste disposal. We investigated the instant release fraction using the irradiated oxide exfoliated from a Zircaloy-2 water rod, whose bundle burnup was 53.0 GWd/MTU. We performed a rapid leaching test in a dilute NaOH solution (pH of 12.5) for 10 min in an ultrasonic bath to ensure the release of radionuclides. The activity ratios of the leached amount to the total amount for 14C, 60Co, and 125Sb were extremely low at approximately 10-4 to 10-3, among which the maximum value was 2.65 × 10-3 for 125Sb. These ratios were higher than that predicted from the thermodynamic solubility of ZrO2, i.e., less than 10-6. However, given the low ratios, it is too conservative to regard the inventory of all radionuclides in the Zircaloy oxide as instantaneous release. A small part of the released 14C was found as volatile species.

Determination of tolerance limits for fuel assembly fission gas release summary statistics

Instant release fraction is used in spent fuel final disposal studies to estimate the quickly released inventory from a spent fuel final disposal capsule. The instant release fraction has been correlated with fission gas release, which can be readily calculated with fuel performance codes. In spent fuel final disposal, individual rod fission gas releases are not important, as the fuel is treated in complete capsules containing multiple assemblies. Therefore assembly summary statistics are more suitable for this purpose.Summary statistics such as mean or maximum are useful in this application. However, the nominal values for the assembly mean fission gas release may be nonconservative estimates, whereas the assembly maximum fission gas release is a highly conservative assumption. This results from the “cutoff log-normal” distribution of fission gas release in an assembly and in a more general level, the reactor core. The tail of the distribution contains a small amount of rods with a high fission gas release, whereas a large number of rods typically have low fission gas releases.Compared to nominal summary statistics, a better method to show compliance with regulatory criteria is to determine the distribution of the assembly mean, taking into account the effects of manufacturing tolerances and uncertainty of the power history. The upper tolerance limit is straightforward to calculate with order statistics and with suitable parameters it is accepted by many regulatory authorities. In this work, the calculation of tolerance limits has been applied to mean fission gas release in complete fuel assemblies. A calculation system for such analyses has been developed at VTT, and by calculating each rod multiple times it can be used to determine the tolerance limits for complete nuclear reactor cores.In statistical analyses, typically a large number of simulations must be run. This increases the probability of failure of the simulation program due to nonconvergence of the calculation. This results in missing data, and this could prevent the use of order statistics which are based on the sorting of observations. With missing data, the real order of the observations is unknown. This work suggests a suitable data treatment method to overcome this problem.

Powder Leaching Study for Grain Boundary Inventory of Two High Burnup Fuels

ABSTRACTIn the context of safety assessment, the fraction of inventory that is expected to rapidly dissolve when water contacts the spent fuel is called the Instant Release Fraction (IRF). Conceptually, this fraction consists of radionuclides outside of the uranium dioxide matrix and therefore the fraction can be further divided into the radionuclides in the fuel/cladding gap and radionuclides in the grain boundaries. The relative importance of these two fractions is investigated here for two Swedish high burnup fuels through simultaneous grinding and leaching fuel fragments in simplified groundwater for a short period of time. The hypothesis is that this will expose grain boundaries to leaching solution and provide an estimate of the release of the grain boundary inventory upon contact with water. The studied fragments were used in previous leaching experiments and thus pre-washed to remove any pre-oxidized phases. The results showed a significant release of iodine, cesium and rubidium and to a lower extent molybdenum and technetium. The fraction of inventory in the aqueous phase of actinides and lanthanides was 1-2 orders of magnitude lower than for the elements associated to the IRF. Both fuels displayed a very similar behavior and no correlation as a function of burnup or fission gas release was found.

Overview OF14C release from irradiated zircaloys in geological disposal conditions

ABSTRACTCarbon-14 (radiocarbon,14C) is a long-lived radionuclide (5730 yr) of interest regarding the safety for the management of intermediate level wastes (ILW). The present study gives an overview of the release of14C from irradiated Zircaloy cladding in alkaline media.14C is found either in the alloy part of Zircaloy cladding due to the neutron activation of14N impurities by14N(n,p)14C reaction, or in the oxide layer (ZrO2) formed at the metal surface by the neutron activation of17O from UO2or (U-Pu)O2fuel and water from the primary circuit in the reactor by17O(n,α)14C reaction. Various irradiated and unirradiated Zircaloys have been studied. The total14C inventory has been determined both experimentally and by calculations. The results seem to be in good agreement. Leaching experiments were conducted in alkaline media for several time durations.14C was mainly released as carboxylic acids. Further, corrosion measurements were performed by using both hydrogen measurements and electrochemical measurements. The corrosion rate (CR) ranges from a few nm/yr to 100 nm/yr depending on the surface conditions and the method used for measurement. From a safety assessment point of view, the instant release fraction (IRF) was determined on irradiated Zircaloy-2. The results showed that the14C inventory in the oxide was significantly below the 20% commonly used in safety case assessments.

14C content in CANDU spent fuel cladings and its release under alkaline conditions

ABSTRACTThe total 14C content and its partition between inorganic and organic species were measured on irradiated Zy-4 samples from a CANDU spent fuel rod transferred from Cernavoda Nuclear Power Plant (NPP). Long-term leaching tests and accelerated corrosion tests were carried out to measure the 14C release and corrosion rate, respectively, in chemical conditions relevant to cementitious environment. Experimentally measured 14C inventory was compared to the theoretically one predicted based on the irradiation history and impurity content of Zy-4 by means of ORIGEN computations. CANDU SF claddings have a 14C content of around 2 × 104 Bq/g of Zy-4, mainly as organic compounds (more than 99%). The total 14C content measured by acid dissolution/wet oxidation method is in good agreement with the value estimated by ORIGEN simulations for an average burn-up of 7 MWd/kgU. The total 14C released as dissolved species after 18 days and 18 months of Zy-4 immersing in alkaline solution are similar, indicating that a small amount of 14C was available as instant release fraction (0.05% from the initial 14C content) followed by a very low release rate that could not be measured by liquid scintillation counting. In alkaline conditions, 14C is released predominantly (∼70%) as soluble species, but also inorganic 14C was measured as gaseous species. From the soluble 14C released during leaching test, more than 60% was found to be as organic species. Generally, corrosion rates values ranging between 46 and 130 nm/yr were measured by the linear polarization resistance method. In addition, defects and cracks were observed on the oxide layer by scanning electron microscopy (SEM) investigation.

Aqueous leaching of high burnup UO2 fuel under hydrogen conditions

ABSTRACTLeaching results on fragments of high burnup (65 MWd/kgU) UO2 fuel from a commercial pressurized water reactor are presented. The experiment was performed in simplified granitic groundwater under a hydrogen pressure of up to 5 MPa, representing conditions in a water intrusion scenario for a Swedish KBS-3 design spent nuclear fuel repository. The freshly crushed fragments were pre-washed for 6 days to remove pre-oxidized matrix and part of the instant release fraction of the radionuclide inventory, and then transferred to an autoclave for leaching under hydrogen conditions. Following an initial release of U attributed to dissolution of a pre-oxidized fuel layer caused by the aerated handling mainly during the transfer from pre-washing to autoclave, the U concentration decreased with time to levels of 2-5x10-9 M, which corresponds, approximately, to the solubility of amorphous UO2. The release of radionuclides such as Cs and Sr gradually declined indicating a transition to inhibition of the fuel matrix dissolution.

Influence of the interpellet space to the Instant Release Fraction determination of a commercial UO2 Boiling Water Reactor Spent Nuclear Fuel

The contact of the coolant with the fuel pin during irradiation produces a gradient of temperature in the fuel pellet that segregates the radionuclides (RN) depending on its volatility and reactivity. This segregation determines the Instant Release Fraction (IRF), an important source of radiological risk in the performance assessment (PA) of a Deep Geologic Repository (DGR).RN segregation was studied radially in previous papers. In the present work, it was studied axially, taking into special consideration the cutting position of the solid sample to be studied.Iodine and caesium were the RN with the highest release, while the contribution of rubidium, strontium, molybdenum and technetium to the IRF depended on their chemical state.The interpellet presence (known also as dishing) effect was clearly observed for caesium, increasing its release by one order of magnitude. According to these results, one of the major contributions to the IRF comes from the RN trapped in the dishing and has to be considered in the sampling and data interpretation that will be performed for the PA of the DGR.

Summary of the Euratom Collaborative Project FIRST-Nuclides and Conclusions for the Safety Case

A consortium of 10 partners from seven European countries and the European Commission collaborated in investigating the short-term release of radionuclides from disposed spent nuclear fuel upon canister failure. The Collaborative Project FIRST-Nuclides was implemented in the scope of the 7th Euratom Framework Programme in the period from 2012 to 2014. The objectives and organization of the project are presented, as well as the experiments with highly radioactive materials under investigation. The outcome of the project summarizes the measured instant release fraction (IRF) of safety-relevant isotopes from high burnup spent UO2 nuclear fuels (SNFs). Specifically discussed are the dependencies of the IRF on the sample properties, the gap and grain boundary releases, and the behavior and IRFs of elements such as cesium, iodine, and selenium. The IRFs of nonstandard SNFs were also investigated. The summary is complemented by the presentation of the modeling approaches within the project.

Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples.Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel.The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

Spent Fuel Leaching in the Presence of Corroding Iron

ABSTRACTThe Swedish spent nuclear fuel canister design KBS-3 consists of a cylindrical copper shell surrounding an iron insert that holds the spent fuel. Like in most other canister designs the mass of iron constitutes the majority of the canister weight. In order for groundwater to access the spent fuel in a future repository the copper shell must fail and iron corrosion occur. Spent nuclear fuel dissolution will therefor likely proceed under conditions of simultaneous anoxic iron corrosion. The iron corrosion can likely suppress the spent fuel release by creation of strongly reducing conditions from Fe(II) formation and the generation of large quantities of hydrogen. Redox sensitive radionuclides may either be reductively precipitated by dissolved Fe(II) or from interaction with iron corrosion products such a magnetite or green rusts. The generated hydrogen (up to several MPa) may also inhibit the spent nuclear fuel dissolution at the surface of the fuel via the so called hydrogen effect. In order to probe these effects an autoclave experiment was performed in which a basket with PWR spent nuclear fuel (burnup ∼43 MWd/kgU) was suspended in an autoclave containing a simplified groundwater (10 mM NaCl, 2 mM NaHCO3) with iron powder. The autoclave was sparged and pressurized with argon. Following an initial rise in radionuclide concentrations from dissolution of pre-oxidised phases the U concentration dropped to 3x10-9 M within 76 days, in-line with the solubility of amorphous UO2, expected to form under reducing conditions. Any Cs and Sr release also ceased within 223 days indicating complete transition from dissolution of pre-oxidized phases and instant release fractions to conditions with inhibition of the dissolution of the fuel matrix. Gas phase analysis and pressure monitoring showed a steady build-up of hydrogen at a rate higher than what could be attributed to radiolysis, reaching hydrogen partial pressures of several hundred kPa. The results indicate continuous corrosion of iron, with magnetite as the dominating iron corrosion product.

Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO2) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine (131/129I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50–60 MW d/kgU, and with increasing burn-up, the IRF of 131/129I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO2 fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from 239Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO2 nuclear fuels may reduce the impact of long-lived 129I on the repository performance assessment calculations.

Carbon 14 Distribution in Irradiated BWR Fuel Cladding and Released Carbon 14 after Aqueous Immersion of 6.5 Years

Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 (14C), which is mainly generated by the 14N(n,p)14C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF).In the present work, specific activity of 14C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The 14C specific activity of the base metal was 1.49×104 Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the 17O(n,α)14C reaction, the 14C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, 14C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the 14C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released 14C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the 14C abundance and the low leaching rate suggests that 14C in oxide does not have a significant impact on the IRF in the safety case.

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%.One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Spent UO2 TRISO coated particles – instant release fraction and microstructure evolution

Abstract The impact of burn-up on the instant release fraction (IRF) from spent fuel was studied using very high burn-up UO2 fuel (∼ 100 GWd/t) from a prototype high temperature reactor (HTR). TRISO (TRi-structural-ISO-tropic) particles from the spherical fuel elements contain UO2 fuel kernels (500 μm diameter) which are coated by three tight layers ensuring the encapsulation of fission products during reactor operation. After cracking of the tight coatings 85Kr and 14C as 14CO2 were detected in the gas fraction. Xe was not detected in the gas fraction, although ESEM (Environmental Scanning Electron Micoscope) investigations revealed an accumulation in the buffer. UO2 fuel kernels were exposed to synthetic groundwater under oxic and anoxic/reducing conditions. U concentration in the leachate was below the detection limit, indicating an extremely low matrix dissolution. Within the leach period of 276 d 90Sr and 134/137Cs fractions located at grain boundaries were released and contribution to IRF up to max. 0.2% respectively 8%. Depending on the environmental conditions, different release functions were observed. Second relevant release steps occurred in air after ∼ 120 d, indicating the formation of new accessible leaching sites. ESEM investigations were performed to study the impact of leaching on the microstructure. In oxic environment, numerous intragranular open pores acting as new accessible leaching sites were formed and white spherical spots containing Mo and Zr were identified. Under anoxic/reducing conditions numerous metallic precipitates (Mo, Tc and Ru) filling the intragranular pores and white spherical spots containing Mo and Zr, were detected. In conclusion, leaching in different geochemical environments influenced the speciation of radionuclides and in consequence the stability of neoformed phases, which has an impact on IRF.

Study of the release of the fission gases (Xe and Kr) and the fission products (Cs and I) under anoxic conditions in bicarbonate water

ABSTRACTFor safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of fission products to the instant release fraction (IRF). During the last three years the EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF.Within CP FIRST-Nuclides, a leaching experiment with a cladded SNF pellet was performed in bicarbonate water (19 mM NaCl + 1 mM NaHCO3) under Ar /H2 atmosphere over 333 days. The cladded SNF pellet was obtained from a fuel rod segment which was irradiated in the Gösgen pressurized water reactor; the average burn-up of the segment was 50.4 MWd/kgUO2. In the multi-sampling experiment, gaseous and liquid samples were taken periodically. The moles of the fission gases Kr and Xe released in the gas phase and those of 129I and 137Cs released in solution were measured. Cumulative release fractions of (1.6 ± 0.2)·10-1 fission gases, (1.6 ± 0.1)·10-1129I and (3.9 ± 0.2)·10-2 137Cs, respectively, were achieved after 333 days of leaching. Accordingly the release ratio of fission gases to 129I was 1:1 and the release ratio of fission gases to 137Cs was 4:1, respectively.