Spent Fuel Elements Research Articles

Impact of Radiation-Induced Microstructures on the Integrity of Spent Nuclear Fuel (SNF) Elements in Long-Term Storage

Abstract. The current safety concept provides for a period in the range of 40 years for interim storage of spent fuel elements. Since the requirement for proof of safety for to up to 100 years arises, the integrity of the spent fuel elements in prolonged interim storage and long-term repositories is becoming a critical issue. In response to this safety matter, this study aims to assess the impact of radiation-induced microstructures on the mechanical properties of spent fuel elements, in order to provide reliable structural performance limits and safety margins. The physical processes involved in radiation damage and the effect of radiation damage on mechanical properties are inherently multiscalar and hierarchical. Damage evolution under irradiation begins at the atomic scale, with primary knock-on atoms (PKAs) resulting in displacement cascades (primary damage), followed by the defect clusters leading to microstructural deformations. In this context, we have developed and applied a multiscale simulation methodology consistent with the multistage damage mechanisms and the corresponding effects on the mechanical properties of spent fuel cladding and its integrity. Within the improved hierarchical modelling sequence, the effect of the radiation field on the fuel element cladding material (Zircalloy-4) is assessed using Monte Carlo methods. A molecular dynamics method is employed to model damage formation by PKAs and primary damage defect configurations. The formation of clusters and evolution of microstructures are simulated by extending the simulation sequence to a longer time scale with the kinetic Monte Carlo (KMC) method. Transferring the calculated radiation-induced microstructures into macroscopic quantities is ultimately decisive for the structural/mechanical behaviour and stability of the cladding material, and thus for long-term integrity of the spent fuel elements. Results of the multiscale modelling and simulations as well as a comparison with experimental results will be presented at the conference session.

Utilizing of Spent Fuel Storage Pool in Gamma Irradiation Experiments

Abstract The objective of this study is suggesting a new method for gamma irradiating experiment using spent fuel elements (SFEs). An irradiation device was developed to contain the sample during the gamma irradiation process under water in spent fuel storage pool. The maximum burned fuel elements were used as a source of gamma rays with intensity and energy spectrum depend on the decay time from the reactor core withdrawal. A SFE configuration was introduced to verify an ideal condition of irradiating the sample with uniform gamma flux. The gamma dose rate inside the sample chamber was calculated using SCALE/MAVRIC code for the SFEs configuration source at decay times ranging between 0 and 5 years. A relationship was introduced to obtain the required accumulated dose depending on choosing the proper SFE configuration with the proper irradiation time.

Assessment of Criticality Safety for Spent Fuel Elements in Open Pool Type Storage

Storage of spent fuel elements needs a previous calculation to ensure verifying the criteria of storage from the nuclear criticality point of view. So, arrangement of spent fuel elements in the storage must result in a configuration that veri-fying an effective multiplication factor (k-eff) lower than 0.85. The spent fuel elements storage is an open pool con-nected with the main pool of the reactor via transfer channel to facilitate transporting the irradiated fuel elements under water surface to verify a high radiological safety. The fuel element is a plate type material testing reactor (MTR) with enrichment of 19.97%. The storage pool was designed to store around 800 spent fuel elements as a temporary stage before transporting to a permanent storage. Basket and pool models, loaded with fresh fuel elements, were applied using SCALE/KENO-VI sequence to determine the k-eff during normal storage condition as well as water-drop accident. TRITON and KENO-VI sequence was applied to determine the k-eff for the basket and pool, loaded with spent fuel elements, during the normal and the accidental conditions. All the spent fuel elements were irradiated in a core of 22 MW power to the maximum burn up of discharge. The k-eff values show a proper agreement with previous calculations developed with MCNP5 code.

Effect of blisters and rims on radial hydride precipitation in fuel elements under dry storage conditions

ABSTRACT Hydrogen enters in Zry clads during fuel element life in service; when hydrogen exceeds the terminal solid solubility, circumferential hydrides are precipitated. In the presence of thermal gradients, produced for instance by oxide spallation, localized concentration of hydrides - known as blisters - can be formed. Due to the fact that zirconium hydrides have a lower density that zirconium alloys, stresses are produced around blisters. These stresses may overcome the necessary threshold to precipitate radial hydrides. Moreover, cracks inside blisters may act as initiators of delayed hydride cracking (DHC). In the dry storage facilities developed for Atucha 1, Zry - 4 clads of spent fuel elements (SFE) must withstand environmental conditions which include maximum temperature of 200°C and internal SFE pressure of ~ 75 bar (value at end of cycle). During dry storage, if blisters are present, stresses around them are superposed to circumferential stresses generated by the internal pressure due to fission gasses helping precipitation and/or growth of radial hydrides and DHC process. In order to assess this possibility, in the present work sections of Zry-4 cladding were hydrided. Blisters have been grown by thermal gradient under different conditions of temperature and shapes of thermal contact in order to obtain different blister dimensions and fraction of radial hydrides. After blister formation, some specimens were subjected to Hydride Reorientation Test (HRT) by applying an internal pressure of 75 bar at 200°C. Finally, tubes were sectioned and metallographically analyzed looking for radial hydrides. In some samples radial hydrides were observed, which effectively grew by the internal pressure of Zry-cladding during HRT.

Calculation of dose rates due to loss-of-coolant accident in open-pool spent-fuel storage

The objective of the spent-fuel storage pool is shielding the worker and public from radiation emitted by radioactive decay in the spent fuel and providing a barrier for any radioactive release. In open-pool multipurpose reactor, the spent-fuel storage pool is connected to the main pool through the transfer channel. It was prepared to store 528 spent-fuel elements distributed in two racks that constructed one above the other. The loss-of-coolant accident (LOCA) in spent-fuel storage pool could result in rising of the radiation dose in the reactor building as the water level in the pool falls. The value of the radiation dose rate depends on the height of the water level above the spent fuels, and the number of spent-fuel elements stored in the storage pool during LOCA. The dose rate calculations were carried out starting from the minimum height which the water level could drop above the spent-fuel storage racks. The calculations were carried for two cases as follows: the full capacity of both racks and the full capacity of the lower rack only. Monte Carlo N-Particle Transport MCNP5 code was used to calculate the radiation dose rate above the storage pool and in the control room. The results show that the dose rate in the control room would be lower than the permissible limit when the water level height was 270 and 140 cm for the two cases, respectively. The dose rate above the storage pool would be lower than the permissible limit when the water height above the racks is higher than 385 cm in the first case, and 290 cm for the second case.

Gamma spectrometry inspection of TRIGA MARK II fuel using caesium isotopes

Gamma spectrometry is one of the common methods to inspect the spent fuel from research reactors. This method has been applied to in-pool measurements of the Spent Fuel Elements (SPEs) of the TRIGA Mark II research reactor. Due to mixed nature of the reactor core and complicated irradiation history of the fuel elements (FEs), the gamma spectrometry of the FE establishes improvements in the calculation and measurement of the SPE. In order to inspect the TRIGA SPE from dry storage and cooled fuel from the reactor pool, the selected spend fuels are scanned and measured using the fuel-scanning machine. Gamma spectrometry is performed by HPGe detector for spend fuel inspection and determination of the 137Cs activity and 134Cs/ 137Cs ratio. In this work, the steps of the detector calibration and the use of the Monte Carlo radiation transport code (MCNP5) have been described. In addition, the fuel-scanning machine and the gamma spectrometer are modelled by MCNP5 to simulate the gamma transport from fuel to detector. It also simulate the gamma spectrometer calibration for the burn up determination of the spend fuel. The results from MCNP5 simulation are applied to spectroscopic measurements and compared with the theoretical predictions of the neutronics code ORIGEN2 in this research work.

TRIGA fuel burn-up calculations and its confirmation

The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.

Recovery of Uranium and Plutonium from Spent Fuel Elements of Nuclear Reactors

The results of investigations of the preliminary removal of the products of radioactive decomposition from irradiated nuclear fuel to obtain uranium and plutonium which are suitable for reuse in fuel fabrication are presented. Nitrate-alkali melts are used for the operation. The experiments are performed on simulators and irradiated samples of BOR-60 fuel in remote-controlled hot boxes. The coefficients of removal of fission products are presented. A technological scheme, which will shorten the fuel cycle, for purifying hot nuclear fuel is recommended.

Design of a Spent-Fuel Assay Device Using a Lead Spectrometer

The neutron slowing-down-time method for nondestructive assay of light water reactor spent fuel has been under development for many years. Results for a newly optimized design of a lead slowing-down-time spectrometer for spent-nuclear-fuel assay are presented. Monte Carlo analyses were performed to optimize the design of the assay device, determine its main parameters, investigate the effects of the spent-fuel assembly and the detector impurities on its performance, determine the fission signatures of the fissile isotopes in spent-fuel elements, and simulate the assay signal as a function of the slowing-down time, assuming threshold fission chambers for the assay detectors. The assay signals from the threshold detectors were analyzed to predict the unknown masses of the fissile isotopes in a typical spent commercial light water reactor fuel element. The broadened resolution of the system caused by the presence of the spent fuel inside the spectrometer pile was found sufficient to separate the signatures of the U and Pu fissiles in spent fuel.

Kainate and isoproterenol evoke synergistically glutamate release in the mouse striatum

The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.

Special Safety Aspects of Drift Disposal—The Thermal Simulation of the Drift Storage Experiment

The thermal simulation of drift storage (TSS) full-scale test is being performed in the Asse salt mine in Germany to study the thermomechanical effects of the direct disposal of spent-fuel elements in a nuclear salt repository. The test field comprises two parallel test drifts, in each of which three dummy casks are deposited. The remaining volume of the drifts is backfilled with crushed salt. The casks are equipped with electrical heaters with a thermal power output of 6.4 kW each. The test has been in operation since September 1990. A design temperature of ~210°C at the surface of the heater casks was reached after 5 months. Because the thermal conductivity of the backfill increases with its compaction, the temperature at the surface of the casks subsequently decreased, reaching ~170°C after 5 yr of heating. The drift closure, which causes increasing compaction of the backfill, was considerably accelerated by heating. However, the initial backfill porosity of 35% decreased more slowly than predicted, to ~27% in the heated area at the end of 1995. The average backfill pressure has currently reached 18% of the initial vertical stress in the test field area, which has been estimated at ~12 MPa. Studies of water and gas releases from the backfill material reveal significant increases of carbon dioxide, methane, and hydrogen concentrations due to heating. In situ measurements will be continued in the coming years to study further thermomechanical reactions of the backfill and the surrounding rock salt to the heat input.

A Comparison of Long-Term Safety Aspects—Concepts for Disposal of Spent Fuel and Wastes from Reprocessing

Differences in technical concepts for direct disposal of spent fuel and for disposal of high-level waste (HLW) from reprocessing are discussed. The preferred emplacement sites for spent-fuel elements are drifts instead of boreholes, which are used for vitrified HLW. The nuclide inventories of uranium and plutonium are considerably higher with direct disposal. The impact of these conceptual differences on the long-term safety of a repository in a salt formation is investigated.The deterministically calculated radiation exposures for direct disposal and for disposal of reprocessed waste are both within the limits of the German licensing criterion. Furthermore, the differences in the radiation exposures are low, so from this point of view, neither concept is preferable. This result is surprising because the higher inventories of uranium and plutonium in the concept of direct disposal have a negligible influence on radiation exposure. It is shown that the layout temperature of a repository is a parameter influencing long-term safety, where higher layout temperatures are favorable.

An Assessment of the Results of the Direct Disposal Project by the Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety

The completion of the direct disposal of spent fuel project and thus the implementation of technical developments in the field of direct disposal are within reach. Politically, these developments are reflected in the 7th amendment of the Atomic Energy Act (the so-called Artikelgesetz), which gives safe disposal of spent-fuel elements the same level of priority as safe reprocessing.Since the 1960s, the Federal Republic of Germany pursued the concept of nonretrievable final disposal of radioactive waste in deep geological formations. The results of the direct disposal project show that there are no substantial changes required for this concept. Within the framework of the project, it was demonstrated that the radiological and mining-related safety objectives of final disposal in a mine can be realized. With the termination of the basic activities in the direct disposal project, specific questions about the feasibility of technical implementation were tackled in a comprehensive manner. However, this does not mean that all issues relating to direct disposal have already been resolved, citing for example the verification of long-term safety based on site-specific data with consideration of geochemical processes and the actual planning of emplacement areas for different types of waste (spent fuel, vitrified high-level waste, medium-activity waste, etc.). One topic area for further research and development relates to questions of gas generation and its control. Increased knowledge of the material behavior of rock salt and of crushed salt—as used for backfilling—is prerequisite for a site-specific assessment of geology and the long-term safety.

Thermal and Thermomechanical Analyses for Disposal in Drifts of a Repository in Rock Salt

In the last few years, several repository concepts have been developed for salt formations to dispose of both high-level radioactive waste from reprocessing and spent-fuel elements. The results of a series of thermal and near-field thermomechanical analyses for disposal in drifts at three horizons of a repository are described. The rise of the temperature in the emplacement area and the surrounding rock, the room closure of access and emplacement drifts during operational time, followed by the long-term compaction of the backfill material and the resulting stresses in rock salt, are investigated. Two numerical modeling procedures were used to obtain the results in this study. A computer code based on the closed-form solution for a heat source in a homogeneous medium was applied to predict the temperatures; a finite-element code, taking into account the nonlinear, temperature- and time-dependent behavior of rock salt and backfill material, was used to investigate the thermomechanical effects.

Electrochemical Decontamination of Metallic Surfaces Contaminated by Spent-Fuel Storage Pool Water

It was earlier shown theoretically that the radioactivity released by spent-fuel elements into storage pool water is predominantly carried by positive ions. In this paper a decontamination method is described in which freshly contaminated metallic surfaces are decontaminated electrochemically, resulting in smooth, shiny surfaces. This method, which uses current densities of {approximately}15 {mu}A/cm{sup 2}, is quantitatively and qualitatively different from earlier electrochemical procedures, where higher current densities of the order of milliamperes per square centimeter or even amperes per square centimeter were used.

Heat Transfer from Transport Cask Storage Facilities for Spent Fuel Elements

Interim storage plants for spent fuel elements, based on dry storage technology in transport casks, are planned in the Federal Republic of Germany. The casks are arranged in storage buildings. The decay heat is removed from the outer cask surfaces by natural convection of air entering the building through openings in the walls, and leaving through outlets in the roof. As the differential equations describing the complex three-dimensional flow and temperature field can only be solved for simple boundary conditions, experiments were conducted using scaled-down models of the casks and the building. The relevant similarity conditions have been investigated and used for design and operation of the 1:5 scale test setup. The cask models were heated electrically. Cask temperatures, air temperatures, as well as air flow and velocities, were measured. It was found that the cooling conditions at the different cask positions show very small differences and that the cask surface temperatures inside the building are a maximum of 10/sup 0/C higher than on a free-standing cask.

Hydrothermal Reactions of Clay Minerals and Shales with Cesium Phases from Spent Fuel Elements

AbstractThe immobilization of soluble Cs from spent fuel elements by ion exchange and direct chemical reaction with clay minerals or shales was investigated under hydrothermal conditions. Various clay minerals or shales were reacted with likely Cs sources and water at 300 bars pressure and 100°, 200°, and 300°C for 4, 2, and 1 months, respectively. Pollucite was the principal product, but CsAlSiO4 was also observed, along with unreacted or hydrothermally altered aluminosilicates. From Cs concentrations of the product solutions partition of Cs between liquid and solids was found to vary depending on the Cs source, the clay or shale phase, temperature, and run duration. For example, illite-Cs2MoO4 interactions resulted in 19, 32, and 95% fixation of added Cs at 100°, 200°, and 300°C respectively. Fixation of as much as 97% of the Cs in some solids was observed. In addition to Cs-aluminosilicates, Cs was fixed on cation-exchange sites by interlayer collapse in montmorillonite. Reactions with Cs2MoO4 also produced powellite because Ca was available in the reaction mixture. The U6+ from β-Cs2U2O7 was reduced to form uraninite by sulfide- and/or organic-rich shales. (Cs,Na)2(UO2)(Si2O5)3·4H2O, an analog of weeksite, was produced in reactions with β-Cs2U2O7. The reaction products pollucite and uraninite can immobilize much of the Cs and U from spent fuel elements because Cs in pollucite is extremely difficult to exchange and U in uraninite is insoluble.

Hydrothermal interactions of basalts with Cs and Sr of spent fuel elements: Implications to basalt as a nuclear waste repository

The immobilization reactions of Cs, Sr, U and Mo from spent fuel elements (SFE) in the presence of basalts and basalt phases were investigated under hydrothermal, closed-system conditions. The exact Cs and Sr phases in SFE are not known with certainty, but Cs 2MoO 4, CsI, Cs 2O, β-Cs 2U 2O 7 and SrZrO 3 are some of the predicted possibilities. These phases were mixed with basalts and basalt phases along with water, sealed in gold capsules and reacted at 100, 200 and 300°C under a confining pressure of 300 bars. The Cs and Sr concentrations of the product solutions were measured to determine the partitioning of these elements between the liquid and solid phases. Results varied substantially, depending on the particular basalt or basalt phase or Cs phase, the temperature and the duration of run. For example, a Deep Drill Hole (DDH-3) basalt-Cs 2MoO 4 reactions resulted in 15, 78 and 99% fixation of added Cs at 100, 200 and 300°C respectively. Fixation of up to 99.7 and 99.8% of the added Cs and Sr respectively were observed in some cases. X-Ray diffraction (XRD) analysis of the solid products revealed that Cs was immobilized by the formation of pollucite, (Cs, Na)AlSi 2O 6 in most cases and the formation of CsAlSiO 4 in some cases. The interactions of basalts and basalt phases with Cs 2MoO 4 also resulted in the immobilization of Mo by the formation of powellite, CaMoO 4 because Ca is available in the reaction mixtures. The U 6+ in β-Cs 2U 2O 7 was reduced to form uraninite, UO 2 by the divalent iron component of basalts. The exact nature of Sr reaction with basalts is not clear because no new Sr phases could be detected by XRD. The reaction products, pollucite, CsAlSiO 4, powellite and uraninite from SFE-basalt interactions serve to immobilize Cs, Mo and U because Cs in pollucite and CsAlSiO 4 is difficult to exchange and Mo and U in powellite and uraninite are quite insoluble.

Storage of Spent Fuel Elements

Storage of Spent Fuel Elements

Sloshing in Pools under Earthquake-Like Ground Motions

Spent-fuel elements from nuclear reactors are stored in pool structures while waiting to be reprocessed. Very large pools are being considered for future use and because of the large size, modularization into cells might be desirable. However, the effect of modularization on earthquake resistance needs to be investigated. Hydrodynamic pressures and water slosh-height are important considerations. Because of a lack of experimental data on large pools, the validity of using available hydrodynamic analytical techniques remains uncertain. Therefore, two experiments involving moderately large swimming pools near the Nevada Test Site were carried out during two high-yield, underground nuclear events. The results indicated that a linear hydrodynamic theory could be used and that damping in the water was nearly zero for the pool sizes studied. Spectral displacements at low frequencies govern the slosh and amplitude, but reported values at these frequencies, unfortunately, are not always accurate.