Dry Cask Research Articles

Accident mitigation for spent fuel storage in the upper pool of a Mark III containment

The upper pool (UP) is a specific design of the Mark III containments. Basically, no fuel can be stored in the UP when the reactor criticality is achieved. The Kuosheng plant in Taiwan has two BWR/6 units with Mark III containments. The spent fuel storage capacity of this plant will become insufficient in coming years. Unfortunately, the license application for the dry cask storage has not been approved. Taiwan Power Company has a temporary solution to store spent fuel with cooling time of 15years in the UP. Heat up mitigation for the UP by spray is also planned. In this study, the loss-of-coolant accident of the UP is analyzed using GOTHIC. The calculated results show that the pool spray can make up the UP inventory against small leakage. The spent fuel is sufficiently cooled if the spray mitigation maintains the pool level above the fuel. On the contrary, large leakage resulting in drainage of the entire pool allows airflow entering the fuel region to enhance the cooling effect. The case with the highest fuel cladding temperature occurs with a medium sized leakage because the partially uncovered fuels cannot be adequately cooled by either water or air. With 200-gpm spray, the calculated highest peak cladding temperature is 472.9°C which is well below the threshold causing significant fission product release. The capability of the spray mitigation to maintain the pool level determines whether the spent fuel will be heated up. Based on the results, the activation of spray can ensure the thermal safety of the spent fuel in the UP during a loss-of-coolant accident.

Preferential Water Ingress into a Dry Spent Fuel Cask

Water ingress into a fissile system, such as may occur during transportation of spent fuel casks, or in storage casks, may result in a significant increase in the keff of that system, which, if it were not accounted for in the abnormal operating or accident analysis of the cask, could potentially result in an inadvertent nuclear excursion with severe nuclear and radiological consequences. This paper will provide the results of an analysis of a gradual increase in water level in spent fuel casks and will discuss the way the keff of the system responds to such an increase. It will also provide the results of a preferential water ingress analysis, which sought to answer the question that if there is change in the keff of the system accompanied by a fractional increase in the amount of water in the cask, which one of the four water ingress pathways/channels would have the greatest effect on the keff. The water ingress pathways in question, and their respective unit numbers in brackets, are fuel rod (Unit 1), burnable poison rod (Unit 2), instrumentation tube (Unit 3) and cask air gap (Unit 9). This analysis will also determine which of the four channels has the highest sensitivity coefficient and will compare the effects of material composition on the keff when water is either freshwater or seawater.

ドライキャスクおよび使用済燃料プールラック向け新規中性子吸収材の開発

ドライキャスクおよび使用済燃料プールラック向け新規中性子吸収材の開発

地震動を受ける貯蔵架台付きキャスクのロッキングモデル化手法の提案

地震動を受ける貯蔵架台付きキャスクのロッキングモデル化手法の提案

Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry storage cask with hybrid heat pipes produces relatively low temperature inside a cask and reduces the possibility of structural failure due to thermal degradation.

Multidimensional simulations of hydrides during fuel rod lifecycle

In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

Seismic Response for a One-Third Scale-Down Vertical Cylindrical Cask Specimen Using Shaking Table Testing

This study is aimed to investigate the seismic behavior of the freestanding dry storage cask for spent fuel, several shaking table tests were conducted using a scaled cask model for a real assessment of the characteristics of the seismic response of the cask. First, the harmonic excitation test on the pedestal of the cask was performed to estimate the friction coefficient at the interface between the cask and the pad according to the sliding acceleration response of the pedestal. Then, tests for the seismic response of the cask were conducted for two different setups, the vertical cylindrical concrete cask (VCC) and the concrete cask with an add-on shield which has a square pedestal (VCC+AOS), respectively, using the artificial earthquakes compatible to the design spectrum. In order to verify the often-used analysis method for the seismic response of the cask in engineering practice, the explicit finite element software LS-DYNA was adopted to generate the finite element model of the scale cask with the cask/pad interface modeled by Coulomb’s law of friction and to simulate the shaking table tests. Results indicate that the utilized method gave reasonable cask responses if the variation of the friction coefficient at the cask/pad interface was small during the sliding process.

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.

Uncertainty Quantification of Concrete Utilized in Dry Cask Storage

The objective of this paper is to analyze the uncertainty in radiation dose rate estimates outside of a used nuclear fuel (UNF) dry cask storage unit due to the parametric variability of concrete compositions and densities. This requires the selection of a limited number of concrete compositions from a standardized database and the development of a reference dry cask model, which can be used to estimate dose rate from neutrons and gamma rays. The model was developed for use in calculations with MCNP, with reference data from a UNF assembly source and geometry details based on the Holtec HI-STORM 100S UNF dry cask storage system, both provided by the Comanche Peak nuclear power plant. A number of cases have been developed and analyzed to compare the effects of variations in concrete compositions considering nominal densities, variations from nominal densities, fixed densities regardless of the specific composition, and variations in the decay source. The analysis confirmed that the parametric variability of concrete compositions is a major source of uncertainty in evaluations of dry cask dose rates. While precise results depend on the compositions compared, general trends can be identified. The largest fraction of the dose value in all cases, typically 70%, is due to gamma rays produced by the fission products. Density variation had a dominant effect on the dose rate. Composition variations, while density was held fixed, indicated that the specific composition data significantly impact the dose rates produced by neutrons and associated capture gamma rays. The impact due to composition variation on neutron dose rate was found to be on the order of 70% or higher for these test cases. The analysis indicates that uncertainties in concrete characteristics at the time of on-site pour procedures impact the actual shielding efficiency and, therefore, must be evaluated.

Taking Out the Trash

The management of radioactive waste, particularly of High-Level Radioactive Waste (HLW) containing isotopes, whose half-life exceeds one million years, is a wicked and aporetic problem. The amount of waste increases continuously, while the question of management remains technologically and politically unsolved. Not only do the technological challenges involved exceed the horizon of scientists, but the ethical problems raised by the use of nuclear power have been neglected from the beginning. The history of nuclear power is as well a history of neglecting its consequences. No country is able to provide a suitable concept of storage for HLW. Instead, we have failing approaches: for example, decaying barrels in the Channel and the Atlantic Ocean, and in Germany two salt mines stuffed with unregistered amounts of waste on the edge of collapse.After the disaster at Fukushima Daiichi in March 2011, the German Federal Government decided for the second time to stop using nuclear power as an energy source, and on July 5th 2013, it passed a law mandating a new search for a storage site. In the beginning of 2013, an interdisciplinary project—ENTRIA—was launched to investigate options for radioactive waste management from the perspectives of the technical sciences and the humanities. ENTRIA is evaluating three possible options for storage: deep geological disposal, retrievable deep geological repository with monitoring, and permanent dry cask storage above ground. Each of the three options bears particular technological and ethical challenges.

Dry Storage Casks Monitoring by Means of Ultrasonic Tomography

Spent nuclear fuel (SNF) is one of the most hazardous types of nuclear power plant waste. This fact emphasizes the importance of careful handling and storage of SNF. There are two current state-of-the art technologies of SNF storage facility: wet and dry. It is important to mention that IAEA does not determine which kind of handling strategy should be chosen, however it is noted that dry storage of SNF could be used for one hundred years. Mining and Chemical Enterprise (MCE) is one of the leading Russian companies that deals exclusively with the dry storage of SNF. This company has implemented a long-term storage scheme. At the same time MCE faced the challenge of nondestructive monitoring of the degradation process of structural material of cask and its sealing with weld seam. Currently, X-ray testing is used for this purpose but in order to provide an effective nonradioactive method of monitoring MCE has initiated a collaborative R&D project with TPU supported by the Russian Government. Ultrasonic industrial tomography technique was proposed as the solution. The method is based on application of phased and sparse arrays transducer with real-time visualization algorithm. Received acoustic data is processed and realized by means of Sampling Phased Array technology which is a collaborative development of TPU and I-Deal Technology, GmbH. The multichannel ultrasonic set-up of immersion control was assembled for performing testing of seven experimental specimens with representative defects (side drill-holes, notches, natural welding flaws). X-ray tomography of high-resolution was chosen as the reference method. All indications were successfully reconstructed in B and C-scans and 3D image. The next step is to automate the monitoring procedure completely and to introduce an evaluation tool for current flaw state and prediction of its further behavior.

Why America should move toward dry cask consolidated interim storage of used nuclear fuel

Despite the recommendations of the 2012 Blue Ribbon Commission Report, the US government has made no substantial progress toward the permanent, or even temporary, consolidated storage of used1 civilian nuclear fuel. To complicate matters, a November 2013 decision by the United States Court of Appeals (2013) in Washington, DC eliminated the very fee designed to finance used-fuel storage—which had accumulated over $30 billion so far—introducing a further obstacle (Nuclear Energy Institute, 2014a).It was not supposed to be this way. The 1982 Nuclear Waste Policy Act bound the federal government by law to take custody of all civilian waste from power companies for final disposal, under the assumption that the waste would be permanently stored in a deep geological repository at Nevada’s Yucca Mountain and overseen by the Energy Department. The Act obligated the US government to begin accepting waste by 1998, but the government was unable to fulfill this promise, forcing it to remunerate the utility companies for continuing to store the waste and assume liability. This situation cannot continue.As a solution, the commission argues that the US government should establish a widely distributed series of regional, government-run sites that would take in the used fuel from the cooling pools of several reactors, thereby consolidating the interim storage of used fuel and putting this nuclear waste into stronger, safer, more secure, more manageable—and ultimately more affordable—dry casks, as a first step toward ultimate disposal. Dry casks have withstood earthquakes and floods, and are designed to withstand the heat of fires and the impact of airplanes; the 100-ton structures are hard to steal or damage, and require no cooling systems or power supplies. These are some of the many reasons why making the transition to dry cask-based interim storage should be made as quickly as possible, regardless of one’s opinion of civilian nuclear power.

Optimizing Laminar Natural Convection for a Heat Generating Cylinder in a Channel

Natural convection heat transfer from a horizontal cylinder is of importance in a large number of applications. Although the topic has a rich history for unconfined cylinders, maximizing the free convective cooling through the introduction of sidewalls and creation of a chimney effect is considerably less studied. In this investigation, a numerical model of a heated horizontal cylinder confined between two vertical adiabatic walls is employed to evaluate the natural convective heat transfer. Two different treatments of the cylinder surface are investigated: constant temperature (isothermal) and constant surface heat flux (isoflux). To quantify the effect of wall distance on the effective heat transfer from the cylinder surface, 18 different confinement ratios are selected in varying increments from 1.125 to 18.0. All of these geometrical configurations are evaluated at seven distinct Rayleigh numbers ranging from 102 to 105. Maximum values of the surface-averaged Nusselt number are observed at an optimum confinement ratio for each analyzed Rayleigh number. Relative to the “pseudo-unconfined” cylinder at the largest confinement ratio, a 74.2% improvement in the heat transfer from an isothermal cylinder surface is observed at the optimum wall spacing for the highest analyzed Rayleigh number. An analogous improvement of 60.9% is determined for the same conditions with a constant heat flux surface. Several correlations are proposed to evaluate the optimal confinement ratio and the effective rate of heat transfer at that optimal confinement level for both thermal boundary conditions. One of the main application targets for this work is spent nuclear fuel, which after removal from the reactor core is placed in wet storage and then later transferred to cylindrical dry storage canisters. In light of enhanced safety, many are proposing to decrease the amount of time the fuel spends in wet storage conditions. The current study helps to establish a fundamental understanding of the buoyancy-induced flows around these dry cask storage canisters to address the anticipated needs from an accelerated fuel transfer program.

An investigation on the seismic response of free-standing dry storage cask for spent fuel

For most dry storage facilities for spent fuel worldwide, storage casks are freestanding on concrete pads. Thus, relative motion between the cask and the pad may be induced during earthquakes, leading to stability concerns. In this study, the seismic stability conditions of a free-standing cask were examined quasi-statically. Then, a finite element (FE) cask-pad model considering the frictional contact at the interface was established, and a dynamic analysis was performed to simulate a scaled cask shaking table test, which was conducted at the Central Research Institute of Electric Power Industry in Japan. Furthermore, the influence of friction coefficient of the cask/pad interface on the cask response was investigated using this FE model. Apparent rocking of the cask was induced at a higher friction coefficient, while sliding dominated the cask motion at a lower one. Since the friction coefficient is critical to the seismic stability of a free-standing cask, cyclic loading tests of a scaled pedestal model of a cask system called the High Performance System of the Institute of Nuclear Energy Research (INER-HPS), which will be used in Taiwan, were conducted to investigate the actual friction coefficient at the cask/pad interface as well as the effect of normal stress and sliding rate. Based on the test results, numerical simulations were performed to ensure the seismic stability of this INER-HPS cask under the design base earthquake.

Quantification of U.S. spent fuel inventories in nuclear waste management

The goal of this research was to develop a methodology to collect inventory estimates for the analysis of spent fuel in the United States. Data sets were collected from publicly available information on 103 operating U.S. reactors over the lifespan of each unit. SCALE 6.1’s ORIGEN-ARP module was used to produce plant models for each reactor and inventory estimates for spent fuel removed 0, 1, 3, 5, 10, and 20years ago. The analysis includes assessments of radiotoxicity and heat load levels. Reductions in radiotoxicity and heat load were quantified for the oldest spent fuel, and it was noted that the values were smaller for BWR assemblies due to lower enrichment levels. The obtained results serve as a guiding tool in evaluating spent fuel management strategies, including spent fuel pool conditions, dry cask loading sequences of interim storage options, and expectations for permanent disposal needs.

Walking the Wall

Walking the Wall

Aging management and post-storage transport of dual purpose cask for spent fuel

AbstractAging management of spent fuel storage facility may follow lessons learned from literature for nuclear power plant and a review for spent fuel dry cask storage system by US NRC, DOE, by German BAM, that by Japan NISA, etc. Namely, the essence of systematic approach to aging management includes Understanding aging, Plan (Development and optimisation of activities for aging management), Do (Managing aging mechanisms), Check (Monitoring, inspection and assessment), and Act (Maintenance). The PDCA cycle will optimise the systematic approach to the aging management. An aging management programme (AMP) for the storage system over the period of extended storage will address uncertainties in the safety relevant functions of the system that may otherwise be impaired by aging mechanisms. The AMP identifies system, structure and components (SSCs) that need specific actions to mitigate aging and ensures that no aging effects result in a loss of their intended function of the SSCs, during an intended licensed ...

Comparisons of prediction methods for peak cladding temperature and effective thermal conductivity in spent fuel assemblies of transportation/storage casks

When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature. There are three models to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas’s two-region model, Bahney Lotz’s effective thermal conductivity model, and Wooton–Epstein correlation. The peak cladding temperatures of Babcock & Wilcox (B&W) 15×15 PWR spent fuel assembly under helium backfill gas were evaluated by using two-dimensional CFD simulation and compared with two models (Wooton–Epstein correlation, two-region model). The peak cladding temperature difference between the two-region model and CFD simulation ranges from −0.2K to 9K. Two-region model over-predicts the measured peak cladding temperature that performs in a spent fuel dry storage cask. Therefore the simulation could be used to calculate peak cladding temperature of spent fuel assemblies.Application using CFD simulation was conducted to investigate the peak cladding temperature and effective thermal conductivity of spent fuel assembly used in Korea NPPs: 16×16 (CE type) and 17×17 (WH type) PWR spent fuel assembly. CFD simulation results are similar to each other, and the difference of temperature drop between the three arrays occurs slightly in all basket wall temperatures. The effective thermal conductivity calculated from the 16×16 PWR spent fuel assembly results was more conservative than those for the 17×17 array.

Modeling of Particulate Behavior in Pinhole Breaches

A model is presented for calculating depressurization time for and particulate release from used nuclear fuel dry storage containers that have developed a pinhole breach. Particular attention is given to particulate deposition and transmission within the breach pathway. The model is modular in nature and is developed in a way that allows for more advanced treatments of internal temperature, internal component geometry, or aerosol flow to be readily incorporated. The model can be treated as a basis for addressing concerns associated with monitoring and verification efforts during long-term dry cask storage.

Analysis of processes in RBMK-1500 fuel rods during the operation, short and intermediate term storage

Abstract Recently the reactors of both units of Ignalina NPP in Lithuania were shutdown for decommissioning (in 2004 and 2009). According to the design, the spent fuel should be returned for reprocessing to Russia. However, up to now, all assemblies of spent fuel are still stored in the spent fuel pools and in the dry storage facility on-site of the Ignalina NPP. Thus, the safety justification for short and intermediate term spent fuel assemblies storage in Ignalina NPP is very important. The most important barrier, preventing release of radioactivity from the fuel matrix to the environment is the fuel rod cladding. The condition (integrity) of cladding at the end of intermediate storage of spent fuel assemblies may be evaluated by simulating processes in fuel rods during the whole “life cycle” of the fuel assembly: beginning from the first loading into the reactor core, until the end of dry intermediate storage in special casks. This paper presents the modelling of processes in fuel rods during normal operation of fuel assemblies in reactor, short term wet storage in spent fuel pool and intermediate storage in dry cask. The analysis was performed using FEMAXI-6 integral code for the analysis of processes in fuel rod. The behaviour of thermal (pressure inside fuel rod, temperatures of fuel pellets and cladding, etc.) and mechanical (change of the gap between pellets and cladding, stresses and plastic deformation in fuel pellets and cladding, etc.) parameters were calculated. The analysis performed demonstrates the possibility to describe the state of the fuel rods after normal operation and short and intermediate term fuel storage.