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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

TN{sup R} MW- Multi Waste Package 'All In One Solution' For Conditioning - Transport - Storage - Up To Final Disposal - 17196

For the management of radioactive waste, either during nuclear facility operation or decommissioning, one must consider a diverse array of nuclear waste forms of various types, volumes and activities. Examples are: activated metallic waste; legacy waste; sludge; resins; and orphan waste. Customers are frequently faced with the obligation of undertaking multiple, costly waste management operations. These operations activities include handling, reconditioning or material transfers between packages, as in the case when moving material from on-site storage to transportation, and then from transportation to final disposal. To address this issue, a new - highly flexible up to reversible - cask system, the TN{sup R} MW, is being developed as an 'All in One Solution'. TN{sup R} MW has a total weight of 10 metric tons and is compliant with the 2012 International Atomic Energy Agency (IAEA) regulations. It was developed to provide flexibility and designed with the capability to be adaptable to the various needs of the nuclear industry. From IP-2 and Type A containers to Type B(U) and B(U)F casks, this cask offers different options to sufficiently address the requirements of each step of the waste stream from on-site and international transport to long-term interim storage and final disposal. Licensing and manufacturing of a number of packagings in this TN{sup R} MW family are underway. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Neutron radiation shielding material polyethylene: consequences of gamma irradiation - 15062

High and ultra-high molecular weight polyethylenes are high performance materials, which are predestined for a wide range of applications due to characteristics like high density, low weight, good dimensional stability, high chemical resistance, and high hydrogen content. These polymers are used in two demanding areas: as implant material in medical technology (only ultra-high molecular weight polyethylenes) and as a component for neutron shielding purposes in casks for storage and transport of radioactive waste. In the medical field (joint replacements) as well as during neutron radiation shielding application, high and ultra-high molecular weight polyethylenes are exposed to gamma irradiation: in the first case requested as sterilization process and for surface-crosslinking and in the second case existing as a side effect of inserting the radioactive material in the cask. Given that polyethylene as shielding material has to withstand any type of degradation affecting safety relevant aspects to be applicable for long-term radiation shielding purposes for instance over a period of 40 years, the durability of the material is of special interest. High molecular weight polyethylene (HMW-PE; LUPOLEN 5261Z; Lyondell Basell) and ultra-high molecular weight polyethylene (UHMW-PE; GUR 4120; Ticona) were subjected to gamma radiation and afterwards thermally treated. The gamma doses used are in the range of 50 to 600 kGy and irradiation takes place at RT using a Co-60 source. The planned thermal treatment will take place at a temperature of 125 deg.C for periods of minimum 30 days and maximum of five years. With the applied conventional analytical techniques it is possible to detect structural changes of both types of polyethylene induced by gamma irradiation and certainly of thermal treatment. Through gamma irradiation melting temperature, crystallinity, and density, respectively increased. Furthermore both polyethylenes get oxidized and cross-linked. With regard to the special application as neutron radiation shielding material in casks for storage and transport of radioactive materials, the impact of irradiation lead to changes of material properties. A consolidated view indicates that the detected changes of the irradiated (U)HMW-PE are not safety relevant for long-term neutron radiation shielding purposes over a period of 40 years in Germany. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

High-level waste relocation requires use of new technology in aging nuclear facility - 15269

The West Valley Demonstration Project (WVDP) is preparing to relocate 278 canisters of vitrified high-level waste (HLW) from their current location inside the Main Plant Process Building (MPPB) to an outdoor pad for long-term, on-site storage. The waste was produced at the WVDP between 1996 and 2002 and was originally placed in interim storage inside the MPPB in anticipation of off-site shipment for disposal. Relocation of the canisters to a new on-site location is now necessary as part of WVDP Phase 1 Site Decommissioning work to enable pre-demolition activities inside the MPPB to proceed. The logistics of packaging and moving the HLW canisters within an aging facility to another on-site location has provided numerous technical challenges with subsequent innovations. Significant preparations have taken place to resolve structural inadequacies, including sub-surface soil subsidence issues, antiquated equipment, and facility reconfigurations to support up to 134 metric ton (148 ton) loads. To perform the transfer and storage of the HLW canisters, new equipment and fabrications were required including a robust storage system consisting of stainless steel canister overpacks within steel-reinforced concrete storage casks, three specialty cask transporters, a remotely operated welding system and construction of an outdoor engineered storage pad to store the 56 casks containing the relocated HLW canisters. Packaging and relocation of vitrified canisters is scheduled to begin in 2015. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Experimental Seismic Response of Scaled Storage Containments under Identical Loading Conditions - 16318

Dry Storage Casks (DSCs) store spent nuclear fuel (SNF), usually at sites contiguous to nuclear power plants known as Independent Spent Fuel Storage Installation (ISFSI). DSCs have been considered as a temporary storage solution, and are usually licensed for 20 years, which can be extended up to an operating periods of 60 years. Recently, however, DSCs have been considered as a potential mid-term solution for hundreds of years. This consideration requires reevaluation of DSC performance under a larger seismic hazard. That leads to larger horizontal and vertical accelerations. This study is a part of a research project evaluating the seismic response of free-standing DSCs under long return period seismic events. This article assess whether the response of free-standing structures under seismic excitation is repeatable. For this purpose, experimental tests were conducted using a six degree-of-freedom (6DOF) shake table. During the experimental tests, several ground motions were repeated multiple times to obtain the dynamic response under identical loading conditions. Scaled casks of four aspect ratios (radius to centroidal height ratio, r/h{sub cg}) were studied: 0.39, 0.43, 0.55 and 0.62. The specimens studied are considered to be 1:2.5 and 1:3.5 scaled model of generic prototype casks. The experimental results show that a small change in initial conditions leads to change in the boundary condition of moving bodies and in combination with acceleration being applied at different frequencies contained within the applied ground motions may lead to a large variation in the response. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Simulated Gamma Doses to Coating Materials for Use in High Level Waste Packages

An outer shield barrier for spent nuclear fuel is being developed in order to improve dry cask performance under mechanical, chemical, and radiological degradation and stresses. The materials proposed for the outer shield barrier are titanium nitride (TiN), a composite oxide layer consisting of ZrO{sub 2}, TiO{sub 2}, and Al{sub 2}O{sub 3}, and a final layer of molybdenum disulfide (MoS{sub 2}). The ultimate aim of this research is to develop a multi component shield material either as multiple layers or a multi material composite that will improve the performance of steel used in long-term storage casks. Two approaches are being examined through this research effort: multilayer and composite. The composition of the oxide layer is crucial to optimizing the shield's properties and performance. At equal oxide compositions the mixture of these specific oxides is commonly known as zirconolite. This material has been shown to provide a water barrier and to provide substantial corrosion resistance to the dry cask. In order to optimize the radiological performance of this layer the compositions of each oxide have been varied and analyzed for chemical, mechanical, and radiological effects on performance. Through the course of this research effort materials performance changes after gamma and neutron exposure as well as gamma and neutron attenuation of the potential shield materials have been measured and correlated to coating composition. As it is particularly difficult to design a novel, five component material with the goal of improving corrosion, wear behavior, heat tolerance, radiation tolerance, and general survivability, it is necessary to understand which coating components contribute to each property enhancement, and which are critical to design and eventual coating survivability. To compliment the experimental research effort, a study of gamma radiation doses using the relevant zirconolite constituent oxides and assorted combinations off all three, ZrO{sub 2}-TiO{sub 2}-Al{sub 2}O{sub 3}, has been conducted in order to gain insight into which, if any, ratios of the three constituent oxides may provide optimal gamma radiation shielding. Analyzing and comparing the oxides to each other it is apparent that the outer shield barrier with no zirconia has the highest absorbed dose rate at 4.88 E6 mrad/hr. The simulations run with no alumina and no titania are remarkably similar in absorbed dose rates. Alumina and titania compositions increase the absorbed dose rate increase. This is to be expected as during these simulations the zirconia composition is decreasing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Apparent Monte Carlo Source Convergence Problem with BWR Fuel Depleted with Partial Control Blade Insertion

Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate application of burnup credit (BUC) to boiling-water reactor (BWR) fuel in storage and transportation casks and spent fuel pools (SFPs). Ideally, a single combination of all operating parameters including moderator density and axial burnup profiles, control blade exposure, and others can be identified to allow a simple, bounding approach to BWR BUC. One aspect of this investigation involves examining the effect of control blade insertion during operation on discharged assembly reactivity. In the course of these calculations, an apparent source convergence problem was discovered for fuel that has experienced partial blade insertion for the entire lifetime of the assembly. The source convergence problem appears not to impact k{sub eff}, but it manifests in predictions of the axial fission density profile. This phenomenon is of interest in identifying areas of concern with BWR BUC since it has not been studied as thoroughly as BUC for pressurized-water (PWR) reactor fuel. There may also be interest in this configuration for further study of a realistic source Monte Carlo convergence problem as opposed to some earlier manufactured source convergence problems through careful selection of initial conditions. This summary presents the depletion conditions used to generate the problem, a description of the models, a background based on well-behaved calculations, the identification of the problem, and attempts made to resolve the problem. Only a brief description of the depletion conditions and models is possible within the limited scope of this summary. Finally, conclusions are presented. This summary identifies a source convergence problem in the process of investigating BUC for BWR spent fuel. The depletion scenario which created the depleted fuel and burnable absorber compositions included a control blade inserted in the bottom 14 nodes throughout the entire depletion. This insertion is slightly over half the length of the assembly, and it appears to create two regions of nearly equal reactivity. As expected, one region is near the top of the assembly, and the other region is near the middle of the assembly. This node includes significant {sup 239}Pu generation as a result of rodded operation, but it is also the top node of the full lattice zone. While the calculated k{sub eff} values appear to be statistically equivalent, the axial neutron production profile varies significantly among independent simulations of the same model. A range of standard approaches has been used to attempt to resolve the source convergence problem and determine the true axial neutron production profile. This profile is of interest in cask criticality safety analysis because it influences the interaction among assemblies loaded in the cask. The forthcoming SCALE 6.2 release will include source convergence testing based on Shannon entropy, which may assist in identifying these types of problems in the future. It should be noted that this source convergence problem results from the depletion conditions and the final distribution of fissile and absorber material within the discharged fuel assembly. The results, while entirely generated within the SCALE code system, are representative of any Monte Carlo program attempting to calculate k{sub eff} for such a system. The most important conclusion of this work is that the BUC criticality safety community must be aware of the potential for this type of source convergence problem. Analysts must approach such partially rodded scenarios with caution and critically examine the results. Fortunately, it is extremely unlikely that BWR assemblies would experience the bladed operation necessary to create this particular source convergence problem. These results may also have some bearing on the analysis of PWR BUC for new plant designs such as AP1000 that incorporate significant rodded operations. In a broader context, the source convergence problem identified here could have been magnified if it had occurred in a Monte Carlo depletion simulation. It may be possible to identify this type of problem in a depletion calculation by comparing the apparently converged fluxes between steps to ensure that no oscillations or other serious shifts are occurring. The depletion calculations performed in this work were performed with a two-dimensional deterministic code and are therefore not subject to this problem. These results may also raise questions about the validity of 2D depletion models with extruded geometry in the 3D k{sub eff} calculation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Corrosion Testing of TiN on Steel Substrates for use in Waste Packages

An outer shield barrier for spent nuclear fuel casks is being developed in order to improve dry cask performance under mechanical, chemical, and radiological burdens. The materials proposed for the outer shield barrier are titanium nitride (TiN), a composite oxide layer consisting of ZrO{sub 2}, TiO{sub 2}, and Al{sub 2}O{sub 3}, and a final layer of molybdenum disulfide (MoS{sub 2}). A proposed multilayer and multicomposite approach are being investigated in order to optimize the outer shield barriers performance. Spent nuclear fuel casks are frequently made, in part, of 304 and 316 stainless steels. These materials have been chosen as substrates to deposit single layer films of the proposed materials. In order to understand how these materials will behave layered and composited it is first necessary to understand how they perform individually. Several corrosion studies have been performed and analyzed for cyclic corrosion and static corrosion in different brine solutions and at different temperatures. The corrosion rate, over potential, activation enthalpy, and entropy of corrosion have been documented in previous publications. One of the main functions of the TiN coating is to serve as a corrosion barrier on the waste package. In this study we investigate pitting corrosion on thin layers of TiN through cyclic polarization testing and scanning electron microscopy. The corrosion protection of thin layers of TiN has been investigated on 304 and 316 stainless steels. Cyclic polarization testing has been used to this end. Corrosion rates have been calculated using tefel fits on data obtained from this testing and SEM images were used in conjunction with this data. Pitting is observed to occur even at the thickest deposition conditions tested. Although the corrosion rates decrease at these thicker values - sometimes dramatically - thicker layers will be deposited in the future to provide complete resistance to pitting. Deposition and testing of these layers is currently underway. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2020

An Experimental Feasibility Study on Applying Neutron Tomography to Encapsulated Spent Nuclear Fuel - 20050

Visual inspection makes easier to ensure the integrity and safety of spent nuclear fuel (SNF) than any classical techniques. Various classical techniques have been applied but there are no reliable methods to qualitatively and quantitatively verify spent fuel in dry storage. Thus, the present authors have developed the prototype safeguards apparatus for dry storage employing the array of He-4 gas scintillation detectors (S670E, Arktis Radiation Detectors Ltd., Switzerland), newly designed to simultaneously measure thermal and fast neutrons without any moderators. The S670E detector has a cylindrical shape with a diameter of 52 mm and active length of 600 mm (total length: 875 mm). The detector is filled by He-4 gas with an approximate pressure of 180 bar for fast neutron detection, and its inner wall is coated by Li-6 for thermal neutron detection. The scintillation lights generated via Li-6 nuclear reaction and elastic scattering are collected by 24 SiPMs linearly paired at the center of the detector. The detector delivers a TTL (Transistor-Transistor Logic) output for pulse readout and UART (Universal Asynchronous Receiver Transmitter) for device control. In order to assess feasibility of the apparatus, an experimental system has been designed, built, and optimized via computational studies. Cf-252 neutron sources and linearly arrayed detectors, working as a single detector, were occupied for this study due to the difficulties in working with SNF. The laboratory scale cask (diameter: 0.67 m, height: 1.5 m), minimized by a factor of 10 compared to the actual thickness of a commercial TN-32 cask, was also manufactured. The detector array was designed to rotate the lab-scale cask and obtain 36 image profiles at every 10 degrees. All profiles were aligned in single frame image called a sinogram, and the cross-sectional image was then fabricated by the inverse radon transform algorithm. These experiments have been repeated with different configurations and numbers of sources. Some gamma-ray sources were also measured with neutron sources in order to distinguish between neutron and gamma-ray pulses. Basically, a He-4 detector is designed to run on Linux OS so it is difficult to directly apply to Windows-based equipment widely used in S. Korea. Therefore, a new data acquisition board working on Windows OS was designed and built. The board mainly consists of FPGA (Field Programmable Gate Array) and SoC (System on Chip) for TTL pulse readout, sorting measured data, and transferring data to a user interface. In conclusion, the tomographic system at lab-scale has shown considerable potential to detect a partial or gross defect of encapsulated assemblies in dry storage. Next steps of this study will be to 1) repeatedly carry out experiments to demonstrate scientific reliability and validity, and 2) numerically integrate signals with weight factors to enhance the image quality since the suggested system based on passive interrogation method requires longer measurement time. Finally, the system will apply to a commercial dry storage phased out soon in S. Korea. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Addressing fission product validation in MCNP burnup credit criticality calculations - 14508

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members' review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k{sub eff}) calculations crediting the presence of a limited set of fission products and minor actinides (FP and MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP and MA worth - in addition to bias and bias uncertainty resulting from validation of k{sub eff} calculations for the major actinides in SNF - to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP and MAs. The ISG recommends (1) use of 1.5% of the FP and MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP and MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP and MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP and MA worth bias is shown to be acceptable by comparison of FP and MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII-based nuclear data. The comparison supports use of the 1.5% FP and MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP and MA worth is no more than 0.1 Δk{sub eff} (ISG-8, Rev. 3, Recommendation 4). (authors)

42 ENGINEERING↗
DOE OSTI2015

Regulatory perspective of boiling water reactor peak reactivity credit in spent fuel storage and transportation - 14592

The U.S. Nuclear Regulatory Commission (NRC) performs reviews of applications for spent nuclear fuel transportation packages and storage casks to ensure pertinent safety regulations are met. Among other acceptance criteria, these regulations require that the package be subcritical under a variety of normal, off-normal and accident conditions. Currently, all spent nuclear fuel transportation packages and storage casks assume fresh fuel (i.e., the fuel is unirradiated) in their criticality safety analyses for boiling water reactor (BWR) fuels. Reviews using this conservative assumption are less complex. The fresh fuel assumption for BWR fuel does not allow the presence of gadolinium (often called the 'fresh fuel no-Gad' assumption), an integral burnable absorber present in nearly all BWR fuel assemblies. More recently, applicants began requesting credit for the reactivity reduction due to depletion. This credit is commonly referred to as 'burnup credit' (BUC). As of today, NRC recommends that only analyses for Pressurized Water Reactor (PWR) spent fuel implement BUC. This is due to the fact there is limited directly applicable data available to benchmark codes for depletion and reactivity calculations for BWR BUC analyses, and determining the most reactive irradiation conditions is much less straightforward than for PWR BUC analyses. Similar to what was done for PWR BUC, the NRC staff and its contractors have begun identifying and prioritizing significant technical issues so that a technical basis for the allowance of BWR BUC can be developed. BWR fuel exhibits a peak in reactivity where initial gadolinium is depleted. Crediting burnup at this point is generally referred to as using a 'peak reactivity method'. This is a conservative approach and a method routinely used in BWR spent fuel pool criticality analyses. NRC is implementing a two-phased approach in investigating BWR BUC. Phase 1 investigates peak reactivity credit and Phase 2 evaluates BUC at a typical discharge exposure. This paper discusses peak reactivity credit and its history in regulatory analyses, challenges the NRC expects in reviewing an application requesting peak BWR reactivity credit for a spent fuel transportation package or storage cask, the work that is being performed by NRC contractors to address these challenges, and future work on peak reactivity credit and full BWR BUC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Advancing the Fork detector for quantitative spent nuclear fuel verification

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations. A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. Finally, the results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

A sensing material-free and simple readout MEMS sensor for detecting Helium

Abstract In this work, we report a method that enables a standard electrostatic MEMS device to perform complex sensing functionalities, such as detecting the presence of helium without a sensing material or a conditioning circuit. Helium is a noble, odorless, non-reactive gas that is very challenging to detect. It is used in critical applications such as storing nuclear fuel waste inside a dry cask. In these applications, its leakage from the dry cask may indicate the cask's safe operation's degradation. A departure from the common practice of exciting the MEMS around its mechanical resonance, the method is based on exciting the MEMS around its electrical resonance circuit. This method shows that the tiny difference between the air dielectric constant (1.00059) and helium (1.000067) corresponding to only a few Femtofarad level capacitances produces a 25 mV difference without a conditioning circuit. Simulation results confirmed those findings and explored the sensor response at different operation conditions. This method eliminates the need for a heated microstructure and the need for absorption material. This method is not limited to gas sensing. It can be applied to other sensing mechanisms, such as acceleration and pressure measurements, and eliminate the complex circuit to read small capacitance in these applications.

Mohaidat, Sulaiman↗
DOE OSTI2013

PACKAGE IMPACT MODELS AS A PRECURSOR TO CLADDING ANALYSIS

The evaluation of spent nuclear fuel storage casks and transportation packages under impact loading is an important safety topic that is reviewed as part of cask and package certification by the United States Nuclear Regulatory Commission. Explicit dynamic finite element models of full systems are increasingly common in industry for determining structural integrity during hypothetical drop accidents. Full cask and package model results are also used as the loading basis for single fuel pin impact models, which evaluate the response of fuel cladding under drop conditions. In this paper, a simplified package system is evaluated to illustrate several important structural dynamic phenomena, including the effect of gaps between components, the difference in local response at various points on a package during impact, and the effect of modeling various simplified representations of the basket and fuel assemblies. This paper focuses on the package impact analysis, and how loading conditions for a subsequent fuel assembly or fuel cladding analysis can be extracted.

Klymyshyn, Nicholas A.↗
DOE OSTI2013

Safety aspects of dry spent fuel storage and spent fuel management

As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. TUV and BAM, who work as independent experts for the competent authorities, present the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All safety relevant issues like safe enclosure, shielding, removal of the decay heat or behavior of cask and building under accident conditions are checked and validated with state-of-the-art methods and computer codes before the license approval. It is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Numerical Estimation of the Spent Fuel Ratio

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations

Radial hydride formation in high-burnup used fuel cladding has the potential to radically reduce its ductility and suitability for long-term storage and eventual transport. To avoid this formation, the maximum post-reactor temperature must remain sufficiently low to limit the cladding hoop stress, and so that hydrogen from the existing circumferential hydrides will not dissolve and become available to re-precipitate into radial hydrides under the slow cooling conditions during drying, transfer and early dry-cask storage. The objective of this research is to develop and experimentallybenchmark computational fluid dynamics simulations of heat transfer in post-pool-storage drying operations, when high-burnup fuel cladding is likely to experience its highest temperature. These benchmarked tools can play a key role in evaluating dry cask storage systems for extended storage of high-burnup fuels and post-storage transportation, including fuel retrievability. The benchmarked tools will be used to aid the design of efficient drying processes, as well as estimate variations of surface temperatures as a means of inferring helium integrity inside the canister or cask. This work will be conducted effectively because the principal investigator has experience developing these types of simulations, and has constructed a test facility that can be used to benchmark them.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Management of the Cs/Sr Capsule Project at the Hanford Site. Technology Readiness Assessment Report

The Federal Project Director (FPD) for the U.S. Department of Energy (DOE), Richland Operations Office (RL) Waste Management and D&D Division (WMD) requested a Technology Readiness Assessment (TRA) for the Management of the Cesium/Strontium Capsule Storage Project (MCSCP) at the Waste Encapsulation and Storage Facility (WESF) on the Hanford Site in Washington State. The MCSCP CD-1 TRA was performed by a team selected in collaboration between the Office of Environmental Management (EM) Chief Engineer (EM-3.3) and RL, WMD FPD. The TRA Team included subject matter and technical experts having experience in cask storage, process engineering, and system design who were independent of the MCSCP, and the team was led by the Director of Operations and Processes from the EM Chief Engineer's Office (EM-3.32). Movement of the Cs/Sr capsules to dry storage, based on information from the conceptual design, involves (1) capsule packaging, (2) capsule transfer, and (3) capsule storage. The project has developed a conceptual process, described in 30059-R-02, "NAC Conceptual Design Report for the Management of the Cesium and Strontium Capsules Project", which identifies the five major activities in the process to complete the transfer from storage pool to pad-mounted cask storage. The process, shown schematically in Figure 1, is comprised of the following process steps: (1) loading capsules into the UCS; (2) UCS processing; (3) UCS insertion into the TSC Basket; (4) cask transport from WESF to CSA and (5) extended storage at the CSA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Consequence Analysis of Residual Water in a Storage Canister (Preliminary Report)

The objective of this work is to examine the consequences of a residual water on change in condition of cladding and SNF during dry storage. Small amounts of residual water may remain in SNF dry storage canisters after the fuel assemblies are transferred to the canister in the spent fuel pool (SFP), the canister is removed from the SFP and drained or pumped, vacuum dried in several pressure reducing steps, and backfilled with helium. This residual water from incomplete drying could degrade fuel rod cladding, oxidize exposed fuel pellets, and cause cladding failure resulting in SNF fines’ release in canister cavity. If residual water content is high enough, it could corrode internal components inside the canister, e.g., fuel basket, and neutron absorber plates, but CNWRA work indicated that corrosion of the internal component is expected to be negligible even with 55 moles of residual water. Considering this, focus of this work to determine partitioning of residual water towards fuel and cladding oxidation. Residual water will decompose by radiolysis and create reactive oxygen species, which, at sufficient high temperatures will oxidize cladding, and UO 2 exposed by clad breaches. If a significant fraction of UO 2 in a fuel rod is oxidized to U 3 O 8 , swelling of the fuel pellets will rupture the cladding and release fuel particles, contaminating the inside of the canister and complicating retrieval of fuel assemblies, transport and ultimate disposal. The report describes an evaluation of the potential cladding and exposed SNF pellets due to residual water. An integrated quantitative approach was used to estimate the effects. The approach consisted of (i) the evolution of environmental conditions, i.e., temperature and radiation field, in the cask, (ii) the physicochemical processes that affect degradation of the materials within the cask, and (iii) the potential damage states that may result from these conditions and processes. The assessment considers the following set of cask environmental conditions: (i) the quantity of residual water after drying, (ii) internal temperature fields and their intensity decrease over time, and (iii) the strength of the internal radiation field. The quantity of residual water will affect the mass of cladding, SNF, and other components that are degraded by oxidation. Temperature controls the rate of the oxidation reactions, the phase composition of water, and, to a lesser extent, the rate at which water vapor is radiolyzed. The strength of the radiation field controls the radiolysis rate of residual water. These processes, i.e., time evolution of temperature fields and their effects on radiolysis of residual water, fuel and cladding oxidations, were modeled using the integrated approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗