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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

This report presents a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown nuclear power plant sites. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; a description of the on-site infrastructure and conditions relevant to transportation of UNF and GTCC waste; an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF and GTCC waste, including identification of gaps in information; and, an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste. The primary sources for the inventory of UNF and GTCC waste are the U.S. Department of Energy (DOE) RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included observations and information collected during visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Spent Nuclear Fuel Transportation: Rail Conveyance Transmissibility

Equipos Nucleares, S.A. (ENSA) and the US Department of Energy (DOE) are preparing a full scale Spent Nuclear Fuel (SNF) transportation test. This transportation testing will include road, rail, coastal, and trans-Atlantic shipments. The test campaign will use a full scale commercial transportation package and cradle. The package will be loaded with two instrumented fuel assemblies to measure strains at cladding locations and accelerations on the fuel assemblies, and thirty “dummy” assemblies. This testing is designed to closely match an actual SNF shipment. Accelerometers will also be used at various locations throughout the full conveyance system (i.e. rail car/truck, cask, transport cradle, and basket) to study the transmission of loads through the system and to provide validation for numerical models. Previous testing and modeling work has shown how the structural transmissibility of the transport system can affect the magnitude of these loads, and the importance of modeling all aspects of the transport system (i.e. rail car/truck, transport cradle, cask, basket, and fuel) (Ref 1, 2). This paper describes preliminary models that were constructed to estimate load transmission during rail transport, from the bottom of the cradle to an individual fuel rod within the package. The modeling studies in this paper evaluate the system response to postulated shock pulses and random vibration loads. These models describe the transmissibility of the conveyance, and demonstrate how loads can be amplified or attenuated as they are transmitted through the structure. This is done by coupling the rail vehicle dynamics code NUCARS, to the general finite element modeling code ANSYS, and the explicit dynamics code LS-DYNA. Models such as the ones presented herein will be used during the test campaign to help analyze and evaluate the test data as it is collected.

Cladding, Spent Fuel and Radiactive Waste, Used Fu↗
DOE OSTI2013

Demonstrating the Safety of Long-Term Dry Storage - 13468

Commercial nuclear plants in the United States were originally designed with the expectation that used nuclear fuel would be moved directly from the reactor pools and transported off site for either reprocessing or direct geologic disposal. However, Federal programs intended to meet this expectation were never able to develop the capability to remove used fuel from reactor sites - and these programs remain stalled to this day. Therefore, in the 1980's, with reactor pools reaching capacity limits, industry began developing dry cask storage technology to provide for additional on-site storage. Use of this technology has expanded significantly since then, and has today become a standard part of plant operations at most US nuclear sites. As this expansion was underway, Federal programs remained stalled, and it became evident that dry cask systems would be in use longer than originally envisioned. In response to this challenge, a strong technical basis supporting the long term dry storage safety has been developed. However, this is not a static situation. The technical basis must be able to address future challenges. Industry is responding to one such challenge - the increasing prevalence of high burnup (HBU) used fuel and the need to provide long term storage assurance for these fuels equivalent to that which has existed for lower burnup fuels over the past 25 years. This response includes a confirmatory demonstration program designed to address the aging characteristics of HBU fuel and set a precedent for a learning approach to aging management that will have broad applicability across the used fuel storage landscape. (authors)

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DOE OSTI2013

Pressure Build-Up During the Fire Test in Type B(U) Packages Containing Water - 13280

The safety assessment of packages for the transport of radioactive materials with content containing liquids requires special consideration. The main focus is on water as supplementary liquid content in Type B(U) packages. A typical content of a Type B(U) package is ion exchange resin, waste of a nuclear power plant, which is not dried, normally only drained. Besides the saturated ion exchange resin, a small amount of free water can be included in these contents. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific issues. An overview of these issues is provided. The physical and chemical compatibility of the content itself and the content compatibility with the packages materials must be demonstrated for the assessment. Regarding the mechanical resistance the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could for example be caused by radiolysis of the liquid and must be taken into account for the storage period. If the package is stressed by the total inner pressure, this pressure leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The total pressure in any calculation is the sum of partial pressures of different gases which can be caused by different effects. The pressure build-up inside the package caused by the regulatory thermal test (30 min at 800 deg. C), as part of the cumulative test scenario under accident conditions of transport is discussed primarily. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from beginning of the thermal test until cooling-down. In this case, while calculating the temperature distribution, conduction and radiation as well as evaporation and condensation during the associated process of transport have to be considered. This paper discusses limiting amounts of water inside the cask which could lead to unacceptable pressure and takes into account saturated steam as well as overheated steam. However, the difficulties of assessing casks containing wet content will be discussed. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. (authors)

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DOE OSTI2013

Stakeholder Transportation Scorecard: Reviewing Nevada's Recommendations for Enhancing the Safety and Security of Nuclear Waste Shipments - 13518

As a primary stakeholder in the Yucca Mountain program, the state of Nevada has spent three decades examining and considering national policy regarding spent nuclear fuel and high-level radioactive waste transportation. During this time, Nevada has identified 10 issues it believes are critical to ensuring the safety and security of any spent nuclear fuel transportation program, and achieving public acceptance. These recommendations are: 1) Ship the oldest fuel first; 2) Ship mostly by rail; 3) Use dual-purpose (transportable storage) casks; 4) Use dedicated trains for rail shipments; 5) Implement a full-scale cask testing program; 6) Utilize a National Environmental Policy Act (NEPA) process for the selection of a new rail spur to the proposed repository site; 7) Implement the Western Interstate Energy Board (WIEB) 'straw man' process for route selection; 8) Implement Section 180C assistance to affected States, Tribes and localities through rulemaking; 9) Adopt safety and security regulatory enhancements proposed states; and 10) Address stakeholder concerns about terrorism and sabotage. This paper describes Nevada's proposals in detail and examines their current status. The paper describes the various forums and methods by which Nevada has presented its arguments and sought to influence national policy. As of 2012, most of Nevada's recommendations have been adopted in one form or another, although not yet implemented. If implemented in a future nuclear waste program, the State of Nevada believes these recommendations would form the basis for a successful national transportation plan for shipments to a geologic repository and/or centralized interim storage facility. (authors)

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DOE OSTI2013

Dynamic Impact Analyses and Tests of Concrete Overpacks - 13638

Concrete cask is an option for spent nuclear fuel interim storage which is prevailingly used in US. A concrete cask usually consists of metallic canister which confines the spent nuclear fuel and concrete overpack. When the overpack undergoes a severe missile impact which might be caused by a tornado or an aircraft crash, it should sustain acceptable level of structural integrity so that its radiation shielding capability and the retrievability of canister are maintained. Missile impact against a concrete overpack involves two damage modes, local damage and global damage. Local damage of concrete is usually evaluated by empirical formulas while the global damage is evaluated by finite element analysis. In many cases, those two damage modes are evaluated separately. In this research, a series of numerical simulations are performed using finite element analysis to evaluate the global damage of concrete overpack as well as its local damage under high speed missile impact. We consider two types of concrete overpack, one with steel in-cased concrete without reinforcement and the other with partially-confined reinforced concrete. The numerical simulation results are compared with test results and it is shown that appropriate modeling of material failure is crucial in this analysis and the results are highly dependent on the choice of failure parameters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A Cable Car Transfer System for the Cigeo Access Ramps - An Original Approach - 17035

Andra is currently implementing the detailed engineering phase of the step-wise design of Cigeo (the French HLW and IL-LLW Deep Geological Repository). Various technical issues are at stake, including the definition and development of a key component: a funicular ('cable-car') system dedicated to the transfer (down the access ramp and leading to the underground infrastructures) of the shielding casks housing the waste disposal containers (packages). The present paper starts with a brief description of the Cigeo architecture (i.e. the surface utilities, the service and nuclear ramps, the various shafts and finally the underground facilities). A focus is then made on the 'nuclear ramp' dedicated to the transfer of the waste containers lodged inside the casks (those latter weighing some 100 to 130 metric tons each) and its main geometrical characteristics (slope, length, diameter, structure). Then a discussion follows to detail the rationale having led to the choice of a somehow 'sophisticated' funicular transfer system, at the expense of what could be called 'more conventional' and more straightforward transport solutions (like a transfer system by trucks or by rack and pinion trains). A systematic comparison is finally made between the various solutions, explaining and rating the pros and cons of each concept, via a multi-criteria analysis. This technical choice (and the related justification approach) was submitted to an Experts review at the end of the Preliminary Design phase of Cigeo and comforted by their strong support before the design freeze of this solution was integrated in Cigeo. The following part of the paper is then focused on a description of the main technical components and the mechanical performance of the cable-car system, its current state of design and the set of qualification tests (of key safety components) to be implemented before its effective construction. This article concludes with a timeline of the qualification tests and developments to come. The positive outcomes of this future technical test campaign will also pave the way for the future instruction of the Cigeo licensing file, scheduled as of mid-2018. (authors)

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DOE OSTI2015

Solutions for management of defective fuel assemblies - 15318

Defective fuel management is a major challenge for nuclear operators when the time comes for finding a long-term solution for their used (or spent) fuel assemblies. This question of defective fuel management can be raised in several cases: - For optimizing the remaining operating costs of shutdown reactors, the spent fuel pools must be emptied as soon as possible; a solution needs therefore to be found for evacuating all the spent fuel assemblies, defective fuel assemblies and in particular defective fuel rods included; - Some utilities are today facing a saturation of their reactor spent fuel pool; evacuating the defective fuel assemblies and extracted defective fuel rods allows optimizing the space in the pool. Whereas the solutions for managing intact used fuel assemblies are well-known, a specific technical solution must be implemented for managing the defective spent fuel assemblies as well. The purpose of this paper is to describe the existing solutions for the management of all types of defective fuel assemblies. Interim Dry Storage Solution Interim dry storage of defective spent fuel assemblies is being implemented for decades in the USA. Specific operations are performed for preparing the defective fuel assemblies for interim dry storage and they are then stored with intact used fuel assemblies in the dry storage cask. The advantage of such a solution is its short-term cost effectiveness. The paper presents the existing dry interim storage solutions for defective used fuel rods, the way they are implemented, the associated operational experience. Long-term Dry Storage Solution A long-term dry storage solution for defective fuel assemblies is being developed, compliant with direct disposal requirements. The solution is based on the encapsulation of each defective fuel rod into a fuel rod capsule and an associated specific dry storage licensing approach: this is the best available technology regarding long term storage safety requirements. The paper presents this unique capsule technology, the transport and storage casks that can be used for transportation of defective fuel rods to the storage facility, and the associated transport licensing approach currently used in Europe. Treatment Solution. The most definitive way to get rid of defective fuel assemblies is their treatment (or reprocessing) in a treatment plant. There is then no need to assess the behaviour of these specific used fuel assemblies as for an interim storage facility or a final repository. The advantage of such a solution is its sustainable approach. The paper presents the existing transport and treatment solutions, the way they are implemented, the associated operational experience. (authors)

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DOE OSTI2015

Development of the computer code for multi-calculations of source term of spent fuel - 15351

Many PWR spent fuels have been transferred between the neighboring nuclear power plants at Kori, Hanbit, Hanwul site, and from nuclear power plant to the KAERI(Korea Atomic Energy Research Institute) for the research purpose in Korea. In case of the PHWR type spent fuels, these have been transferred from the spent fuel pool to an on-site dry storage facility. Total radioactivity of spent fuels has to be calculated before transport in advance by the regulation requirement. In case of PWR, KN-12 transport cask can transfer 12 spent fuel assemblies at one time and in case of PHWR, Hi-STAR 63 transport cask can transfer 120 spent fuel assemblies at one time. Therefore, total radioactivity of all transported spent fuel assemblies that are transported generally are calculated by using the ORIGEN-ARP to meet this regulatory requirement. Basically, ORIGEN-ARP calculates the one fuel assembly at one time. SCALE code including this code provides the batch mode to perform the multi-calculations of the source term. In this study, in order to use the batch mode of ORIGEN-ARP, ARP-auto program was developed to generate input files and batch file of ORIGEN-ARP automatically and to make the summary file extracting the result from output files of ORIGEN-ARP. This code enhances the efficiency at the field. The working time to complete the calculation of source terms of 120 spent fuel assemblies take only 20 minutes. In order to confirm the credibility of this code, total radioactivity of 12 spent fuel assemblies of 17x17 Westinghouse type were calculated and compared to the previous result. The initial enrichments of these spent fuels are 3.46 ∼ 3.50 w/o and burnups are 29 ∼ 43 GWD/MTU. As the results, the relative errors are about 0.9 ∼ 3.0%. (authors)

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

Higher harmonic ultrasonic guided waves for structural integrity assessment of dry storage canisters - 15381

Dry storage of used fuel from nuclear power plants has gone from an interim storage solution to a longer-term solution, which necessitates monitoring the integrity of the stainless steel canister that confines the used fuel. Since the storage facilities are located around the country in a variety of different environments there is a wide range of degradation mechanisms that could occur over time in the cylindrical canister. One potential degradation mode that has created significant concern is stress corrosion cracking. We propose using higher harmonic ultrasonic guided waves as a tool to nondestructively inspect used fuel canisters. Higher harmonics are generated when a wave propagates through a nonlinear elastic media. These higher harmonics have been shown to be quite sensitive to microstructural features such as dislocation substructures, persistent slip bands, precipitates, etc. The use of ultrasonic guided waves to inspect the canister is attractive due to the size of the canister and the ability of guided waves to inspect a reasonably large domain relative to traditional bulk wave ultrasonics. We will highlight our recent progress on theory, numerical simulations, and laboratory experiments to develop higher harmonic guided waves as a tool to detect microstructural changes that precede macro-scale damage. Analyzed structures include plates and hollow cylinders, and we focus on localized damage initiation, as is the case for stress corrosion cracking. Recent experiments focus on using magnetostrictive transducers in activating the fundamental shear horizontal mode (SH0) and the third harmonic that is generated due to the material nonlinearity. The magnetostrictive transducer comprises a thin iron-cobalt foil, a meandering electric coil, and a permanent magnet. These transducers are inexpensive, easy to use, and are very efficient at generating shear waves. In applications where the wavelength is larger than the plate thickness, the waves are guided by the traction free boundary conditions and propagate in the plane of the plate. For this canister application, the shear horizontal waves energize the entire thickness of the canister and propagate along the shell in the direction dictated by the meander coil. Experiments indicate that magnetostrictive transducers can generate waves that propagate in plate thicknesses of 1 mm to at least 16 mm. The SH0 mode is nondispersive and generates a third harmonic of the same mode that is shown to be sensitive to localized plastic strain and unseen fatigue damage. It has a strong potential to detect the very early stages of stress corrosion cracking. Due to the geometric configuration of dry storage casks, the transducers will need to be delivered to the inspection points by a robotic system that navigates through the cask ventilation system. (authors)

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

Laying the groundwork for a large-scale spent nuclear fuel transportation system - 15408

The US Department of Energy Office of Nuclear Energy established the Nuclear Fuels Storage and Transportation Planning Project to lay the groundwork for implementing an interim storage facility, including associated transportation activities. Efforts include the development of a system for the large-scale transport of spent nuclear fuel that will be necessary in an integrated waste management system. Progress is being made on long lead time, destination-independent aspects of the transportation infrastructure. The large-scale transportation system for spent nuclear fuel is divided into three elements: institutional, operational, and hardware. The institutional element refers to the various forms of stakeholder interaction that must occur for this type of transportation system to be successful. It includes activities like development of a national transportation plan, work on policy development for Section 180(c) of the Nuclear Waste Policy Act, and identification of a preliminary suite of national transportation routes that reflect the interests of a broad cross section of stakeholders while meeting regulatory requirements. The operational element refers to the activities that must be undertaken to run a large-scale transportation system. This element is currently focused on development of a new SNF transportation routing analysis tool, study of the infrastructure near SNF storage sites that may be de-inventoried first, and development of tools for modeling transportation activities. The hardware element refers to the casks, railcars, and other items necessary to operate the system. This element currently focuses on development of railcars compliant with Association of American Railroads Standard S-2043, as well as studies related to the use of rail casks and their ancillary equipment. The Nuclear Fuels Storage and Transportation Planning Project is making significant progress in all three of these areas along the path forward to a fully operational transportation system. (authors)

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

Achievements and process improvements in processing transuranic waste at the TRU Waste Processing Center - 15488

Wastren Advantage, Inc., (WAI) is the DOE Prime contractor for the TRU Waste Processing Center (TWPC) located in Oak Ridge, TN. This paper highlights the TWPCs specific achievements and process improvements associated with both Contact (CH) and Remote Handled (RH), Transuranic (TRU) waste receipt processing and repackaging for disposal. The five key improvement areas and achievements include: - The TWPC installed, tested and operated successfully the Drum Venting System (DVS) to vent over-packed CH-TRU waste drums in several configurations including 55-gallon drums over-packed in 79 or 85-gallon drums. - A new Remote Drum Opener (RDO) system was designed, tested and placed in service to vent drums discovered in overpack boxes in the Box Breakdown Area and to vent overpack configurations that could not be handled in the standard DVS (e.g., 110 gallon overpacks and multiple drum configurations). - The Cask Processing Enclosure (CPE) approach to repackaging RH-TRU waste contained in concrete casks was developed and implemented - Hot Cell outage to incorporate planned equipment modifications and improvements to maintain processing rates. - To improve certification and segregation of CH-TRU from drums that meet Low Level Waste criteria for disposal at Nevada the TWPC added additional equipment/facilities for Non-Destructive Assay (NDA) and Real Time Radiography (RTR). (authors)

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

Comparisons of Peak Cladding Temperature and Effective Thermal Conductivity: CFD Simulation on PWR Spent Fuel Assemblies - 16278

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 of spent fuel assemblies. There are four methods 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, Wooton-Epstein correlation, and computational fluid dynamics (CFD) simulation. The CFD simulation based on the FLUENT code is used to calculate the peak cladding temperature and effective thermal conductivity of a spent fuel assemblies used in Republic of Korea NPPs: 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 PWR spent fuel assemblies. The CFD simulation results show that the effective thermal conductivity of the 16x16 ACE7 was more conservative than those for the other assemblies. Therefore the effective thermal conductivity using 16x16 ACE7 is used for thermal evaluation of the KORAD's developing cask which can load 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 spent fuel assembly. (authors)

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

Distribution of Documents through the Robatel Technologies Client Portal - 16328

Robatel Technologies introduced the Client Portal section of the www.robateltech.com web site in 2015. This feature allows our customers to download quality documents at their convenience. Initially, the Client Portal will be used to distribute procedures for the RT-100 Type B transportation cask. After a user registers with the web site, Robatel verifies the user and assigns download privileges. The user next receives a confirmation e-mail and instructions on how to access the files. This system allows Robatel to efficiently distribute updated procedures, files, and drawings to all cask users immediately when new files are uploaded to the site. Users are automatically notified of the update. The web site also records and documents when files are downloaded and by whom. This feature ensures compliance with our quality management system by documenting when files are distributed. (authors)

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

Perspective of Dry Interim Storage of spent nuclear fuel (SNF) in Germany - 16504

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection (BfS). In 1995 a license to store spent nuclear fuel (SNF) and vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2035. During the last two years in Germany preparatory consultations to restart the process towards finding a repository took place, organised by a commission. This commission consists out of members of the scientific community, religious parties, social communities and environmental organizations as well as members of federal states governments and members of the German parliament (Bundestag). Recently the commission decided to extend its working period until summer 2016. Taking this background into account the BfS plans to intensify its research into extended interim storage with respect to potential material degradation leading to interference with safety functions (e.g. neutron shielding material). This includes topics like ageing of gasket materials which are already studied in ongoing experiments (e.g. at the Federal Institute for Materials Research and Testing) as well as studies of long term relaxation processes. Because of the need for transportation towards a repository another important field is the behaviour of the inventory under mechanical loads. For SNF, first calculations showed that because of the lower decay heat load after 60 years due to hydride reorientation the cladding might become much less ductile. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A burnup credit approach for margin estimation of loaded boiling water reactor canisters in UNF-ST-DARDS

This paper discusses the modeling approach applied in the Used Nuclear Fuel Storage, Transportation, and Disposal Analysis Resource and Data System (UNF-ST-DARDS) to take credit for the reduced reactivity associated with the burnup of boiling water reactor (BWR) fuel for canister-specific as-loaded criticality analysis. Burnup credit is routinely used for the analysis of pressurized water reactor (PWR) spent nuclear fuel (SNF) storage (wet), transportation, and disposal. The fresh fuel assumption has historically been applied to BWR SNF criticality analyses for dry storage and transportation, and the peak reactivity method of burnup credit has been applied to pool storage. The fresh fuel assumption ignores the credit for gadolinium burnable absorbers in BWR fuel, as well as the burnup of the fuel, and is therefore conservative. This work employs the framework traditionally used for PWR burnup credit analyses and applies available data and recently published regulatory guidance to justify a new approach to model the as-loaded burnup and enrichment of SNF assemblies in currently loaded dry casks. The following features of this criticality analysis approach are presented in this paper: (1- selection of axial burnup profiles for BWR fuel from publicly available sources using methods derived from recently published research, (2- justification of the selected axial burnup profiles as relevant to all types of BWR fuel, (3- justification for modeling the fuel assemblies with a uniform axial and radial enrichment, and (4- justification for modeling the axial void profile and control blade insertion during depletion. This analysis approach was used to evaluate 153 already-loaded BWR casks at 8 sites as a function of time in UNF-ST-DARDS. Finally, the results of the as-loaded criticality safety margin assessments are presented. The results show the k{sub eff} calculated for storage and transportation cases range from 0.71 to 0.83, and the eigenvalues calculated for the disposal cases range from 0.85 to 1.03, though the highest cases are dominated by very limiting hypothetical damaged fuel assumptions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Sensitivity and uncertainty analysis for a combined depletion and criticality calculation chain of a BWR FA

This paper examines the prerequisites and boundary conditions for extending burnup credit from pressurized water reactors to boiling water reactors with validated depletion codes by performing uncertainty and sensitivity analyses on a modern boiling water reactor fuel assembly design. We developed a computational chain consisting of an axial boiling water reactor axial irradiation model, followed by a series of depletion calculations and a three-dimensional criticality calculation. The irradiation model puts into relation the profiles of power, burnup and moderator density of the fuel assembly at 24 axial locations. In a simplified fashion it also includes the insertion of control blades. The depletion calculations of the fuel assembly were performed in axial slices with the 2D-depletion code HELIOS 1.12. The criticality calculations were done in 3D-geometry with KENO-Va from the SCALE 6.2.1 package, in which several fuel assemblies are placed into a generic transport cask. For the uncertainty and sensitivity analyses a Monte Carlo sampling based approach is used applying the GRS tool SUnCISTT. We performed several generic investigations testing the capabilities and the limitations of the whole computational chain. The impact of variations of the irradiation model on the resulting reactivity of the fuel assemblies inside a transport cask or storage pool was investigated using an uncertainty and sensitivity analysis on the whole computational chain. Geometrical uncertainties were also included in the evaluation and a parameter study on the use of control blades during irradiation was performed. The goal was to develop the procedures and tools to be able to obtain quantitative uncertainty data which can be used in the future to derive conservative irradiation profiles without unnecessary excess reactivity margins. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Updating the Dry Transfer System - 18110

The Dry Transfer System (DTS) was examined for its ability to be redesigned in order to meet needs associated with repackaging used/spent nuclear fuel (UNF/SNF) from existing storage cask systems into alternative systems, while also incorporating design advances based on some lessons learned from Orano's dry UNF handling facility (T0) located at the La Hague recycling plant. In this redesign, the DTS is envisioned to provide some flexibility through the ability to perform repackaging activities from multiple differently designed cask systems, including those with canisters that will require cutting and welding activities, and potentially with waste packages designed for disposal. The redesigned DTS is also intended to provide some operational flexibility by providing lag storage and include a means for examining UNF/SNF during repackaging, using some features implemented at the T0 facility at La Hague. This paper provides a summary of: the original DTS design, which was evaluated by the NRC and portions of which were cold-tested at Idaho National Lab; the lag storage system designed for the DTS, which is able to flexibly handle different fuel assembly designs while providing structural support for design basis events (e.g., seismic); the incorporation of lessons learned from the La Hague T0 facility into the DTS design to facilitate dry handling of UNF/SNF; and identification of other challenges towards designing a complete system. A revised DTS is envisioned to provide an important tool for the future management of UNF/SNF in preparation for transportation, re-storage, re-transportation, and/or disposal and also provides a means for repackaging UNF/SNF in case recovery from a cask/canister system is necessary (e.g., due to extended storage or a transportation event). (authors)

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