A Unified Spent Nuclear Fuel Database and Analysis System

J.M. Scaglione, R.A. LeFebvre, K. Banerjee, G. Radulescu, K.R. Robb
Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA

Abstract. A fully integrated spent nuclear fuel (SNF) management system involves managing SNF from the time it is discharged from a reactor to the time it is disposed of in a geologic repository. To support the US Department of Energy Office of Nuclear Energy (DOE-NE) Nuclear Fuels Storage and Transportation Planning Project (NFST) activities to lay the groundwork for implementing interim storage, including associated transportation (per the Administration’s Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste), an integrated data and analysis tool has been developed. The tool is called the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). UNF-ST&DARDS provides a controlled source of technical data integrated with key analysis capabilities to characterize the inputs to the overall US waste management system from reactor power production through ultimate disposition. This system is a new capability/resource that enables automated assembly-specific and cask-specific evaluations for assessing issues and uncertainties related to the extended storage and transportability of loaded canisters, supporting safety confidence and research and development prioritization, and providing a foundational data and analysis capability resource for the future. This paper provides an overview of the UNF-ST&DARDS architecture and automated analysis capabilities.

Introduction

Commercial spent nuclear fuel (SNF) in the United States is expected to remain in storage for considerably longer than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) result in increased uncertainties related to aging management of the structures, systems, and components important to safety during normal, off-normal, and accident conditions of ES and subsequent transportation. An integrated SNF management system involves managing the SNF from the time it is discharged from the reactor to the time it is disposed of in a geologic repository. As depicted in the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste [1], a set of possible pathways to developing the waste management system is presented in Fig. 1. Because the SNF may be stored, transported, and handled multiple times before resting in a geologic repository, uncertainties associated with the physical characteristics of the fuel assemblies and canister systems will propagate and compound with time. To implement effective aging management strategies, these uncertainties need to be understood and addressed before shipping, handling, and ultimate disposal. Hence, considerations must be made to ensure that the fuel assembly data and other information needed to support SNF characterization activities are available and maintained.

To streamline analyses for the waste management system, a comprehensive, integrated data and analysis tool has been assembled—the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). UNF-ST&DARDS provides a comprehensive domestic SNF system database, referred to as the Unified Database, integrated with key analysis capabilities to support numerous US Department of Energy (DOE) waste management and fuel cycle-related objectives, as well as the foundation for collecting information on SNF from reactor power production through ultimate disposition. UNF-ST&DARDS provides a controlled source of technical data for various waste management system analysis/evaluation tools, as well as fuel cycle system analyses and safeguards and security studies; access to key technical data and analysis capabilities to characterize the SNF inventory; and enhanced capability to assess safety, risks, and uncertainties throughout the waste management system. Initial development of UNF-ST&DARDS has been a collaborative effort among multiple national laboratories and several utilities. Automation sequences have been developed and initiated to generate data that can be used to assess the state of specific populations of SNF over time, including characterization of the SNF inventory [2][3], and criticality safety, dose, containment, and thermal analyses.

FIG. 1. Waste management system pathways

A unique, unprecedented capability within UNF-ST&DARDS is the performance of actual cask-specific evaluations. This capability enables quantification of realistic safety margins and conditions of actual as-loaded SNF in existing dry cask storage systems and potentially in future cask systems. This capability enables informed decision making with the most accurate information available relative to planning for transportation operational safety (e.g., worker dose), storage facility design, handling operations such as any subsequent packaging (i.e., for storage, transport, and/or disposal), and addressing concerns related to fuel failure of high-burnup fuel during and following transportation. Overly conservative analyses based on bounding generic assumptions could lead to prematurely invoking compensatory measures within the SNF management system, such as repackaging already packaged fuel, canning a significant fraction of SNF before storage, or requiring additional criticality control before transport and disposal (e.g., inserting control rod assemblies). For example, licensing-basis thermal analysis methods typically overpredict the time-dependent temperature profiles by an amount that is not readily quantifiable in general and that would vary widely based on the individual characteristics of a given cask. This would hinder efforts to accurately predict the potential for confinement breaches from low-temperature degradation phenomena (e.g., deliquescence) and would tend to overpredict the number of fuel rods that may fail during extended storage and/or transportation because of thermally driven mechanical property changes (e.g., hydride reorientation) [4].

System Architecture

The UNF-ST&DARDS is illustrated in Fig. 2 and consists of six primary elements: (1) a user interface (graphical as well command line), (2) the Unified Database, (3) a Unified Database software development kit (SDK), (4) a template repository that contains constructs for specific nuclear safety analysis tools that have been integrated, (5) an application agnostic template engine, and (6) a process manager that handles interactions between the different elements. The nuclear safety analysis tools and the tool interfaces are externally developed software that function independently. The software infrastructure has been designed with the capability to integrate virtually any set of tools, tool interfaces, and analysis packages, but it would require development of appropriate templates for inclusion in the template repository and any specific input attributes added to the Unified Database as necessary. Currently, Standard Computation Analyses for Licensing Evaluations (SCALE) [5] version 6.2 beta and COolant Boiling in Rod Arrays–Spent Fuel Storage (COBRA-SFS) [6] have been fully integrated. A tool interface that has been integrated is a new feature in SCALE 6.2 called Fulcrum that enables dynamic SCALE input editing, model previews, and results visualization.

FIG. 2. Schematic of UNF-ST&DARDS

Performing the different types of analyses required to understand the changing nuclear and mechanical characteristics of the SNF, and how these characteristics affect the different storage, transportation, and disposal options, can require many tools and types of data. The following datasets are stored in the Unified Database:

Facility Attributes—include utility, site, reactor, pool, and independent spent fuel storage installation (ISFSI) data

Federal Government Radioactive Waste Attributes—include government-owned high-level waste and government-owned SNF data

Future Facility Attributes—include interim storage facility (ISF), repackaging facility, and repository data

Transportation Infrastructure Attributes—include rail, heavy haul, legal-weight truck, and barge assets data in addition to transfer times between these transportation modes

Assembly-Specific Attributes—include both actual and projected assembly discharge information and characteristics (e.g., isotopic concentrations, source terms)

Cask/Canister Attributes—include overpack, capacity, physical properties, unit processing times, unit costs, and characteristics (e.g., dose, component temperatures, keff)

Economic Attributes—include reactor sites unit cost, transportation infrastructures unit cost, interim storage facility, repackaging facility, and repository surface costs

These data have been collected from the DOE Office of Civilian Radioactive Waste Management and General Council nuclear data forms (RW-859/GC-859) [7] database and other sources. These sources include DOE studies, the US Nuclear Regulatory Commission Agencywide Documents Access and Management System, open literature, vendor data, and utility data. One essential requirement of the Unified Database is that all of the data within the database be traceable. All data references are uploaded to the Centralized Used Fuel Resource for Information Exchange (CURIE) website at curie.ornl.gov, allowing permanent storage and simple retrieval for all imported references.

The user interface with UNF-ST&DARDS is performed via a graphical user interface (GUI) or command line interface. The command line interface provides access to launch all UNF-ST&DARDS analysis capabilities. In the GUI, an analysis page allows the user to select an analysis type (e.g., criticality) and either (1) reactor type batch (e.g., all pressurized water reactors [PWRs]), or (2) specific analysis. In addition to conducting analysis, ISFSI specific results can be explored via an integrated geographic information system (GIS). Additional results that are accessible include assembly-specific decay heat and isotopic inventory data, and cask-specific criticality, thermal, and dose data. Plotted data can be exported to Microsoft Excel. Commonly requested data sets also are available in the GUI in tabular format (e.g., wet and dry storage inventory and reactor information). The GUI also provides read-only access to all data in the Unified Database via an integrated structured query language (SQL) viewer for advanced users.

The process manager manages the workflow across the Unified Database, integrated nuclear safety analysis tool templates and applications, and results retrieval, storage, and interrogation. This software uses the Java Persistence application programming interface entity, controller, and annotation classes to abstract the communication with the Unified Database. Model templates are stored in the template repository and are implemented in conjunction with structured parameter sets in JavaScript Object Notation (JSON) [8] and the template engine to abstract the management of nuclear safety analysis tool and system processes.

The template engine (or template processor) is used to combine site-specific input parameters from the unified database with the model templates developed for the nuclear safety analysis tool calculations to produce complete input files for those calculations. The template engine is a string substitution program designed to take advantage of repeated structures in text files. The template engine takes the input parameter data structures represented by a JSON data structure and the root template file. With these two components, the template engine conducts attribute replacement and sub-template imports. The model templates contain three basic components: (1) input data blocks that do not vary as a function of fuel assembly characteristics (e.g., description of canister/cask dimensions and construction materials for criticality, dose rate, or thermal calculations), (2) input parameters that vary as a function of assembly characteristics (e.g., fuel pin dimensions in an assembly model for depletion calculations or nuclide concentrations in a cask model for criticality calculations), and (3) sub-templates to be imported (e.g., templates describing fuel pin arrays for depletion or criticality calculations). Model template development, update, and review are conducted in a controlled manner.

One of the key features of the Unified Database is the capability for integration with multiple computational tools and models. Integration with the Unified Database is streamlined with a Java programming language SDK. As other internally or externally developed tools, such as system analysis tools, are identified that can benefit from data in the Unified Database, a bottom-up approach is used to develop the interface requirements. This ensures that the parameter sets that the tool needs are available in the Unified Database or are added to it with appropriate references. Multiple interface layers are maintained for the specific tools, which allow multiple projects to move forward independently while maintaining the data relationships between them. The primary interface with the unified database is with the SDK.

Automation of Computational Analyses

Because of the complexity of the complete computational analysis process to characterize the SNF as it traverses though the SNF management system, considerable time and effort can be expended simply building input files and troubleshooting errors in the process. UNF-ST&DARDS simplifies the computational analysis input generation process, allowing analyst(s) to focus more time on analyzing results, trends in results, and sensitivities. A key database design element facilitating computational analysis automation is how the data relations are defined within the Unified Database. The UNF-ST&DARDS automation features have been built around the SCALE and COBRA-SFS codes, but the automated input building process is independent of code choice. It could be readily adapted for other software packages so that information from the Unified Database could be used directly once input requirements for the other software packages are defined. The data relationships that have been defined allow the inputs to the respective codes to be built autonomously when basic information about the SNF and cask system is provided and essentially eliminate user interaction required to build the large number of computer code inputs needed for characterizing the SNF at the assembly-specific and cask-specific level for each site. Current analysis capabilities include neutronics, dose rate, containment, and thermal analyses. Neutronics analyses involve simulating neutron transport and include depletion, decay, and criticality calculations. Thermal analyses are used to calculate component temperatures (e.g., cladding temperature) and are fundamental to understanding cask surface degradation mechanisms. Dose analyses provide radiation dose rate maps around an as-loaded transportation cask configuration, which could be useful for operational planning, emergency response, and interactions with state and regional groups. Containment analyses are used to evaluate potential radionuclide release rates against regulatory limits.

Because SCALE has been linked with UNF-ST&DARDS, automated execution of the following modules is available: Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion (TRITON) transport and depletion module, the Oak Ridge Isotope Generation (ORIGEN) assembly isotope generation code, the Criticality Safety Analysis Sequence with KENO-VI (CSAS6), and Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) shielding sequence. A new SCALE 6.2 feature also provides the ORIGen AsseMbly Isotopics (ORIGAMI) interface for performing automated ORIGEN calculations. Note that unlike licensing evaluations, which are designed to cover a broad range of SNF characteristics, assembly- and cask-specific asloaded analyses are performed without attempting to establish bounding parameters and conditions but are used to establish best-estimate approximations. This means that average parameter values may be used in place of bounding parameter values for a desired effect. For example, the criticality calculations are based on the burnup credit isotopes recommended by NUREG/CR-7108 and -7109 [9][10], which are different from most licensing evaluations that typically used a fresh fuel assumption for meeting dry cask storage requirements. This allows quantification of the inherent uncredited margin that may be useful at some future time. Conservatisms from the final safety analysis reports are retained for the different analyses where insufficient information is available to justify a change, so some level of conservatism is retained in the generated results for a specific application.

Upon execution, UNF-ST&DARDS creates JSON files with values provided by the Unified Database that are used with the template engine and previously generated templates to create an input file that is then run by the appropriate code. This process is used to perform analyses (depletion, decay, criticality, shielding, containment, and thermal) for all SNF assemblies and canisters/casks at a given site, or alternatively for all SNF assemblies and canisters/casks for which data are available. Upon execution, SCALE modules TRITON and ORIGEN perform depletion and decay calculations in parallel for all SNF assemblies presently stored at the site for the user-prescribed date(s). Subsequently, continuous-energy Monte Carlo criticality (CSAS6), shielding (MAVRIC), and thermal (COBRA-SFS) input files are generated with the appropriate time-decayed SNF isotopic compositions, radiation source term, and decay-heat source term information, respectively. These files are executed automatically and in parallel for all the loaded canisters for each date specified. Full output files are archived, but pertinent results such as peak and minimum clad temperatures, component temperatures, cask surface temperatures, total decay heat, keff, dose rates at select locations, and allowable leakage rates are extracted and stored in the Unified Database for future use.

Data and Results Visualization

Cask-specific as-loaded analyses for storage and transportation configurations have been performed with UNF-ST&DARDS to assess the realistic state of SNF for various sites for which data has been collected. Once the data is available to perform analyses, the assembly-specific characteristics are calculated and stored within the Unified Database. Cask-specific analyses are subsequently performed with pertinent results also retrieved and stored in the Unified Database. Currently, the Unified Database contains assembly-specific charactersitics for approximately 150,000 individual fuel assemblies and 300 different as-loaded canisters/casks. These values will continue to grow, eventually encompassing all discharged fuel and loaded canisters/casks in the United States.

As discussed earlier, individual site results can be viewed through the GUI. Screen snapshots from the interface using the integrated GIS are shown in Fig. 3. Commercial SNF locations within the United States are identified, and additional drill-down capabilities through the GUI are available, including going from the individual sites to the individual casks, to the as-loaded contents, and to the isotopic compositions of the individual fuel assemblies. An as-loaded transportion cask dose rate map generated through the interface Fulcrum is illustrated in Fig. 4. Estimated as-loaded decay heat load per cask as a function of time is presented in Fig. 5 for selected reactor sites referred to as Sites A, B, C, D, E, F, and G that include both decommissioned PWR sites and decommissioned boiling-water reactor sites. Criticality analysis results are shown for all casks at a single site in Fig. 6, and Fig. 7 shows a horizontal cross-sectional view heat map illustrating the temperature distribution and other relevant information of an as-loaded cask.

FIG. 3. Example of GUI visualization capabilities for site characteristics

FIG. 4. Radial and axial radiation dose rate maps for an as-loaded transportation cask (dose rate in mrem/h)

FIG. 5. Total cask decay heat as a function of calendar year for selected reactor sites

FIG. 6. As-loaded keff values for casks at single site accounting for burnup credit

FIG. 7. Heat map for selected cask with various as-loaded contents specified, including calculated peak clad temperature

Summary

One of the most important aspects for analysis of large, complex systems is a reliable source of information with a capability to update system parameters dynamically based on new or changing information. UNF-ST&DARDS provides this capability through a controlled, comprehensive relational Unified Database coupled with nuclear safety analysis tools, and it represents the quintessential element of a fully integrated SNF management system. Integrating the Unified Database with the analysis pipeline streamlines analyses for the SNF management system by automating input file development. This capability allows UNF-ST&DARDS to perform assembly-specific and canister-specific analyses for all SNF at the different reactor sites, enabling informed decision making with the most accurate information available. The Unified Database provides a consistent set of reliable data that can be controlled and updated to support subsequent analyses and a variety of tools, preserve the results and parameters that will be required for downstream assessments at different operational transitions within the SNF management system, and address and understand the significance of any emerging issues.

References

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