Joshua L. Peterson

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

Abstract A novel assessment has been completed to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing-basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed, calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014), and significant uncredited transportation dose rate margins were also observed. The results demonstrate that at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.

Development of Streamlined Nuclear Safety Analysis Tool for Spent Nuclear Fuel Applications

Abstract To understand the changing nuclear and mechanical characteristics of spent nuclear fuel (SNF) or used nuclear fuel (UNF) and the different storage, transportation, and disposal systems at various stages within the waste management system, different types of analyses are required. These analyses require the use of assorted tools and numerous types of data. Using the appropriate modeling and simulation (M&S) parameters and selecting from the diversity of analytic tools to conduct SNF analyses can be a tedious, error-prone, and time-consuming undertaking for analysts and reviewers alike. A new, integrated data and analysis system was designed to simplify and automate performance of accurate, efficient evaluations for characterizing the input to the overall U.S. nuclear waste management system—the UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database has been assembled to provide a standard means by which UNF-ST&DARDS can succinctly store and retrieve M&S parameters for specific SNF analysis. A library of various analysis model templates is used to communicate M&S parameters for the most appropriate M&S application. A process manager facilitates performance of actual as-loaded, assembly-specific, and cask-specific evaluations. Interactive visualization capabilities facilitate data analysis and results interpretation. To date, UNF-ST&DARDS has completed (1) explicit depletion and decay analysis of every fuel assembly (~245 000) discharged from commercial U.S. reactors through June 2013, with 13 cooling time steps (results include isotopic compositions for 142 isotopes, and radiation and thermal source terms); (2) SNF radiation dose rate evaluations at 1 m for all the fuel assemblies discharged through June 2013; and (3) criticality, shielding, thermal, and containment analyses of hundreds of loaded casks. UNF-ST&DARDS also provides various automated report generation capabilities with dynamic figure and table update capabilities based on changes to the Unified Database.

The Multi-Step CADIS method for shutdown dose rate calculations and uncertainty propagation

Abstract Shutdown dose rate (SDDR) analysis requires (a) a neutron transport calculation to estimate neutron flux fields, (b) an activation calculation to compute radionuclide inventories and associated photon sources, and (c) a photon transport calculation to estimate final SDDR. In some applications, accurate full-scale Monte Carlo (MC) SDDR simulations are needed for very large systems with massive amounts of shielding materials. However, these simulations are impractical because calculation of space- and energy-dependent neutron fluxes throughout the structural materials is needed to estimate distribution of radioisotopes causing the SDDR. Biasing the neutron MC calculation using an importance function is not simple because it is difficult to explicitly express the response function, which depends on subsequent computational steps. Typical SDDR calculations do not consider how uncertainties in MC neutron calculation impact SDDR uncertainty, even though MC neutron calculation uncertainties usually dominate SDDR uncertainty. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) hybrid MC/deterministic method was developed to speed SDDR MC neutron transport calculation using a deterministically calculated importance function representing the neutron importance to the final SDDR. Undersampling is usually inevitable in large-problem SDDR simulations because it is very difficult for the MC method to simulate particles in all space and energy elements of the neutron calculation. MS-CADIS can assess the degree of undersampling in SDDR calculations by determining the fraction of the SDDR response in the space and energy elements that did not have any scores in the MC neutron calculation. It can also provide estimates for upper and lower limits of SDDR statistical uncertainties resulting from uncertainties in MC neutron calculation. MS-CADIS was applied to the ITER SDDR benchmark problem that resembles the configuration and geometrical arrangement of an upper port plug in ITER. Without using the hybrid MC/deterministic methods to speed MC neutron calculations, SDDR calculations were significantly undersampled for all tallies, even when MC neutron calculation computational time was 32 CPU-days. However, all SDDR tally results with MC neutron calculations of only 2 CPU-days converged with the standard Forward-Weighted CADIS (FW-CADIS) method and the MS-CADIS method. Compared to the standard FW-CADIS approach, MS-CADIS decreased the undersampling in the calculated SDDR by factors between 0.9% and 0.3% for computational times between 4 and 32 CPU-days, and it increased the computational efficiency of the SDDR neutron MC calculation by factors between 43% and 69%.

Shutdown Dose Rate Analysis Using the Multi-Step CADIS Method

Abstract The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) hybrid Monte Carlo (MC)/deterministic radiation transport method was proposed to speed up the shutdown dose rate (SDDR) neutron MC calculation using an importance function that represents the neutron importance to the final SDDR. In this work, the MS-CADIS method was applied to the ITER SDDR benchmark problem. The MS-CADIS method was also used to calculate the SDDR uncertainty resulting from uncertainties in the MC neutron calculation and to determine the degree of undersampling in SDDR calculations because of the limited ability of the MC method to tally detailed spatial and energy distributions. The analysis that used the ITER benchmark problem compared the efficiency of the MS-CADIS method to the traditional approach of using global MC variance reduction techniques for speeding up SDDR neutron MC calculation. Compared to the standard Forward-Weighted-CADIS (FW-CADIS) method, the MS-CADIS method increased the efficiency of the SDDR neutron MC calculation by 69%. The MS-CADIS method also increased the fraction of nonzero scoring mesh tally elements in the space-energy regions of high importance to the final SDDR.