Spent Nuclear Fuel Assemblies Research Articles

Measurement of homogeneity of U and Pu isotopes in PWR spent fuel powder mixture

We present a feasibility study of the homogenization of pressurized water reactor spent nuclear fuel (SNF) powder through a mechanical mixing process. Because burn-up of the SNF depends on the position in the SNF assembly, concentrations of uranium, plutonium, and fission products are distributed differently according to the burn-up profile. The heterogeneity of the material elements affects the selection of a representative sample for quantitative analysis. Homogenization process improvement to reduce the sampling error is thus required to precisely determine the amount of uranium, plutonium, and fission products in the SNF. In this study, fine powders (<70 μm) extracted from one SNF rod were mixed, and the degree of homogenization was determined as a function of the mixing time indicating the relative standard deviations of the 134Cs/137Cs, Pu, and U isotope ratio measurements.

Uncertainty Quantification for New Approaches to Spent Fuel Assay

Estimating plutonium (Pu) mass in spent nuclear fuel assemblies (SFAs) helps inspectors ensure that no Pu is diverted. Therefore, nondestructive assay (NDA) methods are being developed to assay Pu mass in SFAs. Uncertainty quantification is an important task in most assay methods, and particularly for SFA assay. A computer model (MCNPX) is being used to predict isotope masses and the spatial distribution of masses in virtual SFAs for 64 combinations of initial fuel enrichment (IE), fuel utilization [burnup (BU)], and cooling time (CT) values. Additional MCNPX modeling for the same 64 virtual SFAs provided the expected detector responses (DRs) for several NDA techniques such as the passive neutron albedo reactivity method and the 252Cf interrogation with prompt neutrons method.A previous paper describes one uncertainty quantification approach involving Monte Carlo (MC) simulation using individually any of six new NDA options together with IE, BU, and CT. This paper provides an interpretation of the MC approach that is suited for a numerical Bayesian alternative, separately assesses the impact of MCNPX interpolation error, and compares several options to use subsets of IE, BU, CT, and one DR.

Heuristic Methods for Assigning Spent Nuclear Fuel Assemblies to Canisters for Final Disposal

The disposal of spent fuel assemblies (SFAs) by companies currently producing nuclear power in Finland is the responsibility of a company named Posiva Oy. Posiva Oy has decided to use the KBS-3 (Swedish abbreviation for nuclear fuel safety; version 3) concept. In KBS-3, SFAs are placed in metal canisters, which are themselves deposited deep into crystalline rock. The disposal process in Finland will last many decades. To efficiently assign SFAs to canisters, in this paper we study the minimax canister formation problem. In this problem, we assume we are given two sets of data: (a) a schedule specifying the number of disposal canisters per year and (b) the decay heat of each SFA for every disposal year. The goal in the problem is to assign SFAs to canisters so that the largest canister heat load is minimized. The minimax canister formation problem is a variant of a well-known optimization problem: makespan minimization on unrelated parallel machines. We developed heuristic methods for solving the minimax canister formation problem. Using our methods and predicted SFA amounts and properties for Finland, we obtained high-quality solutions in numerous test cases. We also investigated how the uncertainty in SFA burnups affects the canister heat loads.

Very Long-Term Dry Storage Systems for Spent Nuclear Fuel: Effect of Canister Weld Corrosion on System Integrity

ABSTRACTTermination of the Yucca Mountain repository program implies that U.S. spent nuclear fuel (SNF) and high-level radioactive waste may be stored for very long periods, i.e., more than 100 years. Many of the storage systems for commercial SNF are large, cylindrical, welded, stainless-steel canisters containing 24-80 SNF assemblies in an inert gas (helium) environment. Each canister is contained in a massive reinforced concrete structure for shielding, safety, and security reasons. Airflow in the annular space between the canister and the concrete structure removes SNF decay heat by natural convection.Undetected defects in the canister-lid welds and stress corrosion cracking (SCC) in the canister-lid weld area may combine to form pathways for slow escape of the inert fill gas from the canister. SCC may not occur until many decades after SNF is loaded into the canister because SNF decay heat initially would keep the canister surface temperatures above the temperature where liquid water, a prerequisite for SCC, could condense. The period required for inert gas to escape from the canister once pathways have formed and for air subsequently to enter the canister may also be measured in decades, depending on the size and shape of the pathways.In this paper, the authors explore the limits of weld-flaw detectability; when necessary conditions for SCC could occur in the canister-lid weld area; the period for loss of inert fill gas if through-weld cracks occur; and the possible consequences of fill gas loss. The authors also develop a list of research and study needs to address the possibility of SCC resulting in the loss of fill gas.

Effect of Depleted-Uranium Dioxide Particulate Fill on Spent-Nuclear-Fuel Waste Packages

The use of depleted uranium dioxide (DUO2) particulates as fill material for repository waste packages (WPs) containing light-water reactor (LWR) spent nuclear fuel (SNF) was investigated. A repository WP would be loaded with SNF, and small DUO2 particulates (0.5 to 1.0 mm) would be added to fill the void space inside the WP - including the coolant channels inside SNF assemblies. The DUO2 fill slows release of radionuclides from the SNF by (a) creating a local chemically reducing environment that slows degradation of the SNF UO2 and (b) reducing groundwater flow through the WP. The depleted uranium (DU) fill minimizes the potential for long-term criticality in the repository by isotopic dilution of 233U and 235U. The potential for criticality is primarily determined by 235U (a) originally in the SNF and (b) from decay of 239Pu. The use of DU consumes excess DU from the production of enriched uranium. The mechanisms for improvements in repository performance with DUO2 fill are defined, but additional work is required to fully quantify the benefits and costs of such an approach.

Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel—I: Methodology Overview

A conservative methodology is presented that would allow taking credit for burnup in the criticality safety analysis of spent nuclear fuel (SNF) packages. The method is based on the assumption that the isotopic concentration in the SNF and cross sections of each isotope for which credit is taken must be supported by validation experiments. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentration with burnup. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps:1. Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. Chemical assay benchmarks are used for this purpose, in conjunction with a method for assessing the calculational bias and uncertainty for each isotope.2. Validate a computer code system to predict the subcritical multiplication factor keff of an SNF package by use of UO2 and UO2/PuO2 critical experiments. The method uses an upper safety limit on keff (which can be a function of trending parameters) to ensure that the calculated keff when increased for the bias and uncertainty is <0.95.3. Establish conditions for the SNF (depletion analysis) and package (criticality analysis) that bound keff. Bounding axial and horizontal profiles must be established to ensure that the “end effect” and “horizontal effect” are accounted for conservatively.4. Use the validated codes and bounding conditions to generate package-loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment.5. Verify by measurement that SNF assemblies meet the package-loading criteria, and confirm proper assembly selection prior to loading.