Engineering topics

Eidelpes, Elmar

Publications and source records attributed to Eidelpes, Elmar.

3 records

DOE OSTI2022

A risk-informed assessment tool to support review of license amendment requests for spent nuclear fuel dry cask storage systems

We report that while risk-informed safety evaluations are common methodologies used by the U.S. Nuclear Regulatory Commission (NRC), their application in the field of commercial spent nuclear fuel (SNF) dry storage has received limited attention. For instance, license amendment requests (LARs) for SNF dry cask storage systems (DCSSs) are evaluated according to U.S. NRC standard review plans (SRPs). However, risk-informed LAR review strategies could improve the understanding of critical system modifications, and lead to more predictable, efficient, and consistent LAR review processes. A methodology to develop a risk-informed LAR review tool is presented. Although focused on DCSSs, this methodology can be used to develop similar tools for a broad spectrum of safety-relevant systems. The end product (i.e., the tool) includes a tree diagram to help visualize the review risk significances assigned to a predefined set of system modifications and supporting rationale documentation. Instructions support correct tool application and provide the user with a path to incorporate currently unevaluated modifications in the tool structure. The demonstration of the methodology leads to the conclusion that risk-informed reviews of DCSS LARs are possible, although the tool precision could benefit from additional risk information. Further, the results indicate that many typical LARs involve low-risk or medium-risk modifications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

A High-Assay Low-Enriched Uranium Fuel Transportation Concept

The uranium 235U enrichment commonly used in fuel production for U.S. light water nuclear reactors typically does not exceed 5 wt%. In contrast, many of the currently investigated advanced reactor concepts demand fuel with higher enrichments. This includes high-assay low-enriched uranium (HALEU), characterized by a 235U enrichment of 5 to 20 wt%. The necessity of HALEU transportation in the fuel production cycle leads to new challenges caused by various technical and regulatory hurdles. Current U.S. Nuclear Regulatory Commission–approved transportation package designs for UF6 with enrichments above 5 wt% provide relatively small payloads [=116 kg (250 lb)]. Furthermore, in accordance with 10 CFR 71.55, package design activities for fissile material enriched above 5 wt% need to consider water infiltration in the containment as part of the criticality safety evaluations. This study presents a transportation package concept for HALEU advanced nuclear reactor fuel with a significantly higher payload of up to 376 kg (830 lb) of fissile material per package and up to 1881 kg (4149 lb) of HALEU per legal weight truck. The anticipated chemical form of the transported material is UO2 downblended from available highly enriched uranium. The concept utilizes a combination of existing transportation packaging, 18 inner canisters, and a novel basket design that includes a borated aluminum flux trap. Criticality and shielding evaluations; fundamental structural, confinement, and thermal assessments; and studies on package operations are presented. The results of this study build significant confidence in the technical feasibility of a high-capacity HALEU transportation package concept while demonstrating the concept’s potential to meet U.S. regulatory requirements.

Eidelpes, Elmar↗

DOE OSTI2020

Assessment of Critical Experiment Benchmark Applicability to a Large-Capacity HALEU Transportation Package Concept

This work presents an assessment of the applicability of existing benchmark critical experiments to the criticality safety code validation for a large-capacity high-assay low-enriched uranium (HALEU) transportation package concept. Numerous next-generation nuclear reactor designs require HALEU fuel, which is characterized by an enrichment between 5 and 20 wt% 235 U. The U.S. Department of Energy (DOE) has proposed to recover and downblend highly enriched uranium from DOE-owned used nuclear fuel to accelerate the demonstration of commercially viable microreactor technologies. One element of the infrastructure needed to demonstrate HALEU-fueled reactors is the ability to safely transport enriched product to be used for fuel fabrication. There is uncertainty as to whether existing critical benchmark experiment data are sufficient to support criticality safety code validation for HALEU transportation applications. The anticipated chemical form of the HALEU in the proposed transportation concept is UO 2 with 20 wt% 235 U/U. The concept uses a combination of an existing transportation packaging design and a novel basket design, including borated aluminum flux traps. The basket provides space for 18 reusable, stainless steel canisters that contain the HALEU. In 10 CFR 71, normal conditions of transport (NCTs) and hypothetical accident conditions (HACs) are defined for fissile material transportation packages. NCT and HAC KENO-VI models of the transportation package were developed using the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.2.3 computer code package, and optimum moderation conditions were determined using the SCALE SAMPLER sequence. The SCALE Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) sequences were then used to compare the neutronic characteristics of 1584 International Criticality Safety Benchmark Evaluation Project benchmark critical experiments with the NCT and HAC HALEU transportation models. The TSUNAMI integral correlation coefficient c k was the criterion used to rank neutronic similarity. Thirty-four experiments were identified as similar (c k ≥ 0.9) to the NCT model, and 55 experiments were identified as similar to the HAC model. Hundreds of experiments were also identified as at least marginally similar (c k ≥ 0.8) to both models. The results indicate that additional critical experiments are unlikely to be needed to support HALEU transportation criticality safety analyses for package concepts similar to the concept package analyzed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗