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Puig, Francesc

Publications and source records attributed to Puig, Francesc.

6 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

MITR & NBSR DDE Irradiations in BR2 – Fluence in LEU Cladding and Structural Materials

The BR2 nuclear reactor is a material testing reactor (MTR) located in Mol, Belgium, and operated by the Belgian Nuclear Research Centre (SCK CEN) since 1963. The reactor is highly versatile as the number and location of fuel elements and control rods can change significantly from cycle to cycle to accommodate different needs. Argonne National Laboratory (ANL or Argonne) Reactor Conversion (RC) team has collaborated with SCK CEN for over a decade on the conversion of domestic and international research reactors from highly enriched uranium (HEU, ≥20 wt.% of 235 U) to low enriched uranium (LEU, <20 wt.% of 235 U) fuel. The U.S. High-Performance Research Reactor (USHPRR) project within the M3 Reactor Conversion Program aims at converting five U.S. high performance research reactors (MITR, MURR, NBSR, HFIR, and ATR) and one critical facility (ATR-C) to LEU fuel. These USHPRRs still use and regularly refuel with HEU fuel. Each facility has a unique reactor design, operating conditions, and fuel element design to accomplish its mission. The goal of the USHPRR project is to convert the USHPRRs and the critical facility to LEU fuel while maintaining experimental performance and ensuring safe facility operation. The current technical report focuses on two reactors requiring very high-density LEU fuel: the Massachusetts Institute of Technology Reactor (MITR) and the National Bureau of Standards Reactor (NBSR). To support the conversion of these reactors, so-called design demonstration elements (DDE) are planned to be irradiated in the BR2 reactor under conditions similar to the targeted reactors and using a prototypic geometry. In support of this experiment, SCK CEN studied and modeled the DDE irradiations using MCNP6.2 to investigate the feasibility of irradiating the MITR DDE and NBSR DDE in BR2. Argonne reviewed and confirmed the conclusions of this study. Structural analysis is another step toward converting USHPRR to LEU fuel. The objective of the current report is to provide information useful to the structural analysis of the NBSR & MITR DDEs to support its irradiation in BR2. Specifically, the goal is to provide the fast neutron (E>0.1MeV) fluence in the cladding of the fuel plates in BR2 for the whole period of irradiation (8 cycles for MITR DDE and 10 cycles for NBSR DDE). Additionally, fast neutron fluences in the side plates and in the NBSR DDE’s outside plates were calculated and reported. Neutronic calculations were performed using MCNP6.2 on the RTRHPC cluster.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

JM-1 SLOWPOKE-II Research Reactor Supporting Analysis Report for the Conversion from HEU to LEU

The JM-1 research reactor (SLOWPOKE-2 type) at the International Centre for Nuclear and Environmental Sciences (ICENS) of the University of West Indies, in Kingston (Jamaica), was successfully converted from HEU to LEU fuel on September 2015. As part of the work to produce the corresponding Safety Analysis Report (SAR), which was to be reviewed and approved by the Jamaican regulator in order to provide authorization for the reactor commissioning, a detailed safety analysis was carried out. This analysis, involving a variety of calculations, is summarized in the present report; describing the basic methodology and the obtained results. The report covers both the neutronic and thermal hydraulic aspects of the analysis, for which MCNP and PLTEMP/ANL codes, respectively, were used. On one hand, the HEU core was modeled and analyzed, subsequently comparing the obtained results with the available experimental data. On the other hand, the modifications involved with the change to an LEU core were introduced into the models and the results examined to ensure that the differences in the expected behavior did not negatively affect the reactor safety or performance. In fact, the LEU core was found to present significant advantages in different areas. Consequently, the present report provided a solid basis for the SAR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Used Fuel Management System Interface Analyses - 13578

Preliminary system-level analyses of the interfaces between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Used Fuel Management System Architecture Evaluation, Fiscal Year 2012

In the 1990s the U.S. Department of Energy (DOE) completed a number of system analyses investigating consolidated interim storage as a part of the waste management solution. These analyses are “dated” and do not reflect the present situation regarding at-reactor used nuclear fuel (UNF) management, alternatives for away from reactor management of used nuclear fuel, and alternatives for the ultimate disposal of UNF. The Blue Ribbon Commission for America’s Nuclear Future and the Nuclear Waste Technical Review Board have also pointed out the need for such analyses. The Used Fuel Management System Architecture Evaluation effort provides the DOE and others with information regarding the various alternatives for managing UNF generated by the current fleet of light water reactors operating in the U.S. The objectives are to: • Provide quantitative information with respect to a broad range of UNF management alternatives and considerations • Develop an integrated approach to evaluating storage, transportation, and disposal options, with emphasis on flexibility • Evaluate impacts of storage choices on UNF storage, handling, and disposal options • Identify alternative strategies and evaluate with respect to cost and flexibility • Consider a broad range of factors including repository emplacement capability, thermal constraints, repackaging needs, storage and transportation alternatives, and impacts on utility operations. In Fiscal Year 2012 system-level analyses of the overall interface between at-reactor, consolidated storage, and ultimate disposition along with the development of supporting logistic simulation tools were initiated. The objective of the Fiscal Year 2012 effort was two-fold: 1) develop methodologies, approaches, and tools (capability development), and 2) evaluate select UNF disposition scenarios (capability demonstration). The scenarios chosen for evaluation and the assumptions, inputs, and boundary conditions selected allowed for an initial set of analyses to gain insight regarding integrated system dynamics and an understanding of trends. This initial set of analyses also points to where additional system architecture analyses should focus.

Nutt, Mark↗