scholarly journals Quantifying and Qualifying Alloys Based on Level of Homogenization: A U-10Mo Alloy Case Study

2019 ◽  
Vol 142 (1) ◽  
Author(s):  
Chao Wang ◽  
Zhijie Xu ◽  
Deborah Fagan ◽  
David P. Field ◽  
Curt Lavender ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Thermomechanical Processing ◽  
Measurement Data ◽  
Statistical Strategy ◽  
Minimum Number ◽  
Statistical Tolerance ◽  
Consistent Performance ◽  
Enriched Uranium ◽  
Formed Product ◽  
Molybdenum Concentration

Abstract Homogenization heat treatment is performed to attain uniformity in microstructure which is helpful to achieve the desired workability and microstructure in final products and, eventually, to gain predictive and consistent performance. Fabrication of low-enriched uranium alloys with 10 wt% molybdenum (U-10Mo) fuel plates involves multiple thermomechanical processing steps. It is well known that the molybdenum homogeneity in the final formed product affects the performance in the nuclear reactor. To ensure uniform homogenization, a statistical method is proposed to quantify and characterize the molybdenum concentration variation in U-10Mo fuel plates by analyzing the molybdenum concentration measurement data from scanning electron microscopy energy dispersive spectroscopy line-scan. Statistical tolerance intervals (TI) are employed to determine the qualification of the U-10Mo fuel plate. We formulate an argument for the minimum number of independent samples to define fuel plate qualification if no molybdenum measurement data are available in advance and demonstrate that the given TI requirements can be equivalently reduced to a sample variance criterion in this application. The outcome of the statistical analysis can be used to optimize casting design and eventually increase productivity and reduce fabrication costs. The statistical strategy developed in this paper can be implemented for other applications especially in the field of material manufacturing to assess qualification requirements and monitor and improve the process design.

Preliminary Core Design and Optimization Analysis of Travelling Wave Reactor

2014 ◽  
Vol 1070-1072 ◽  
pp. 357-360
Author(s):  
Dao Xiang Shen ◽  
Yao Li Zhang ◽  
Qi Xun Guo
Keyword(s):  
Monte Carlo ◽  
Rational Design ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Optimization Analysis ◽  
Design And Optimization ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Ignition Zone

A travelling wave reactor (TWR) is an advanced nuclear reactor which is capable of running for decades given only depleted uranium fuel, it is considered one of the most promising solutions for nonproliferation. A preliminary core design was proposed in this paper. The calculation was performed by Monte Carlo method. The burning mechanism of the reactor core design was studied. Optimization on the ignition zone was performed to reduce the amount of enriched uranium initially deployed. The results showed that the preliminary core design was feasible. The optimization analysis showed that the amount of enriched uranium could be reduced under rational design.

CAREM X INPRO Case Study

2004 ◽  
Author(s):  
Pablo C. Florido ◽  
Dari´o Delmastro ◽  
Daniel Brasnarof ◽  
Osvaldo E. Azpitarte
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
High Level Waste ◽  
Fuel Cycles ◽  
Surface Intermediate ◽  
Geological Repository ◽  
Enriched Uranium ◽  
Intermediate Storage ◽  
High Level

Argentina is performing CAREM X Nuclear System Case Study based on CAREM nuclear reactor and Once Through Fuel Cycle, using SIGMA for enriched uranium production, and a deep geological repository for final disposal of high level waste after surface intermediate storage in horizontal natural convection silos, to verify INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) methodology. Projections show that developing countries could play a crucial role in the deployment of nuclear energy, in the next fifty years. This case study will be highly useful for checking INPRO methodology for this scenario. In this contribution to ICONE 12, the preliminary findings of the Case Study are presented, including proposals to improve the INPRO methodology.

A Design-and-Build Project to Introduce Concepts Relating to Dimensional Variation

1998 ◽  
Vol 26 (4) ◽  
pp. 259-272
Author(s):  
S. M. Panton ◽  
P. R. Milner
Keyword(s):  
Classroom Environment ◽  
Tolerance Analysis ◽  
Cylinder Block ◽  
Worst Case ◽  
Good Design ◽  
Minimum Number ◽  
Dimensional Variation ◽  
Assembly Problem ◽  
Statistical Tolerance ◽  
Statistical Tolerance Analysis

A design-and-build project which has been used to introduce Year 2 students of Mechanical Engineering to the concepts of dimensional variation and the influence of dimensional variation on function and assembly. The project simulates the cylinder head cylinder block assembly problem and specifies requirements in terms of a tolerance on concentricity of the cylinders in the head and block, and the interchangeable assembly of the head and block. Materials which are easily and cheaply sourced and tools which are easily manufactured and safe to use in a classroom environment are used throughout. During the project the students are exposed to concepts such as worst-case and statistical tolerance analysis, sensitivity analysis, geometric moment effects, minimum constraint design, co-variance and gauging. The exercise also emphasizes that good design means components that function and assemble with the minimum number of tight tolerances.

scholarly journals Problem Decomposition and Information Minimization for the Global, Concurrent, On-line Validation of Neutron Noise Signals and Neutron Detector Operation

2020 ◽  
Vol 11 (6) ◽  
pp. 1-12
Author(s):  
Tatiana Tambouratzis
Keyword(s):  
Nuclear Reactor ◽  
Time Windows ◽  
Divide And Conquer ◽  
Neutron Detectors ◽  
Pressurized Water ◽  
Core Dataset ◽  
Neutron Noise ◽  
Minimum Number ◽  
On Line ◽  
Cross Correlations

This piece of research introduces a purely data-driven, directly reconfigurable, divide-and-conquer on-line monitoring (OLM) methodology for automatically selecting the minimum number of neutron detectors (NDs) – and corresponding neutron noise signals (NSs) – which are currently necessary, as well as sufficient, for inspecting the entire nuclear reactor (NR) in-core area. The proposed implementation builds upon the 3-tuple configuration, according to which three sufficiently pairwise-correlated NSs are capable of on-line (I) verifying each NS of the 3-tuple and (II) endorsing correct functioning of each corresponding ND, implemented herein via straightforward pairwise comparisons of fixed-length sliding time-windows (STWs) between the three NSs of the 3-tuple. A pressurized water NR (PWR) model – developed for H2020 CORTEX – is used for deriving the optimal ND/NS configuration, where (i) the evident partitioning of the 36 NDs/NSs into six clusters of six NDs/NSs each, and (ii) the high cross-correlations (CCs) within every 3-tuple of NSs, endorse the use of a constant pair comprising the two most highly CC-ed NSs per cluster as the first two members of the 3-tuple, with the third member being each remaining NS of the cluster, in turn, thereby computationally streamlining OLM without compromising the identification of either deviating NSs or malfunctioning NDs. Tests on the in-core dataset of the PWR model demonstrate the potential of the proposed methodology in terms of suitability for, efficiency at, as well as robustness in ND/NS selection, further establishing the “directly reconfigurable” property of the proposed approach at every point in time while using one-third only of the original NDs/NSs.

PBMR Fuel Kernel Model for the Prediction of Accurate Temperature Profiles

2008 ◽  
Author(s):  
Onno Ubbink ◽  
Pieter S. du Toit ◽  
Pierre Lourens ◽  
Wessel R. Joubert
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Fission ◽  
Calculation Model ◽  
Tennis Ball ◽  
Fuel Temperature ◽  
Fuel Kernel ◽  
Graphite Sphere ◽  
Enriched Uranium ◽  
Definition Of ◽  
Fuel Elements

PBMR (Pebble Bed Modular Reactor) is a High-Temperature Gas-cooled Reactor (HTGR) utilizing spherical pebble like fuel elements. A pebble is a moulded graphite sphere about the size of a tennis ball that contains approximately 15000 homogeneously distributed, triso-coated low-enriched uranium dioxide (UO2) particles, about 1mm in diameter. In the case of fast reactivity transients the accurate time-dependent calculation of the uranium temperature is essential as the neutron balances in the nuclear reactor are strongly influenced by the actual fuel temperature. This paper presents a calculation model that calculates the temperature profile through a representative fuel kernel, its coating layers and the associated graphite moderator. The local nuclear fission heat is deposited directly in the fuel itself. Great care is taken with the definition of the boundary conditions and implementation thereof to ensure that the kernel temperature calculation model describes the physics as accurately as possible. This paper reports on this in detail. A sample calculation is included to illustrate the effect of and need for a more accurate model.

High Temperature Intrinsically Safe Nuclear Reactor

2015 ◽  
Author(s):  
Robin J. McDaniel
Keyword(s):  
High Temperature ◽  
Alternative Energy ◽  
Nuclear Reactor ◽  
Cooling System ◽  
Clean Energy ◽  
Normal Operation ◽  
The Core ◽  
Small Modular Reactor ◽  
Loss Of Coolant ◽  
Minimum Number

Small Modular Reactor (SMR) technologies have been recently included by the DOE as clean energy, a low carbondioxide emitting “alternative energy” source. The objective of this paper is to further the discussion of intrinsically safe nuclear reactors in the context of passive safety design principles and introduction of a novel conceptual reactor design. After a multiple year research study of past fast neutron reactor designs and recent reactor proposals, the following design is the result of analysis of the best concepts discovered. An improved fast reactor of the liquid metal cooled type including a core configuration allowing for only two operational states, “Power” or “Rest”. The flow of the primary cooling fluid suspends the core in the “Power” state, with sufficient flow to remove the heat to an intermediate heat exchanger during normal operation. This design utilizes the force of gravity to shut down the reactor after any loss of coolant flow, either a controlled reactor shut down or a Loss of Coolant Accident (LOCA) event, as the core is controlled via dispersion of fuel elements. Electromagnetic pumps incorporating automatic safety electrical cut-offs are employed to shutdown the primary cooling system to disassemble the core to the “Rest” configuration due to a loss of secondary coolant or loss of ultimate heat sink. This design is a hybrid pool-loop pressurized high-temperature reactor unique in its use of a minimum number of components, utilizing no moving mechanical parts, no rotating seals, and no control rods. This defines an elegantly simple Gen IV intrinsically safe nuclear reactor. [Advanced Small Modular Reactor (aSMR)]

scholarly journals Powering Down

2003 ◽  
Vol 125 (04) ◽  
pp. 46-48
Author(s):  
Harry Hutchinson
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
The United States ◽  
Nuclear Industry ◽  
Significant Feature ◽  
Pressurized Water ◽  
Safety Testing ◽  
Fuel Assemblies ◽  
Enriched Uranium

This article reviews that after a half century of safety testing for the nuclear industry, a key heat-transfer lab is losing its home. Columbia University’s Heat Transfer Research Facility has been the only place to go for key safety testing. Since the days of the Atoms for Peace program during the Eisenhower years, the lab has tested generations of nuclear reactor fuel assemblies. The lab’s clients over the years have included all the designers of pressurized water reactors in the United States and others from much of the world. The tests are primarily concerned with one small, but significant feature of a reactor core. A core contains as many as 3000 fuel assemblies, bundles of long, slender rods containing enriched uranium. Controlled fission among the bundles heats water to begin the series of heat-transfer cycles that send steam to the turbines that will drive generators.

Thermal-Mechanical Analysis of Annular Target Design for High Volume Production of Molybdenum-99 Using Low-Enriched Uranium

2009 ◽  
Author(s):  
Kyler K. Turner ◽  
Gary L. Solbrekken ◽  
Charlie W. Allen
Keyword(s):  
Contact Resistance ◽  
Thermal Contact Resistance ◽  
Nuclear Reactor ◽  
Thermal Contact ◽  
Cost Effective ◽  
High Volume ◽  
The United States ◽  
Mechanical Analysis ◽  
Effective Manner ◽  
Enriched Uranium

Techenetium-99m is a diagnostic radioactive medical isotope that is currently used 30,000 times a day in the United States. All supplies of techenetium-99m’s parent isotope molybdenum-99 currently originate from nuclear reactor facilities located in foreign countries and use highly enriched uranium (HEU). In accordance with the Global Threat Reduction Initiative all uranium used in future molybdenum-99 production will use low enriched uranium (LEU). A design approach to using LEU in a cost-effective manner is to use a target that is based on LEU foil. A potential failure mode for the LEU foil based target is temperature excursion during irradiation due to poor thermal contact between the foil and the target cladding. The purpose of this study is to establish the theoretical basis for experimentally measuring the thermal contact resistance. Replicating in service heating conditions is nearly impossible when testing the thermal contact resistance as part of a study to establish LEU foil warpage tolerances, thus it is necessary to establish an alternate heating configuration that will allow a conservative estimate of the contact resistance. Thermal and mechanical analysis suggests that external heating of an annular target will place the interface into a state that will over-estimate the contact resistance relative to use conditions. Further, the magnitude of the heat load used for testing can be adjusted to control the degree of overestimation.

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