Fabrication and Assembly of the First Accident Tolerant Fuel Concept for TREAT Testing

2020 ◽  
Author(s):  
Connor Woolum ◽  
D. Devin Imholte ◽  
Austin Fleming ◽  
David Kamerman ◽  
Korbin Tritthart
Keyword(s):  
Phase I ◽  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Reactor Performance ◽  
Operating Characteristics ◽  
Test Vehicle ◽  
Pressurized Water ◽  
Vehicle System ◽  
Reactor Test ◽  
Initial Fuel

Abstract Following the tragic events at the Fukushima Daiichi power plant in 2011, priority was given to increasing the accident tolerance of fuel systems for the current fleet of nuclear reactors. These enhanced Accident Tolerant Fuel (ATF) concepts include a wide variety of fuel and cladding materials, both as variants of the current Zircaloy-UO2 system and also as novel fuel and cladding concepts. In addition to testing at steady-state, prototypic, conditions within a nuclear reactor, performance of these ATF concepts in off-normal and transient conditions must be evaluated. The Transient Reactor Test (TREAT) facility at Idaho National Laboratory’s (INLs) Materials and Fuels Complex (MFC) was restarted in the Fall of 2017 and is well-suited to serve this purpose. September of 2018 marked the first fueled specimen to be tested in TREAT since its restart; testing of fuel specimens has been ongoing since then. Initial fuel tests focused on the traditional Zircaloy-UO2 fuel system in order to gain a more thorough understanding of operating characteristics of both the test vehicle system and also the interactions between the reactor and the experiment itself. These tests also served to commission new test vehicles using the well-characterized Zircaloy-UO2 system. The Separate Effects Test Holder, SETH Capsule, is a modular capsule designed such that it can support a wide variety of specimen geometries ranging from prototypic pressurized water reactor (PWR) fuel samples, heat sink based experiments, and more. The capsule itself is an additively manufactured titanium capsule, within which the experimental specimen is loaded. The SETH Phase I series of tests included five individual SETH capsules, each with a single fuel rodlet and instrumentation to measure temperature during irradiation in TREAT. Each fuel rodlet is representative of a fuel rod in a PWR, with UO2 in Zircaloy-4 cladding. In August of 2019, TREAT irradiated the first ATF candidate fuel, U3Si2. This marked the first transient test of an ATF concept and is part of a larger campaign that will irradiate a total of four capsules containing ATF concepts. This test campaign, SETH Phase II, built upon the previous SETH Phase I campaign with a nearly identical design except for the fuel rodlet itself. Two of the four SETH capsules contained U3Si2 fuel within Zircaloy-4 cladding, and the other two capsules contained U3Si2 fuel within SiC cladding. This paper reviews the design, fabrication, and assembly efforts resulting in the four qualified SETH capsules for TREAT irradiation of these ATF concepts.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

Study on the Viability of the Recycling by Electric Arc Melting of Zirconium Alloys Scraps Aiming the Scalability of the Process

2018 ◽  
Vol 930 ◽  
pp. 495-500
Author(s):  
Cristiano Stefano Mucsi ◽  
L.A.M. dos Reis ◽  
Maurilio Pereira Gomes ◽  
L.A.T. Pereira ◽  
Jesualdo Luiz Rossi
Keyword(s):  
Nuclear Reactor ◽  
Zirconium Alloys ◽  
Vacuum Arc ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Vacuum Arc Remelting ◽  
Arc Melting ◽  
Vacuum Arc Furnace ◽  
Fuel Elements

Turning chips of zirconium alloys are produced in large quantities during the machining of alloy rods for the fabrication of the end plugs for the Pressurized Water Reactor (PWR) fuel elements parts of Angra II nuclear reactor (Brazil – Rio de Janeiro). This paper presents a study on the search for an efficient way for the cleaning, quality control and Vacuum Arc Remelting (VAR) of pressed zirconium alloys chips to produce a material viable to be used in the production of the fuel rod end plugs. The process starts with cutting oil clean out. The first step in this process consists in soaking a bunch of chips in clean water, to remove soluble cutting oils, followed by an alkaline degreasing bath and a wash with a high-pressure flow of water. Drying is performed by a flux of warm air. The oil free chips are then subjected to a magnet in order to detect and collect any magnetic material, essentially ferrous, that may be present in the original chips. Samples of the material are collected and then melted in a small non consumable electrode vacuum arc furnace for evaluation by Energy Dispersive X-ray Fluorescence Spectrometry (EDXRFS) in order to define the quality of the chips. The next step consists in the 15 ton hydraulic pressing the chips in a die with 40 mm square section and 500 mm long, producing an electrode with 20% of the Zircaloy bulk density. The electrode was finally melted in a laboratory scale modified VAR furnace located at the CCTM–IPEN, producing 0.8 kg ingots. The authors conclude that the samples obtained from the fuel element industry can be melting in a VAR furnace, modified to accommodate low density electrodes, allowing a reduction up to 40 times the original storage volume, however, it is necessary to remelt the ingots to correct their composition in order to recycle the original zirconium alloys chips. in a process to reduce volume and allow the reutilization of valuable Zircaloy scraps.

scholarly journals Finite Element Based Full-Life Cyclic Stress Analysis of 316 Grade Nuclear Reactor Stainless Steel Under Constant, Variable, and Random Fatigue Loading

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Bipul Barua ◽  
Subhasish Mohanty ◽  
Joseph T. Listwan ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Analysis ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Fatigue Loading ◽  
Cyclic Stress ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fatigue Evaluation

Although S∼N curve-based approaches are widely followed for fatigue evaluation of nuclear reactor components and other safety critical structural systems, there is a chance of large uncertainty in estimated fatigue lives. This uncertainty may be reduced by using a more mechanistic approach such as physics based three-dimensional (3D) finite element (FE) methods. In a recent paper (Barua et al., 2018, ASME J. Pressure Vessel Technol., 140(1), p. 011403), a fully mechanistic fatigue modeling approach which is based on time-dependent stress–strain evolution of material over the entire fatigue life was presented. Based on this approach, in this work, FE-based cyclic stress analysis was performed on 316 nuclear grade reactor stainless steel (SS) fatigue specimens, subjected to constant, variable, and random amplitude loading, for their entire fatigue lives. The simulated results are found to be in good agreement with experimental observation. An elastic-plastic analysis of a pressurized water reactor (PWR) surge line (SL) pipe under idealistic fatigue loading condition was performed and compared with experimental results.

scholarly journals Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

2011 ◽  
Vol 32 (4) ◽  
pp. 67-79
Author(s):  
Tomasz Bury
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Computer Code ◽  
Hydrogen Combustion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Lumped Parameter ◽  
Loss Of Coolant

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

MRS Advances ◽  
2016 ◽  
Vol 1 (35) ◽  
pp. 2495-2500
Author(s):  
Thomas Winter ◽  
James Huggins ◽  
Richard Neu ◽  
Preet Singh ◽  
Chaitanya S. Deo
Keyword(s):  
Nuclear Reactor ◽  
Wear Behavior ◽  
Sem Analysis ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Rod

ABSTRACTIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.

Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach

2015 ◽  
Author(s):  
Subhasish Mohanty ◽  
William K. Soppet ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Nuclear Reactor ◽  
Axial Stress ◽  
Cyclic Plasticity ◽  
Model Parameters ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Test Conditions ◽  
Over Time

At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S∼N) curves and Coffin-Manson type empirical relations. In most cases, the S∼N curves are generated from uniaxial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S∼N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.

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