scholarly journals Hydrogen Isotopes Permeation in Clean or Unoxidized FeCrAl Alloys: A Review

2022 ◽  
Author(s):  
Yogendra S. Garud ◽  
Andrew K. Hoffman ◽  
Raul B. Rebak
Keyword(s):  
Hydrogen Permeation ◽  
Austenitic Stainless Steels ◽  
Major Elements ◽  
Nuclear Industry ◽  
Department Of Energy ◽  
Loss Of Coolant Accident ◽  
Austenitic Alloys ◽  
Fecral Alloys ◽  
Water Reactors ◽  
Accident Scenarios

AbstractThe US Department of Energy is working with fuel vendors to develop accident tolerant fuels (ATF) for the current fleet of light water reactors (LWRs). The ATF should be more resilient to loss of coolant accident scenarios and help extending the life of the operating LWRs. One of the proposed ATF concepts is to use iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel. A concern in using ferritic FeCrAl is that this type of cladding may result in an increase in the concentration of tritium in the coolant. The objective of the current critical review is to collect and assess information from the literature regarding diffusion or permeation of hydrogen (H) and its isotopes deuterium (D) and Tritium (T) across industrial alloys (including FeCrAl) used or intended for the nuclear industry. Over a hundred years of data reviewed shows that the solubility of hydrogen in ferritic alloys is lower than in austenitic alloys but hydrogen permeates faster through a ferritic material than through austenitic materials. The tritium permeation rates in FeCrAl alloys are between those in austenitic stainless steels and in ferritic FeCr steels. The activation energy for hydrogen permeation is approximately 30 pct higher in the austenitic alloys compared with the ferritic (typically ∼ 50 kJ/mol in ferritic vs. ∼ 65 kJ/mol in the austenitic). None of the major elements in FeCrAl alloys react with hydrogen to form detrimental hydride phases. The effect of surface oxides on FeCrAl delaying hydrogen entrance into FeCrAl alloy is not part of this review.

scholarly journals Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

Coatings ◽  
2021 ◽  
Vol 11 (5) ◽  
pp. 557
Author(s):  
Egor Kashkarov ◽  
Bright Afornu ◽  
Dmitrii Sidelev ◽  
Maksim Krinitcyn ◽  
Veronica Gouws ◽  
...  
Keyword(s):  
Protective Coatings ◽  
Zirconium Alloys ◽  
Nuclear Industry ◽  
Normal Operation ◽  
Severe Accidents ◽  
Recent Developments ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Directed Research

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed.

FeCrAl Alloys for Accident Tolerant Fuel Cladding in Light Water Reactors

2016 ◽  
Author(s):  
Raul B. Rebak ◽  
Kurt A. Terrani ◽  
Russ M. Fawcett
Keyword(s):  
Fuel Assembly ◽  
Department Of Energy ◽  
Normal Operation ◽  
Fuel System ◽  
Light Water Reactors ◽  
Light Water ◽  
Oak Ridge ◽  
Operation Conditions ◽  
Fecral Alloys ◽  
Water Reactors

The goal of the U.S. Department of Energy (DOE) Accident Tolerant Fuel Program (ATF) for light water reactors (LWR) is to identify alternative fuel system technologies to further enhance the safety of commercial nuclear power plants. An ATF fuel system would endure loss of cooling in the reactor for a considerably longer period of time than the current systems. The General Electric (GE) and Oak Ridge National Laboratory (ORNL) ATF design concept utilizes an iron-chromium-aluminum (FeCrAl) alloy material as fuel rod cladding in combination with uranium dioxide (UO2) fuel pellets currently in use, resulting in a fuel assembly that leverages the performance of existing/current LWR fuel assembly designs and infrastructure with improved accident tolerance. Significant testing was performed in the last three years to characterize FeCrAl alloys for cladding applications, both under normal operation conditions of the reactor and under accident conditions. This article is a state of the art description of the concept.

scholarly journals Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

2017 ◽  
Vol 2017 ◽  
pp. 1-13 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Hao-ran Ju
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Safety Issue ◽  
Primary System ◽  
Light Water Reactors ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Significant Difference ◽  
Water Reactors ◽  
Accident Scenarios

Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA) is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR). RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches) is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

Hydrogen Diffusion in FeCrAl Alloys for Light Water Reactors Cladding Applications

2016 ◽  
Author(s):  
Raul B. Rebak ◽  
Young-Jin Kim
Keyword(s):  
Hydrogen Diffusion ◽  
Zirconium Alloys ◽  
The United States ◽  
Department Of Energy ◽  
Light Water Reactors ◽  
Alumina Layer ◽  
Fecral Alloys ◽  
Water Reactors ◽  
Tritium Permeation ◽  
Storage Pools

There is a worldwide effort to develop nuclear fuels that are resistant to accidents such as loss of coolant in the reactor and the storage pools. In the United States, the Department of Energy is teaming with fuel vendors to develop accident tolerant fuels (ATF), which will resist the lack of cooling for longer periods of times than the current zirconium alloy - uranium dioxide system. General Electric (GE) and its partners is proposing to replace zirconium alloys cladding with an Iron-Chromium-Aluminum (FeCrAl) alloy such as APMT, since they are highly resistant to attack by steam up to the melting point of the alloy. FeCrAl alloys do not react with hydrogen to form stable hydrides as zirconium alloys do. Therefore, it is possible that more tritium may be released to the coolant with the use of FeCrAl cladding. This work discusses the formation of an alumina layer on the surface of APMT cladding as an effective barrier for tritium permeation from the fuel to the coolant across the cladding wall.

Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients

2002 ◽  
Author(s):  
Ruwan K. Ratnayake ◽  
S. Ergun ◽  
L. E. Hochreiter ◽  
A. J. Baratta
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Reactors ◽  
Computer Code ◽  
Boiling Water Reactors ◽  
Validation Process ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Accident Scenarios

In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena.

scholarly journals Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

2021 ◽  
Vol 9 (2A) ◽  
Author(s):  
Claudia Giovedi ◽  
Alfredo Abe ◽  
Rafael O. R. Muniz ◽  
Daniel S. Gomes ◽  
Antonio T. Silva ◽  
...  
Keyword(s):  
New Materials ◽  
Fuel Performance ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fukushima Daiichi ◽  
Iron Based ◽  
Water Reactors ◽  
Accident Scenarios

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron-based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes.

ASTEC: An Integral Code for Simulation of Severe Light Water Reactor Accidents

2006 ◽  
Author(s):  
N. Reinke ◽  
K. Neu ◽  
H.-J. Allelein
Keyword(s):  
Fission Products ◽  
Feasibility Studies ◽  
Current Status ◽  
Light Water ◽  
Fast Running ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Station Blackout ◽  
Water Reactors ◽  
Reactor Types

The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. Thus, the main fields of ASTEC application are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 studies as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses, as well as a combination of multiple modules for coupled effects testing and integral analyses. Among activities concentrating on the validation of individual ASTEC modules describing specific phenomena, the applicability to reactor cases marks an important step in the development of the code. Feasibility studies on plant applications have been performed for several reactor types such as the German Konvoi PWR 1300, the French PWR 900, and the Russian VVER-1000 and −440 with sequences like station blackout, small- or medium-break loss-of-coolant accident, and loss-of-feedwater transients. Subject of this paper is a short overview on the ASTEC code system and its current status with view to the application to severe accidents sequences at several PWRs, exemplified by selected calculations.

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
S. P. Saraswat ◽  
P. Munshi ◽  
A. Khanna ◽  
C. Allison
Keyword(s):  
Heat Flux ◽  
Cooling System ◽  
Heat Removal ◽  
Hydraulic Modeling ◽  
Carbon Fiber Composites ◽  
High Heat Flux ◽  
Vacuum Vessel ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Accident Scenarios

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

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