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.

scholarly journals Demonstration of Multi-Physics Evaluation of Fuel Rod Burst Probability With Burnup Extension Under LBLOCA Conditions

2020 ◽  
Author(s):  
Hongbin Zhang ◽  
Cole Blakely ◽  
Jianguo Yu
Keyword(s):  
Operating Time ◽  
Economic Benefits ◽  
Pressurized Water ◽  
Fuel Rod ◽  
Current Limit ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Accident Conditions ◽  
Operating Cycles

Abstract Extending the fuel discharge burnup level, e.g., from the current limit of rod averaged discharge burnup limit of 62 GWD/MT to a proposed new limit of 75 GWD/MT, can provide significant economic benefits to the current fleet of operating light water reactors (LWRs). It allows for longer operating cycles and improved resource utilization. The major economic gain of longer operating cycles is attributable to the increased capacity factor resulting from decreased refueling time as a fraction of total operating time, as well as fewer assemblies to be discharged for a given amount of energy produced. The main licensing challenges for higher burnup fuel are to ensure fuel rod safety under design basis accident conditions, especially under large-break loss-of-coolant accident (LBLOCA) and reactivity insertion accident (RIA). In this work, two-year cycle core design for a typical 4-loop pressurized water reactor (PWR) is performed with enrichment increased up to 6% and burnup extended to 75 GWD/MT. The fuel rod burst potential evaluations under large-break loss-of-coolant accident (LBLOCA) conditions are subsequently performed using the multi-physics best estimate plus uncertainty analysis framework LOTUS (LOCA Toolkit for the U.S. LWRs) and the preliminary results are presented.

A Preliminary Safety Analysis of the Whole Assembly Seed-Blanket Heterogeneous PWR Core

2004 ◽  
Author(s):  
Jeongik Lee ◽  
Pradip Saha ◽  
Mujid S. Kazimi ◽  
Won-Jae Lee
Keyword(s):  
Temperature Rise ◽  
Safety Analysis ◽  
Complete Loss ◽  
Pressurized Water Reactors ◽  
Primary Flow ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Safety Limit

The “Whole Assembly Seed and Blanket” (WASB) design, which utilizes mostly thorium in the blanket, consists of 84 seed and 109 blanket assemblies which may be backfitted into existing Pressurized Water Reactors (PWRs). Since the seed assemblies produce significantly more power than the blanket assemblies, a preliminary safety analysis of this design has been performed. Three accidents/transients (Large Break Loss of Coolant Accident (LBLOCA), Complete Loss of Primary Flow (LOPF) and Loss of Off-site Power (LOSP)), have been analyzed for both the WASB design and a typical all UO2 design for a typical 4-Loop Westinghouse PWR plant. LBLOCA results show that the peak cladding temperature (PCT) for the WASB design is approximately 260 K higher than that for a typical PWR design. However, this higher PCT for the WASB design is still about 200 K lower than the present regulatory safety limit. The response of the WASB and all UO2 core for LOPF and LOSP transients are very similar, and no post-DNB type rapid cladding temperature rise was observed in either of the two calculations.

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 Simulation of Mechanical Behavior of Fuel Pellets With Different Geometries

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Amanda Abati Aguiar ◽  
Danilo Faria ◽  
José Berretta ◽  
Paulo Afonso Rodi ◽  
Marcelo Santos ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Exothermic Reaction ◽  
Mechanical Analysis ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Pellets ◽  
Iron Based ◽  
Water Reactors ◽  
Accident Conditions

Typical Pressurized Water Reactors (PWR) fuel rods are manufactured using zirconium-based alloys as cladding and slightly enriched UO2 sintered pellets as fuel. However, in the last years efforts have been made to develop Accident Tolerant Fuels (ATF) focusing mainly in new materials to replace the cladding in order to avoid the exothermic reaction with steam experienced by zirconium-based alloys under accident conditions as observed during the Fukushima Daiichi accident. In this sense, iron-based alloys appear as a possibility to replace conventional zirconium-based alloys, and the effect of the pellet geometry in the performance of iron-based alloys fuel rods shall be investigated. The fuel pellet geometry experiences changes due to irradiation can promote early gap closure, mechanical loadings to the cladding and/or bamboo effects due to the combination of loads and irradiation creep, and all these effects depend also on the cladding properties. The objective of this paper was to address the influence of geometric parameters in the fuel pellet behavior of a stainless steel fuel rod by means of structural mechanical analysis using the well-known ANSYS software. The parameters evaluated in this paper considered fuel pellet with and without chamfer and dish. The data related to the fuel pellet performance under irradiation were obtained using a modified version of the FRAPCON code considering stainless steel as cladding. Results obtained from mechanical evaluation considering the effects through the responses of the axial, radial, plastic deformations, and resulting tensions were evaluated.

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