scholarly journals Analysis of Computer Modelling Results on Fuel Rods Strength and Condition at Reduced or Absent Cooling Caused by Accident

2021 ◽  
Vol 7 (1) ◽  
pp. 7-16
Author(s):  
Stepan Lys ◽  
Keyword(s):  
Stainless Steel ◽  
Fuel Assembly ◽  
Computer Modelling ◽  
Severe Accident ◽  
Research Centre ◽  
Steel Component ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Standard Problem ◽  
Post Test

The paper describes the phenomenology of fuel rod behaviour in severe accident. As an example, an experiment is described resulting in severe damage of 19 fuel rod assembly of VVER type; it was carried out in the CORA facility in 1993 (Research Centre, Karlsruhe, Germany). Testing conditions and results of post-test investigations of fuel assembly are given. The fuel rod code RAPTA-SFD is briefly dealt with; the code was a participant in the International Standard Problem ISP-36. The basic results are presented acquired by computer modelling CORA-W2 experiment using RAPTA-SFD code. Among the presented experimentally acquired and calculated results, the scope of the data on stainless steel component behaviour is substantial. The tested CORA-W2 fuel assembly contained a significant quantity of steel components, viz., spacer grids, a guide thimble, and a cladding of an absorber element. It is to be borne in mind that the spacer grids and a guide thimble of the updated and upgraded fuel assembly of VVER-1000 are fabricated from Zr-alloy, hence, the relative quantitative characteristics of chemical interactions between materials and stainless steel (Cr-Ni alloy) will be much lower for the up-to-date upgraded fuel assembly under identical conditions.

Investigation Into the Influence of Fuel Rods on the Anti-Seismic Mechanical Characteristics of Fuel Assembly Spacer Grid by FEM Method

2017 ◽  
Author(s):  
Jingya Sun ◽  
Yu Dang ◽  
Song Liu ◽  
Jiazheng Liu ◽  
Libing Zhu
Keyword(s):  
Fuel Assembly ◽  
Simulation Method ◽  
Experimental Results ◽  
Spacer Grid ◽  
Fuel Rods ◽  
Fem Model ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Fem Method ◽  
Crush Strength

The anti-seismic performance of fuel assembly is mainly determined by the critical crush load and the stiffness of spacer grids. To comprehensive know about the influence of fuel rods on the spacer grid, a 5×5 spacer grid FEM model which including fuel rods is established. Basing the fact that the grid spring has remarkable influence on the grid crush strength which is found in experiment, some cases are carried out, which are used to analyze effects of grid with/without fuel rod, friction between the grid spring/dimple and the fuel rod, the deflection of grid spring on the static buckling strength. Results show that grids with fuel rods will have higher crush strength than those without fuel rods; at certain range, increasing grid spring deflection at working point will do help to increase the grid crush strength; higher friction coefficient of grid spring and fuel rod can enhance the crush strength. Comparing with experimental results in literatures, results from simulations show the same tendency with the experimental results. The conclusion and the simulation method involved in this paper can provide some guidelines to optimize the performance of spacer grid assembly.

Flow-Induced Grid-to-Rod Fretting Test and Simulation for NFI PWR Fuel Assembly

2013 ◽  
Author(s):  
Shota Okui ◽  
Yuichiro Kubo ◽  
Shumpei Kakinoki ◽  
Roger Y. Lu ◽  
Zeses Karoutas ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Fuel Assembly ◽  
Wear Test ◽  
Flow Induced Vibration ◽  
Fuel Rods ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Non Linear ◽  
Computational Fluid Dynamics Cfd

A long-term flow-induced vibration and wear test was performed for a full-scale 17×17 PWR fuel mockup, and the test results were compared with numerical simulations. The flow-induced vibration on a fuel assembly or fuel rods may cause Grid-to-Rod Fretting (GTRF) and result in the leakage of fuel rods in PWRs. GTRF involves non-linear vibration of a fuel rod due to the excitation force induced by coolant flow around a fuel rod. So, the numerical simulation is performed by VITRAN (Vibration Transient Analysis Non-linear) and Computational Fluid Dynamics (CFD). VITRAN code was developed by Westinghouse to simulate fuel rod flow induced vibration and GTRF. In this paper, it was confirmed that the code can reproduce GTRF wear for NFI fuel assembly. CFD calculation is performed to obtain the axial and lateral flow velocity around the fuel rods, reflecting detailed geometries of fuel assembly components like bottom nozzle, spacer grids. The numerical simulation reasonably reproduced the vibration and wear test for NFI fuel assembly.

Improvement of SAMPSON Code for BWR Severe Accident Analysis

2013 ◽  
Author(s):  
Maolong Liu ◽  
Yuki Ishiwatari ◽  
Koji Okamoto
Keyword(s):  
Stainless Steel ◽  
Oxidation Reaction ◽  
Reactor Core ◽  
Severe Accident ◽  
Accident Analysis ◽  
Light Water Reactors ◽  
Fuel Cladding ◽  
Fuel Rod ◽  
Water Reactors ◽  
The Impact

The SAMPSON code has been developed in the IMPACT project in Japan to investigate severe accident phenomena for light water reactors. It integrates various analysis modules into a single code. The authors improved the fuel rod heat-up module of SAMPSON code by modeling the oxidation reaction of various core structures, including Zircaloy, stainless steel and B4C. And the creep failures of the Zircaloy fuel cladding and stainless steel monitoring guide tubes of the source range monitor (SRM) in the reactor core was also modeled for severe accident analysis.

Analysis of the Quench-14 Bundle Test With M5® Cladding

2009 ◽  
Author(s):  
Jonathan C. Birchley ◽  
Bernd Jaeckel ◽  
Timothy J. Haste ◽  
Martin Steinbrueck ◽  
Juri Stuckert
Keyword(s):  
Severe Accident ◽  
Local Version ◽  
Test Protocol ◽  
Pressurized Water ◽  
Fuel Rod ◽  
Post Test ◽  
The Subject ◽  
Water Reactors ◽  
Accident Conditions ◽  
Cladding Material

The QUENCH experimental programme at Forschungszentrum Karlsruhe (FZK) investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions, but where the geometry is still mainly rod-like and degradation is still at an early phase. The QUENCH test bundle is electrically heated and consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The cladding and grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO2 pellets. Experiment QUENCH-14 was successfully performed at FZK in July 2008 and is the first in this programme where Zr-Nb alloy M5® is used as the fuel rod simulator cladding. QUENCH-14 was otherwise essentially the same as experiment QUENCH-06, which was the subject of the CSNI ISP-45 exercise. It is also the first of three experiments in the QUENCH-ACM series, recently launched to examine the effect of advanced cladding materials on oxidation and quenching under otherwise similar conditions. Pre- and post-test analyses were performed at PSI using a local version of SCDAP/RELAP5 and MELCOR 1.8.6, using input models which had already been benchmarked against QUENCH-06 data. Preliminary pre-test calculations with both codes and alternative correlations for the oxidation kinetics indicated that the planned test protocol would achieve the desired objective of exhibiting whatever effects might arise from the change in cladding-material in the course of a transient similar to QUENCH-06. Several correlations were implemented in the models, namely Cathcart-Pawel, Urbanic-Heidrick, Leistikow-Schanz and Prater-Courtright for Zircaloy-4 (Zry-4), and additionally a new candidate correlation for M5® based on recent separate-effects tests performed at FZK on M5® cladding samples. Analyses of the QUENCH-14 data demonstrate strengths and limitations of the various models. Some tentative recommendations are made concerning choice of correlation and effect of cladding material.

Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

Kerntechnik ◽  
2013 ◽  
Vol 78 (4) ◽  
pp. 362-370
Author(s):  
H. György ◽  
I. Trosztel
Keyword(s):  
Severe Accident ◽  
Post Test

LDX 2101

Alloy Digest ◽  
2015 ◽  
Vol 64 (8) ◽  
Keyword(s):  
Stainless Steel ◽  
Corrosion Resistance ◽  
Physical Properties ◽  
Duplex Stainless Steel ◽  
Tensile Properties ◽  
Lower Cost ◽  
Heat Treating ◽  
General Purpose ◽  
Research Centre ◽  
Excellent Performance

Abstract LDX 2101 is a low-alloyed duplex stainless designed as a general-purpose duplex stainless steel. Designed for excellent performance at lower cost. This datasheet provides information on composition, physical properties, microstructure, hardness, and tensile properties as well as fatigue. It also includes information on corrosion resistance as well as forming, heat treating, machining, and joining. Filing Code: SS-1220. Producer or source: Outokumpu Stainless AB, Avesta Research Centre.

Degradation mechanism of stainless steel by U-Zr-O molten mixture during core degradation of BWR severe accident

2021 ◽  
pp. 1-14
Author(s):  
Ayumi Itoh ◽  
Nathan C. Andrews ◽  
David L. Luxat ◽  
Randall O. Gauntt ◽  
Masaki Kurata ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Degradation Mechanism ◽  
Severe Accident ◽  
Molten Mixture

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

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