scholarly journals TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

1990 ◽  
Author(s):  
N.J. Lombardo ◽  
T.J. Marseille ◽  
M.D. White ◽  
P.S. Lowery
Keyword(s):  
Nuclear Fuel ◽  
Computer Code ◽  
Severe Accident ◽  
Fuel Assemblies ◽  
Accident Conditions

Simulation of Nuclear Fuel Behavior in Accident Conditions With the DIONISIO Code

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Martín Lemes ◽  
Alicia Denis ◽  
Alejandro Soba
Keyword(s):  
Nuclear Fuel ◽  
Computer Code ◽  
Hydraulic Modeling ◽  
Application Range ◽  
Behavior Simulation ◽  
External Conditions ◽  
Capture And Release ◽  
Fuel Behavior ◽  
Accident Conditions ◽  
Thermal Hydraulic Modeling

DIONISIO is a computer code designed to simulate the behavior of one nuclear fuel rod during its permanence within the reactor. Starting from the power history and the external conditions to which the rod is subjected, the code predicts all the meaningful variables of the system. Its application range has been recently extended to include accidental conditions, in particular the so-called loss of coolant accidents (LOCA). In order to make realistic predictions, the conditions in the rod environment have been taken into account since they represent the boundary conditions with which the differential equations describing the fuel phenomena are solved. Without going into the details of the thermal-hydraulic modeling, which is the task of the specific codes, a simplified description of the conditions in the cooling channel during a LOCA event has been developed and incorporated as a subroutine of DIONISIO. This has led to an improvement of the fuel behavior simulation, which is evidenced by the considerable number of comparisons with experiments carried out, many of them reported in this paper. Moreover, this work describes a model of high temperature capture and release of hydrogen in the nuclear fuel cladding, in scenarios typical of LOCA events. The corresponding computational model is being separately tested and will be next included in the DIONISIO thermal-hydraulic module.

Simulation of Gas Stratification Build-Up in the Containment Under Severe Accident Conditions

2014 ◽  
Author(s):  
Michele Andreani
Keyword(s):  
Structural Integrity ◽  
Computer Code ◽  
Severe Accident ◽  
Test Sequence ◽  
Small Scale ◽  
High Concentration ◽  
Initial Transient ◽  
Loss Of Coolant Accident ◽  
Computer Codes ◽  
Accident Conditions

The presence of hydrogen stratification in a NPP containment in the case of a severe accident is a source of concern, as pockets of the gas in high concentration could lead to a deflagration or detonation risk, which might challenge the containment structural integrity. These issues, as well as the capability of various computer codes to predict the evolution of a representative accident, are addressed in the coordinated projects ERCOSAM of the 7th EURATOM FWP and the project SAMARA sponsored by ROSATOM. The projects aim to establish whether in a test sequence representative of a severe accident in a LWR hydrogen stratification can be established during the initial transient following a loss of coolant accident (LOCA) and whether and how this stratification can be broken down by the operation of Severe Accident Management systems (SAMs): sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Experiments with helium (as simulant of hydrogen) have been performed at “small scale” in TOSQAN (IRSN, Saclay), and “medium scale” in the MISTRA (CEA, Saclay), PANDA (PSI, Villigen) and SPOT ((JSC “Afrikantov OKBM”, Nizhny Novgorod) facilities. The present paper presents the analysis of the initial transient of some tests in the PANDA, TOSQAN and SPOT facilities using the GOTHIC computer code. The work therefore addresses the capability of the code and a relatively coarse mesh to simulate the pressurisation and build-up of steam and helium stratification for conditions representative of a postulated severe accident scenario, properly scaled to the various facilities. The prediction of the pressurisation is excellent, and the position of the gas concentration stratification front at the end of the steam and helium releases is generally well captured.

scholarly journals Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

2017 ◽  
Vol 495 ◽  
pp. 49-57 ◽  
Author(s):  
E. Geiger ◽  
C. Le Gall ◽  
A. Gallais-During ◽  
Y. Pontillon ◽  
J. Lamontagne ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Fission Products ◽  
Severe Accident ◽  
Accident Conditions

Steam and Air Oxidation Behavior of Nuclear Fuel Claddings at Severe Accident Conditions

2010 ◽  
Vol 1264 ◽  
Author(s):  
Mirco Große ◽  
Martin Steinbrück ◽  
Juri Stuckert
Keyword(s):  
Nuclear Fuel ◽  
Oxidation Behavior ◽  
Zirconium Alloys ◽  
Severe Accident ◽  
Air Oxidation ◽  
Air Atmosphere ◽  
Parabolic Kinetics ◽  
Oxygen Starvation ◽  
Accident Conditions ◽  
Local Oxygen

AbstractThe oxidation behavior of zirconium alloys used as materials for nuclear fuel rod claddings is investigated in the temperature range between 973 and 1673 K in steam and air atmosphere. Parabolic kinetics was found for all materials, atmospheres and temperatures, at least at beginning of the reactions. The temperature dependence of the reaction rate is of Arrhenius type. The parameters of the Arrhenius functions are determined and given for steam oxidation. Due to the formation of a large amount of cracks an acceleration of the reactions can occur. Reasons of the crack formations are phase transformations in the oxide layer known as the breakaway effect and, in case of air atmosphere, local oxygen starvation conditions resulting in reactions with nitrogen. The paper gives a short overview of the relevant mechanisms and processes.

scholarly journals Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

2017 ◽  
Vol 495 ◽  
pp. 363-384 ◽  
Author(s):  
Y. Pontillon ◽  
E. Geiger ◽  
C. Le Gall ◽  
S. Bernard ◽  
A. Gallais-During ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Fission Products ◽  
Severe Accident ◽  
Lessons Learnt ◽  
Accident Conditions

Simulation of the HDR E11.2 Hydrogen Mixing Experiment With the CONTAIN Code

2002 ◽  
Author(s):  
Aljaz Skerlavaj ◽  
Ivo Kljenak
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Computer Code ◽  
Severe Accident ◽  
Test Facility ◽  
Hydrogen Distribution ◽  
Flow Geometry ◽  
Input Model ◽  
Containment Test ◽  
Accident Conditions

One of the most well-known experiments on atmosphere stratification in a nuclear power plant containment at severe accident conditions is the test E11.2 “Hydrogen distribution in loop flow geometry”, which was performed in the Heissdampf Reaktor containment test facility in Germany. In the present work, the simulation of the test E11.2 with the CONTAIN computer code is presented. An input model consisting of 72 cells and 263 flowpaths was developed. The predicted pressure history and thermal stratification agree relatively well with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was somewhat underestimated.

Simulation of Aerosol Behavior in a Saturated Atmosphere With the CONTAIN Code

2002 ◽  
Author(s):  
Ivo Kljenak ◽  
Borut Mavko
Keyword(s):  
Nuclear Power ◽  
Saturated Vapor ◽  
Computer Code ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Experimental Facility ◽  
Standard Problem ◽  
Model Aerosol ◽  
Saturated Atmosphere ◽  
Accident Conditions

Experiments on aerosol behavior in an atmosphere containing saturated vapor, which were performed in the KAEVER experimental facility and proposed for the OECD International Standard Problem No. 44, were simulated with the CONTAIN thermal-hydraulic computer code. The purpose of the work was to assess the capability of the CONTAIN code to model aerosol condensation and deposition in a containment of a light-water-reactor nuclear power plant at severe accident conditions. Results of dry and wet aerosol concentrations in the test vessel atmosphere are presented and analyzed.

Modeling and Phenomenological Aspects of Severe Accidents in Integral Pressurized Water Reactors

2011 ◽  
Author(s):  
I. K. Madni ◽  
M. Khatib-Rahbar
Keyword(s):  
Computer Code ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Modeling Framework ◽  
Pressurized Water ◽  
Product Release ◽  
Severe Accidents ◽  
Water Reactors ◽  
Accident Conditions ◽  
Coolant System

This paper focuses on modeling and phenomenological issues relevant to analysis of severe accidents in integral Pressurized Water Reactors (iPWRs). It identifies relevant thermal-hydraulics, melt progression and fission product release and transport phenomena, and discusses the applicability of the MELCOR computer code to modeling of severe accidents in iPWRs. Areas where the current MELCOR severe accident modeling framework has limitations in the representation of phenomenological processes are identified and examples of possible modeling remedies are discussed. The paper identifies modeling and phenomenological issues that contribute to differences in the calculated reactor coolant system and containment response for iPWRs as compared to traditional PWRs under severe accident conditions.

Safety Analysis of Irradiated Nuclear Fuel Transportation Container

2007 ◽  
Author(s):  
E. Uspuras ◽  
S. Rimkevicius
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Analysis ◽  
Normal Operation ◽  
Spent Fuel ◽  
Safety Level ◽  
Fuel Assemblies ◽  
Accident Conditions ◽  
Transportation Container

Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit 1 on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level.

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