RELAP5-3D Code Application for RBMK-1500 Reactor Core Analysis

2002 ◽  
Author(s):  
Evaldas Bubelis ◽  
Algirdas Kaliatka ◽  
Eugenijus Uspuras
Keyword(s):  
3D Model ◽  
Reactor Core ◽  
Core Analysis ◽  
Reactor Operation ◽  
The Core ◽  
Calculation Results ◽  
Plant Data ◽  
Comparison Of The Results ◽  
Real Plant ◽  
Good Agreement

The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed quite good mutual coincidence of the calculation results and good agreement with real plant data.

RELAP5-3D Code Validation for RBMK-1500 Reactors

2004 ◽  
Author(s):  
Evaldas Bubelis ◽  
Algirdas Kaliatka ◽  
Eugenijus Uspuras
Keyword(s):  
Reasonable Agreement ◽  
3D Model ◽  
Reactor Core ◽  
Measured Data ◽  
Primary Circuit ◽  
Hydraulic Parameters ◽  
Physical Processes ◽  
Code Validation ◽  
Calculation Results ◽  
Plant Data

This paper deals with RELAP5-3D code validation through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A successful best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.

Analysis of Dynamic Stress in the Reactor Core Barrel due to the Pressure Pulsations

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Iaroslav Dubyk ◽  
Anatolii Batura
Keyword(s):  
Nuclear Power ◽  
Dynamic Stress ◽  
Reactor Core ◽  
Pressure Pulsations ◽  
Power Station ◽  
Coolant Pump ◽  
The Core ◽  
Calculation Results ◽  
Full Power ◽  
Reactor Pressure

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

Core-polarization-corrected multiconfiguration Dirac–Fock oscillator strengths and excitation energies for the 5s2 1S0 – 5s5p 3P1, 1P1 transitions in Sr I and Y II spectra

1987 ◽  
Vol 65 (12) ◽  
pp. 1612-1619 ◽  
Author(s):  
J. Migdalek ◽  
W. E. Baylis
Keyword(s):  
Electron Correlation ◽  
Oscillator Strengths ◽  
Excitation Energies ◽  
Core Electron ◽  
Polarization Model ◽  
Core Polarization ◽  
The Core ◽  
Relativistic Configuration ◽  
Comparison Of The Results ◽  
Good Agreement

Energies and oscillator strengths for the spin-allowed 5s2 1S0 – 5s5p 1P1 and spin-forbidden 5s2 1S0 – 5s5p 3P1 transitions in neutral strontium and singly ionized yttrium are determined in relativistic multiconfiguration Dirac–Fock computations where modest relativistic configuration mixing to represent intravalence correlation is combined with a polarization model to account for valence–core electron correlation. It is demonstrated, by comparison of the results corrected for electron correlation with those obtained from relativistic intermediate coupling Dirac–Fock calculations, that both intravalence and core–valence correlation are important for achieving good agreement with experiment. However, for neutral strontium it is the intravalence correlation that seems to be more important whereas for the isoelectronic singly ionized yttrium the core-valence correlation, as represented by the core-polarization model, dominates. A delicate balance resulting from the partial collapse of the 4d orbital in Y+ may be a reason for the greater sensitivity to core polarization in this system.

ANDREA 2: Improved Version of Code for Reactor Core Analysis

2014 ◽  
Author(s):  
František Havlůj ◽  
Radim Vočka ◽  
Jiří Vysoudil
Keyword(s):  
Experimental Data ◽  
Cross Section ◽  
Reactor Core ◽  
Object Oriented ◽  
Core Analysis ◽  
Modular Architecture ◽  
The Core ◽  
Code Structure ◽  
Ongoing Development ◽  
Data Library

In recent years, the core physics code ANDREA has been significantly improved and its capabilities were vastly extended. The code implementation has been overhauled to more modern, object-oriented and modular architecture. The code structure was adapted to allow use of multiple different neutronics solvers and to tackle various spatial and energy discretization models. The data library formats and processing workflow have been completely generalized, and different transport codes (e.g. HELIOS, SERPENT or SCALE) can be used to prepare several-group cross-section libraries for ANDREA. The new version of the code has gone through extensive validation both on the benchmarks and experimental data. The modular architecture of ANDREA code allow for its ongoing development. Currently we focus on coupling of ANDREA code with TRANSURANUS code.

scholarly journals Fuel depletion calculation in MTR-LEU NUR reactor

2008 ◽  
Vol 23 (1) ◽  
pp. 11-18
Author(s):  
Foudil Zeggar ◽  
Amrane Halilou ◽  
Brahim Meftah ◽  
Mohamed Mokeddem ◽  
Fathi Letaim ◽  
...  
Keyword(s):  
Data Base ◽  
Fission Products ◽  
Reactor Fuel ◽  
Core Level ◽  
Plutonium Isotopes ◽  
The Core ◽  
Fuel Burn ◽  
Fuel Assemblies ◽  
Calculation Results ◽  
Good Agreement

In this article, we present the results of a few energy groups calculations for the NUR reactor fuel depletion analysis up to 45 000 MWd/tU taken as the maximum fuel burn up. The WIMSD-4 cell code has been employed as a calculation tool. In this study, we are interested in actinides such as the uranium and plutonium isotopes, as well as fission products Xe-135, Sm-149, Sm-151, Eu-155, and Gd-157. Calculation results regarding the five energy groups are in a good agreement with those obtained with only two energy groups which can, therefore, be used in all subsequent calculations. Calculation results presented in this article can be used as a microscopic data base for estimating the amount of radioactive sources randomly dispersed in the environment. They can also be used to monitor the fuel assemblies inventory at the core level.

scholarly journals Estimation of Fuel Burnup Rate for Core Conversion for the Nigeria Research Reactor-1 (NIRR-1) Fueled With 19.75% Enriched UO2 USING VENTURE PC Code

2020 ◽  
pp. 42-49
Author(s):  
J. A. Rabba ◽  
M. Y. Onimisi ◽  
D. O. Samson
Keyword(s):  
Uranium Dioxide ◽  
Research Reactor ◽  
Fuel Burnup ◽  
Reactor Operation ◽  
The Core ◽  
Computer Codes ◽  
Enriched Uranium ◽  
Conversion Study ◽  
Good Agreement

A standardized burnup analysis using VENTURE-PC computer codes system has been performed for the core conversion study of Nigeria Research Reactor-1. The result obtained from this analysis showed that the mass of Uranium decreases with increase in the number of days of reactor operation while the quantity of Plutonium continues to build up linearly. The buildup of the fissile isotope in the Low Enriched Uranium (LEU) core is very much greater than in the Highly Enriched Uranium (HEU) core. The quantity of Uranium-235 consumed and the amount of Plutonium-239 produce in the core of the reactor were 13.95 g and 0.766745 g respectively for the period of 11 years of reactor operation which is in good agreement with other literatures. This results obtained showed that uranium dioxide (UO2) fuel is a potential material for future Low Enriched Uranium (LEU) core conversion of Nigeria Research Reactor.

POLCA-T: Validation of Transient Bundle Flow Predictions During an All Recirculation Pumptrip in a BWR

2006 ◽  
Author(s):  
Ulf Bredolt
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
3D Model ◽  
Critical Power ◽  
Electrical Power ◽  
Measurement Data ◽  
Reactor Core ◽  
Power Company ◽  
3D Effects ◽  
Good Agreement

In this paper predicted results obtained with the POLCA-T code are compared to measurement data from an all recirculation pump trip test made at the Nuclear Power Plant Olkiluoto 1 during its commissioning in 1978. Olkiluoto 1 is one of three nuclear power plants owned by the electrical power company Teollisuuden Voima OY in Finland. The prediction is based on a best estimate reactor analysis code that incorporates a full 3D model of a reactor core into a system model. The use of a coupled code provides a means to simulate interactions between the reactor core and the plant dynamics to increase the understanding of different events and its impact on the core behaviour. The following conclusions are made based on the simulations with POLCA-T of the Olkiluoto 1 commissioning test of tripping all recirculation pumps: • POLCA-T predicts the event accurately, the flow coast down and the power decay is in good agreement with measured data. • 3D effects in channel flow can not be observed. The observed effects support the current use of a 1D approach since this is conservative with respect to power and the evaluation of Operating Limit Critical Power Ratio, (OLMCPR). This conclusion is also supported by measurement data.

Nuclear Analysis During Modification of NPP Core

2010 ◽  
Author(s):  
Andrius Slavickas
Keyword(s):  
Reactor Core ◽  
Reactor Power ◽  
Physical Parameters ◽  
Reactor Operation ◽  
Reactor Control ◽  
The Core ◽  
Burnable Absorber ◽  
Reactivity Coefficients ◽  
Control Rods ◽  
Neutron Activity

Reactor power and neutron activity control is the main key for safe reactor operation. Reactivity coefficients and effects are main measures to estimate reactor control and safety. These characteristics outline reactors behavior during usually exploitation and accident events. Reactivity coefficients and effects quantify the effect, which various parameters (e.g. fuel and graphite temperatures, amount of steam) have for the core neutron activity. Many modifications of RBMK-1500 reactor cores in Ignalina NPP were made during their lifetime. Reactor core modifications like load of higher enriched fuel with burnable absorber and new design control rods affected reactivity coefficients and effects. Neutron-physical parameters calculations of reactor core states with variant fuel loads and new design control rods were performed using QUABOC/CUBBOC-HYCA software. The changes of reactivity coefficients and effects were quantified in this paper.

scholarly journals Analysis Of Doppler Reactivity Coefficient On The Typical Pwr-1000 Reactor With Mox Fuel

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Rokh Madi
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Fuel Temperature ◽  
Safety Level ◽  
The Core ◽  
Mox Fuel ◽  
Calculation Results ◽  
Reactivity Coefficient ◽  
Fuel Elements

<p>Doppler coefficient is defined as a relation between fuel temperature changes and reactivity changes in the nuclear reactor core. Doppler reactivity coefficient needs to be known because of its relation to the safety of reactor operation. This study is intended to determine the safety level of the  typical PWR-1000 core by calculating the Doppler reactivity coefficient in the core with UO<sub>2</sub> and MOX fuels. The  typical PWR-1000 core  is similar to the PWR AP1000 core designed by Westinghouse but without Integrated Fuel Burnable Absorber (IFBA) and Pyrex. Inside the core, there are  UO<sub>2</sub> fuel elements with 3.40 % and 4.45 % enrichment, and MOX fuel elements with 0.2 % enrichment. By its own way, the presence of Plutonium in the MOX fuel will contribute to the change in core reactivity. The calculation was conducted using MCNPX code with the ENDF/B- VII nuclear data. The reactivity change was investigated at various temperatures. The calculation results show that the core reactivity coefficient of both UO<sub>2</sub> and MOX fuel are negative, so that the reactor is operated safely.</p>

Kinematics and Dynamics Simulation Research for Roller Coaster Multi-Body System

2011 ◽  
Vol 421 ◽  
pp. 276-280 ◽  
Author(s):  
Ge Ning Xu ◽  
Hu Jun Xin ◽  
Feng Yi Lu ◽  
Ming Liang Yang
Keyword(s):  
Structural Design ◽  
3D Model ◽  
Structure Design ◽  
Dynamics Simulation ◽  
Body System ◽  
Load Spectrum ◽  
Roller Coaster ◽  
Simulation Research ◽  
Calculation Results ◽  
Multi Body

To assess the roller coaster multi-body system security, it is need to extract the running process of kinematics, dynamics, load spectrum and other features, as basis dates of the roller coaster structural design. Based on Solidworks/motion software and in the 3D model, the calculation formula of the carrying car velocity and acceleration is derived, and the five risk points of the roller coaster track section are found by simulation in the running, and the simulation results of roller coaster axle mass center velocity are compared with theoretical calculation results, which error is less than 4.1%, indicating that the calculation and simulation have a good fit and providing the evidence for the roller coaster structure design analysis.

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