Point Kinetics LOCA Scenarios Assessment at the Using of Hafnium Absorber in Advanced CANDU Fuel Designs

2009 ◽  
Author(s):  
Iosif Prodea ◽  
Cristina Alice Ma˘rgeanu ◽  
Ilie Prisecaru ◽  
Nicolae Da˘nila˘
Keyword(s):  
Power Reactor ◽  
Data Bank ◽  
Safety Analysis ◽  
Core Model ◽  
Major Drawback ◽  
Negative Cell ◽  
Burnable Absorber ◽  
Point Kinetics ◽  
Code Version ◽  
Full Core

The goal of the paper is to compare the results obtained through point kinetics calculations at the simulation of several CANDU® LOCA scenarios in the case of using Hafnium as burnable absorber in advanced CANDU fuel designs (ACR™-1000 based) along with standard CANDU NU fuel. The 3D DIREN_MG diffusion code developed in INR Pitesti was used to accomplish this task after upgrading by original procedures, which allow for Neutronics and ThermalHydraulics coupled calculations. The intervention of SDS1 was modeled until the transient is terminate by SDS1 action. The CANDU community has always interested in the void reactivity reducing-the major drawback of this power reactor. The paper’s novelty elements consist in the using of a newer WIMS code version (WIMSDB5 from NEA Data Bank) and an updated IAEA WIMS library based on newly ENDF/B-VII version along with a simplified DIREN_MG ACR-1000 core model. The results emphasized the critical Hafnium shell thicknesses which allow for a slightly negative cell Void Reactivity (VR) target and the direct consequences in the full core safety analysis.

Development and Application of Multi-Physics Safety Analysis Code for Advanced Liquid Metal Cooled Reactor

2018 ◽  
Author(s):  
Chi Wang ◽  
Xuebei Zhang ◽  
Jingchao Feng ◽  
Muhammad Shehzad Khan ◽  
Minyou Ye ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Coolant Temperature ◽  
Medium Size ◽  
High Tech ◽  
Fuel Temperature ◽  
Flow Phenomena ◽  
Nuclear Reactor Safety ◽  
Cfd Method ◽  
Point Kinetics

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

Identification of Novel Human Serum Albumin (SA) Inhibitors from Scoparia Dulsis for Urolithiasis

2020 ◽  
Vol 16 (3) ◽  
pp. 308-317 ◽  
Author(s):  
Divya Shaji
Keyword(s):  
Binding Energy ◽  
Human Serum Albumin ◽  
Serum Albumin ◽  
Human Serum ◽  
3D Structure ◽  
Data Bank ◽  
Major Drawback ◽  
Structure Based Drug Design ◽  
Lipinski’S Rule ◽  
Extracorporeal Shock Wave

Background:: Urolithiasis is the process of forming stones in the kidney, bladder, and/or urinary tract. It has been reported that kidney stones are the third most common disorder among urinary diseases. At present, surgical procedures and Extracorporeal Shock Wave Lithotripsy (ESWL) are commonly employed for the treatment of Urolithiasis. The major drawback of these procedures is the recurrence of stones. Methods: This study aimed to identify potential natural inhibitors against human Serum Albumin (SA) from the plant Scoparia Dulsis for Urolithiasis. As protein-ligand interactions play a key role in structure- based drug design, this study screened 26 compounds from Scoparia Dulsis and investigated their binding affinity against SA by using molecular docking. The three dimensional (3D) structure of SA was retrieved from Protein Data Bank (PDB) and docked with PubChem structures of 26 compounds using PyRX docking tool through Autodock Vina. Moreover, a 3D similarity search on the PubChem database was performed to find the analogs of best scored compound and docking studies were performed. Drug-likeness studies were made using Swiss ADME and Lipinski’s rule of five was performed for the compounds to evaluate their anti-urolithiatic activity. Results: The results showed that citrusin c (Eugenyl beta-D-glucopyranoside) exhibited best binding energy of -8.1 kcal/mol with SA followed by aphidicolin, apigenin, luteolin and scutellarein. Two compounds (PubChem CID 46186820, PubChem CID 21579141) analogous to citrusin c were selected based on the lowest binding energy. Conclusion: This study, therefore, reveals that these compounds could be promising candidates for further evaluation for Urolithiasis prevention or management.

A 3D BWR Core Model Within the GRS Analysis Simulator ATLAS

2006 ◽  
Author(s):  
K. Velkov ◽  
S. Langenbuch ◽  
G. Lerchl ◽  
W. Pointner
Keyword(s):  
Three Dimensional ◽  
Core Model ◽  
Data Generation ◽  
Generation System ◽  
Water Reactor ◽  
Neutron Kinetics ◽  
On Line ◽  
Using Data ◽  
Point Kinetics ◽  
Good Agreement

The paper describes the application of the ATLAS simulator environment and the coupled three dimensional (3D) neutron-kinetics and thermal-hydraulics system code QUABOX/CUBBOX-ATHLET for a boiling water reactor (BWR) plant transient. A turbine trip (TT) transient is simulated and analyzed once with the 3D core model and once with the point kinetics (PK) model using data generated on the basis of the 3D calculations by the kinetics data generation system SIGMAS developed in GRS. The comparison shows a very good agreement, which is an important precondition for performing transient analyses with an on-line switch from PK to 3D calculation for a BWR plant transient within the ATLAS simulator.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Core Model ◽  
Primary System ◽  
Control Rod ◽  
Transient Behaviour ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients ◽  
Core Conditions ◽  
Point Kinetics

Abstract In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

Generation of Consistent Data for Neutron Kinetics Models of the GRS System Code ATHLET

2004 ◽  
Author(s):  
S. Langenbuch ◽  
K. Velkov
Keyword(s):  
Core Model ◽  
Data Generation ◽  
Kinetics Models ◽  
Neutron Kinetics ◽  
Basic Characteristics ◽  
Point Kinetics

The paper describes the first experience at GRS with a switch algorithm built into the system code ATHLET, which allows to turn to point kinetics or 3D calculations with the neutronics core model QUABOX/CUBBOX and vica versa. The heart of the algorithm is the neutronics data generation code SIGMAS, developed and validated at GRS. Its basic characteristics and possibilities of applications are briefly described. As a demonstration of the algorithm, the results of two boron transient calculations performed with the switch coupling are presented and discussed.

A Study on Doppler Weighting Factor for Control Element Assembly Ejection Accident by Using Newly Developed Nuclear Design Code and Non-LOCA Methodology

2013 ◽  
Author(s):  
Kyungmin Yoon ◽  
Chansu Jang ◽  
Jooil Yoon
Keyword(s):  
Safety Analysis ◽  
Weighting Factor ◽  
Time Effect ◽  
Kinetics Model ◽  
Control Element ◽  
Fuel Temperature ◽  
Design Code ◽  
The Core ◽  
Point Kinetics ◽  
Reactivity Insertion

Among Reactivity Initiated Accidents (RIAs) for Pressurized Water Reactor (PWR), Control Element Assembly Ejection (CEAE) accident causes the rapid positive reactivity insertion to the core. It causes an asymmetric power distortion which results in the rising of local fuel temperature, fuel pellet thermal expansion and cladding ballooning or rupture. In the CEAE accident, Doppler feedback has a profound effect because the negative reactivity insertion due to the rise of fuel temperature reduces the core power after rapid power excursion. But the Doppler reactivity can’t be calculated properly in the safety analysis code, using point kinetics model, because the point kinetics model is not able to consider spatial-time effect of the sudden rise in local fuel temperature on Doppler feedback calculation during CEAE accident. And then the excessively high core power which results from the underestimated Doppler feedback would make more severe results such as PCMI fuel failure, fuel cladding rupture and serious DNB fuel failure. Therefore, Doppler Weighting Factor (DWF) is needed for the safety analysis of CEAE accident to compensate a missing spatial-time effect on Doppler feedback calculation. In this study, the adequacy of the application of DWF for APR1400 was evaluated by using nuclear design code called ASTRA (Advanced Static and Transient Reactor Analyzer)[1] and a methodology called ISAM (Integrated Safety Analysis Methodology)[2]. ASTRA is the 3D nuclear design code newly developed by KNF and has various functions such as the static core design, the transient core analysis and the operational support. ISAM is the methodology which is newly developed by KNF to perform the Non-LOCA safety analysis by using RETRAN[3] code which is widely used in the transient analysis and based on the point kinetics model.

Preliminary Study of Safety Analysis of Pb-Bi Cooled Small Power Reactor with Natural Circulation

2013 ◽  
Vol 772 ◽  
pp. 519-523
Author(s):  
Novi Trian ◽  
Abdul Waris ◽  
Sparisoma Viridi ◽  
Su'ud Zaki
Keyword(s):  
Flow Rate ◽  
Power Reactor ◽  
Natural Circulation ◽  
Safety Analysis ◽  
Coolant Flow ◽  
Coolant Flow Rate ◽  
Outlet Temperature ◽  
Small Power ◽  
Optimum Height ◽  
Lead Bismuth Eutectic

We report our study of the safety analysis in design of small power reactor which design based on the concept of a long-life core reactor cooled by lead-bismuth eutectic (LBE). The motivation of these studies is in order to design a next generation of reactors, we need to design a type of reactor that has inherent safety. We designed the small Pb-Bi cooled reactor with MOX-Nitrate as a fuel. In order to study the safety analysis of this reactor we conducted studies of chimney length effect to coolant flow rate in natural circulation and dependency of outlet temperature with coolant flow rate. From this work we obtained the optimum height of chimney at 15 m for the lead-bismuth eutectic flow rate 3500 kg/s and also we found the dependency of outlet temperature with lead-bismuth eutectic flow rate.

scholarly journals APPLICATION OF THE SPH EQUIVALENCE TECHNIQUE TO THE CABRI FULL CORE IN NON-FUNDAMENTAL MODE

2021 ◽  
Vol 247 ◽  
pp. 03019
Author(s):  
Alain Hébert ◽  
Julien Taforeau ◽  
Jean-Jacques Ingremeau
Keyword(s):  
Fundamental Mode ◽  
Method Of Characteristics ◽  
Core Model ◽  
Mean Square ◽  
Gradient Search ◽  
The Core ◽  
Limited Memory ◽  
Rms Error ◽  
Core Condition ◽  
Full Core

We developed a SPH equivalence technique in non-fundamental mode condition between a CABRI full-core model solved with the method of characteristics (MOC) in 2D and a simplified full-core model solved with the simplified P3 (SP3) method, linear anisotropic sources and discretized with Raviart-Thomas finite elements over a pure Cartesian mesh. The MOC and SP3 calculations are performed with DRAGON5 and DONJON5 codes, respectively. A three-parameter database is generated by DRAGON5 and is interpolated in DONJON5 as a function of the core condition. An objective function is set as the root mean square (RMS) error (MOC-SP3 discrepancy) on absorption distribution and leakage rates defined over the macro-geometry in DONJON5. Our algorithm is a quasi-Newtonian gradient search based on the Limited memory Broyden-Fletcher-Goldfarb-Shanno (LBFGS) method. Numerical results are presented with Hafnium bars withdrawn or inserted.

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