scholarly journals Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

2011 ◽  
Vol 2011 ◽  
pp. 1-16
Author(s):  
Andrew M. Ward ◽  
Benjamin S. Collins ◽  
Marcelo Madariaga ◽  
Yunlin Xu ◽  
Thomas J. Downar
Keyword(s):  
Steady State ◽  
Cross Sections ◽  
Nuclear Power ◽  
Model Development ◽  
Lattice Models ◽  
Transient Behavior ◽  
Injection System ◽  
The Core ◽  
Control Rods ◽  
Good Agreement

In order to analyze the steady state and transient behavior of CNA-II, several tasks were required. Methods and models were developed in several areas. HELIOS lattice models were developed and benchmarked against WIMS/MCNP5 results generated by NA-SA. Cross-sections for the coupled RELAP5/PARCS calculation were extracted from HELIOS within the GenPMAXS framework. The validation of both HELIOS and PARCS was performed primarily by comparisons to WIMS/PUMA and MCNP for idealized models. Special methods were developed to model the control rods and boron injection systems of CNA-II. The insertion of the rods is oblique, and a special routine was added to PARCS to treat this effect. CFD results combined with specialized mapping routines were used to model the boron injection system. In all cases there was good agreement in the results which provided confidence in the neutronics methods and modeling. A coupled code benchmark between U of M and U of Pisa is ongoing and results are still preliminary. Under a LOCA transient, the best estimate behavior of the core appears to be acceptable.

Safety Study of Primary Loop System of Civil Marine Nuclear Power Plant

2014 ◽  
Vol 541-542 ◽  
pp. 916-921 ◽  
Author(s):  
Li Xu ◽  
Ru Chao Deng ◽  
Chu Xu ◽  
Di Zhang ◽  
Chen Xing Sheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Comprehensive Evaluation ◽  
Environmental Parameters ◽  
Loop System ◽  
Injection System ◽  
Safety Study ◽  
Primary Loop ◽  
Control Rods

For evaluate the risk of civil marine nuclear power plant, through the searching related standards for ship, external environmental parameters that the nuclear ship should be suited was found. Based on the characteristics of power plant of civil nuclear-powered ship, the hierarchy system of primary loop system was established and corresponding indicator marking criteria were formulated for the risk assessment. The result shows that the Reactor Safety Injection System (RIS), the Reactor Boron and the Water Supply System (REA), the Control Rods and the Hull of Fuel Canning are the key risk factors in the primary loop system. Finally, the comprehensive evaluation was carried out for collision, stranding and swing of multi-degree of freedom, and put forward relative countermeasures to cope with the possible risks based on the comprehensive evaluation and combined with the literatures.

scholarly journals Validation of Doppler Temperature Coefficients and Component Power Distribution for the Advanced Neutronics Component Program KYLIN V2.0

2021 ◽  
Vol 9 ◽  
Author(s):  
Lei Jichong ◽  
Xie Jinsen ◽  
Chen Zhenping ◽  
Yu Tao ◽  
Yang Chao ◽  
...  
Keyword(s):  
Engineering Design ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Engineering ◽  
Temperature Coefficients ◽  
The Core ◽  
Fuel Assemblies ◽  
Design Requirements ◽  
Design Software ◽  
Good Agreement

This work is interested in verifying and analyzing the advanced neutronics assembly program KYLIN V2.0. Assembly calculations are an integral part of the two-step calculation for core design, and their accuracy directly affects the results of the core physics calculations. In this paper, we use the Doppler coefficient numerical benchmark problem and CPR1000 AFA-3G fuel assemblies to verify and analyze the advanced neutronics assembly program KYLIN V2.0 developed by the Nuclear Power Institute of China. The analysis results show that the Doppler coefficients calculated by KYLIN V2.0 are in good agreement with the results of other well-known nuclear engineering design software in the world; the power distributions of AFA-3G fuel assemblies are in good agreement with the results of the RMC calculations, it’s error distribution is in accordance with the normal distribution. It shows that KYLIN V2.0 has high calculation accuracy and meets the engineering design requirements.

Coupling of Sloshing and Ship Motions

2003 ◽  
Vol 47 (03) ◽  
pp. 208-221 ◽  
Author(s):  
Olav F. Rognebakke ◽  
Odd M. Faltinsen
Keyword(s):  
Steady State ◽  
Transient Behavior ◽  
Frequency Domain Analysis ◽  
Ship Motions ◽  
Steady State Motion ◽  
Constant Coefficients ◽  
Incident Waves ◽  
Violent Sloshing ◽  
Different Levels ◽  
Good Agreement

The coupled effect between ship motions and sloshing is studied. Two-dimensional experiments of a hull section containing tanks filled with different levels of water excited in sway by regular waves have been conducted. Steady-state results are obtained for the sway amplitude. Even if violent sloshing occurs in the tanks, the steady-state motion is almost linear and sinusoidal with the frequency of the linear incident waves. This implies that higher-order harmonics of the sloshing force are filtered out by the system. Simulations of the modeled case are performed using a linear and a nonlinear sloshing model and mainly assuming linear external flow. For steady-state motion, a convolution formulation does not improve the results relative to using constant coefficients in the equation of motion. However, in order to properly model the transient behavior in an irregular sea, a convolution formulation must be included. The treatment of the retardation function for the external problem is discussed in detail. A good agreement between experiments and computations is reported. The calculated coupled motion is sensitive to the damping of the sloshing motion in a certain frequency range where the coupled sloshing and ship motions cause resonant ship motions. A quasilinear frequency domain analysis is used to explain this by introducing the sloshing loads as a frequency dependent spring.

Main Steam Line Break Calculations Using a Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor

2008 ◽  
Author(s):  
Mathias Sta˚lek ◽  
Jo´zsef Ba´na´ti ◽  
Christophe Demazie`re
Keyword(s):  
Steady State ◽  
Fuel Assembly ◽  
Thermal Power ◽  
Steam Line ◽  
Pressurized Water ◽  
The Core ◽  
Different Types ◽  
Main Steam Line ◽  
Main Steam ◽  
Control Rods

A Main Steam Line Break (MSLB) is an important transient for Pressurized Water Reactors (PWR) due to the strong positive reactivity introduced by the over-cooling of the core. Since this effect is stronger when the Moderator Temperature Coefficient (MTC) has a large amplitude, a conservative result will be obtained for a high burnup of the fuel due to the more negative MTC late in the cycle. The calculations have been performed at a cycle burnup of 12.9742 GWd/tHM. The Swedish Ringhals-3 PWR is a three loop Westinghouse design, currently with a thermal power of 3000 MW. The PARCS model has 157 fuel assemblies of 8 different types. Four different types of reflector are used. The cross sections, and kinetic data were obtained from CASMO-4 calculations, using a cross section interface developed at the department. There are 24 axial nodes, and 2×2 radial nodes for each assembly. The transient option for calculating the effect of poisoning was used. The PARCS model has been validated against steady-state measurements from Ringhals-3 of the Relative Power Fraction (RPF) and of the core criticality. The RELAP5 model has 157 channels for the core which means that there is a one to one correspondence between the thermal hydraulics model and the neutronics model. There is eight axial nodes. Originally, the intention was to have 24 axial nodes but this proved not to work because of some limitation in RELAP5. There is currently no mixing between the different channels in the core. The feedwater, and turbines are modelled as boundary conditions. The stand-alone RELAP5 model has been validated against steady state measurements from Ringhals-3. A number of different cases were considered. In the first case, both the isolation of the feedwater for the broken loop, and all the control rods were assumed to work properly. For the second case one of the control rods was assumed to be stuck. The stuck rod was located in the fuel assembly with the highest power. This rod has also one of the highest rod worths. In the final case, the feedwater control valve for the broken loop was fully open. None of the cases led to any recriticality. The increase in power for each fuel assembly was also investigated. With the control rod located in the assembly with the highest power, the maximum power increase before scram turned out to be about 25% compared to the initial power.

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injector: Part 14—Cooling Characteristics of the Particulate Core Material Debris Accumulated in Horizontal Narrow Gap

2006 ◽  
Author(s):  
Y. Hirao ◽  
G. Su ◽  
K. Sugiyama ◽  
T. Narabayashi ◽  
M. Mori ◽  
...  
Keyword(s):  
Heat Supply ◽  
Nuclear Power ◽  
High Efficiency ◽  
Hot Spot ◽  
Research Work ◽  
Lower Surface ◽  
Injection System ◽  
Core Material ◽  
Metal Debris ◽  
The Core

When LOCA occurs in proposed nuclear reactor systems, the coolability of the core would be kept by the SI core injection system and therefore the probability of the core meltdown is negligible small. In this research work, we make it clear that the coolability of the RPV bottom is secured even if a part of the core should melt and a substantial amount of debris should be deposited on the lower plenum. In this report, we examined experimentally the coolability of the RPV bottom that a Zircaloy-based loose debris layer is deposited on. We set up a heat supply section made by SUS304 on the loose debris layer and measured the heat flux released into the loose debris bed and the temperature at the lower surface of the heat supply section. In addition, we measured the temperature distribution at the bottom of the loose debris bed. It became clear in this study that the coolability depends on the amount of coolant supplied, and the hot spot would not occur when coolant is supplied. Even if a hotspot should occur in the oxidization of loose metal debris accompanied with rapid heat generation. It is found that when a small amount of coolant can be supplied, it disappears because of a high capillary force of oxidized loose debris. So it is confirmed that the soundness of RPV is basically maintained.

scholarly journals Safety Analysis of Molten Salt Fast Reactor Fuel Composition using Moltres

2019 ◽  
Author(s):  
Sun Myung Park ◽  
Andrei Rykhlevskii ◽  
Kathryn Huff
Keyword(s):  
Steady State ◽  
Molten Salt ◽  
Core Temperature ◽  
Fast Reactor ◽  
Transient Behavior ◽  
Total Power ◽  
Fuel Composition ◽  
Sources Of Error ◽  
Start Up ◽  
Good Agreement

The Molten Salt Fast Reactor (MSFR) has garnered much interest for its inherent safety and sustainbility features. The MSFR can adopt a closed thorium fuel cycle for sustainable operation through the breeding of 233 U from 232 Th. The fuel composition changes significantly over the course of its lifespan. In this study, we investigated the steady state and transient behavior of the MSFR using Moltres, a coupled neutronics/thermal-hydraulics code developed within the Multiphysics Object Oriented Simulation Environment (MOOSE) framework. Three different fuel compositions, start-up, early-life, and equilibrium, were examined for potentially dangerous core temperature excursions during a unprotected loss of heat sink (ULOHS) accident. The six-group and total neutron flux distributions showed good agreement with SERPENT and published MSFR results, while the temperature distribution and total power showed discrepancies which can be attributed toknown sources of error. For the transient behavior under the ULOHS scenario, while the transition time towards the new steady state core temperature is also in good agreement with existing MSFR simulations by Fiorina et al., Moltres under-estimated the temperature rise by a factor of ten, due to the same sources of error affecting the steady state results. While an MSFR loaded with start-up fuel composition operates at a higher temperature than with the other two fuel compositions, all three cases were shown to be inherently safe due to thestrong negative temperature feedback.

Analysis of Primary Side Feed and Bleed Operation at Armenian NPP Unit 2

2002 ◽  
Author(s):  
Armen Amirjanyan ◽  
Tsolak Malakyan ◽  
Jae Jo
Keyword(s):  
Steady State ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Saturation Temperature ◽  
Steam Generators ◽  
The Core ◽  
Supply Tank ◽  
Primary Water ◽  
Diesel Generators ◽  
Core Cooling

Analyses were performed for the primary side feed and bleed (F&B) procedures for Armenian Nuclear Power Plant (ANPP), using the RELAP5 code. ANPP is a six-loop WWER-440/270 model of Russian design. The purpose of the analyses was to demonstrate the core cooling capability of the plant using the high pressure injection (HPI) system (feed) and pressurizer safety valves (bleed), and to evaluate temperature changes in the Borated Water Supply Tank (BWST). The Mod3.2.2β version of RELAP5 was used for the analyses. The six loops of the plant were modeled by 4 loops; three loops represented three individual loops and one loop combined the other three loops. The input described the primary and secondary loops in detail, as well as all safety systems and most of control systems. The HPI system was connected to the cold leg of each loop. The Emergency Core Cooling (ECC) mixer component was applied in the connection of the HPI system to simulate the phenomena associated with the subcooled ECC injection into a reactor coolant system. Nominal operating parameters were used to establish the steady state conditions. Some of the initiating events and boundary conditions were based on the experience and information obtained from previous transients and incidents in the plant. The incident was assumed to be initiated by a loss of off-site power supply and simultaneous reactor scram. Diesel generators were assumed not to be available for 11,000 seconds. Thus, feedwater to the steam generators, main coolant pumps, and HPI pumps were not available during this period and the core was cooled by natural circulation in the primary loops supported by the remaining water in the steam generators. During this period the primary water temperature and pressure were near or slightly higher than the steady state values, but the water remained subcooled. At 11,000 seconds, upon starting of the diesel generators, the operator switched on two HPI pumps and opened one pressurizer safety valve. At this time, all six steam generators were found to be nearly empty. Once the F&B operation began, the pressure and temperature of the primary loops declined rapidly. The results of F&B analyses show that there is no violation of the core coolability conditions when two HPI pumps are available, and the temperature of BWST tank does not reach the saturation temperature when two heat exchanger pumps are available.

Experience of Xenon Oscillation During an Initial Physics Test of UCN Unit 5 in Korea

2010 ◽  
Author(s):  
Yong Rae Kim ◽  
Tae Young Choi ◽  
Sun Ho Shin ◽  
Ki Bong Seong
Keyword(s):  
Steady State ◽  
Annual Cycle ◽  
Stability Index ◽  
Control Rod ◽  
The Core ◽  
Axial Stability ◽  
Xenon Oscillation ◽  
Burnable Absorber ◽  
Control Rods ◽  
Initial Cycle

Initial core of Ulchin Nuclear Unit 3 (UCN3), which is one of earlier OPR1000 model, was 4 batches and designed as annual cycle after second cycle. The utility requested that UCN Unit 5 (UCN5), which is another of OPR1000 model, had capability of a longer cycle operation from second cycle. Therefore, KNF modified the number of batches from 4 to 3 for OPR1000 initial core, as well as, the number of burnable absorber, and the cutback length of the absorber. However, due to these changes, Xenon oscillation was slightly increased at 100% power during the physics test of UCN5, while that oscillation at 100% power in UCN3 had been gone down without any control rod motion. The xenon oscillation direction is related to axial stability index. The index of UCN3 increased from a slightly negative at BOC to positive at EOC, the index of UCN5 was positive even at BOC, which meant that the core does not go to be stable without the control rod motion. The core of UCN5 became the steady state by the insertion of control rods into the core. To meet the physics test condition, the oscillation was controlled by control rods immediately. After the happening, KNF optimized the cutback length of burnable absorber rods and applied to APR1400, which will keep being stable in xenon oscillation during physics test at the initial cycle.

Validation DualSPHysics Code for Liquid Sloshing Phenomena

2014 ◽  
Author(s):  
Sergei K. Buruchenko ◽  
Alejandro J. C. Crespo
Keyword(s):  
Nuclear Power ◽  
Liquid Sloshing ◽  
Accident Analysis ◽  
The Core ◽  
Particle Hydrodynamics ◽  
In Series ◽  
Core Meltdown ◽  
Smoothed Particle ◽  
Smoothed Particle Hydrodynamics Method ◽  
Good Agreement

The DualSPHysics code is proposed as a numerical tool for the simulation of liquid sloshing phenomena. A particular type of sloshing motion can occur during the core meltdown of a liquid metal cooled reactor (LMR) and can lead to a compaction of the fuel in the center of the core possibly resulting in energetic nuclear power excursions. This phenomenon was studied in series of “centralized sloshing” experiments with a central water column collapsing inside the surrounding cylindrical tank. These experiments provide data for a benchmark exercise for accident analysis codes. To simulate “centralized sloshing” phenomena, a numerical method should be capable to predict the motion of the free surface of a liquid, wave propagation and reflection from the walls. The DualSPHysics code based on the smoothed particle hydrodynamics method was applied to the simulation of “centralized sloshing” experiments. Simulation results are compared with the experimental results. In a series of numerical calculations it is shown that overall motion of the liquid is in a good agreement with experimental observations. Dependence on the initial and geometrical symmetry is studied and compared with experimental data.

scholarly journals Application of Electrical Impedance Measurements’ Method for Studies of Bubble Flows

2021 ◽  
Vol 12 (3) ◽  
pp. 183-193
Author(s):  
I. A. Konovalov ◽  
A. A. Chesnokov ◽  
A. A. Barinov ◽  
S. M. Dmitriev ◽  
A. E. Khrobostov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Electrical Impedance ◽  
Wire Mesh ◽  
Measuring System ◽  
Gas Content ◽  
The Core ◽  
Multicomponent Flows ◽  
Liquid Flows ◽  
Good Agreement

One of the important tasks in carrying out a computational justification of the reliability and safety of equipment that is part of the projected nuclear power plants today is the modeling of the bubbly regime of the coolant flow. In this regard the aim of this work is the use of extended methods of using matrix conductometric systems which are widespread in research practice for study of gas-liquid flows.The work uses a method of primary processing of experimental data aimed at eliminating of excess conductivity in the cells of the developed wire mesh sensor which makes it possible to obtain the values of the true volumetric gas content in the investigated area.Subsequent analysis of the possibilities to estimate the volumes of registered gas bubbles by the gradient method as well as the size of the interface in the sensor cells which plays a key role in modeling the interfacial heat and mass transfer.Comparison of readings values with the control instruments cues showed a good agreement. The presented work is an adaptation of the use of a conductometric measuring system for the study of multicomponent flows with the aim of further application for the study of two-component flows in the channels of the core simulator using wire mesh sensors.

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