scholarly journals Evaluation of Pipe Whip Impacts on Neighboring Piping and Walls of the Ignalina Nuclear Power Plant

2002 ◽  
Author(s):  
Gintautas Dundulis ◽  
Ronald F. Kulak ◽  
Algirdas Marchertas ◽  
Evaldas Narvydas ◽  
Mark C. Petry ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Transient Analysis ◽  
Cooling System ◽  
Nonlinear Material ◽  
Reactor Safety ◽  
Concrete Wall ◽  
Concrete Walls ◽  
Reinforced Concrete Wall ◽  
Core Cooling ◽  
Group Distribution

Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. Therefore, two cases are investigated: • GDH impact on an adjacent GDH and its attached piping; • GDH impact on an adjacent reinforced concrete wall. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.

scholarly journals Study on Radiation Shielding Performance of Reinforced Concrete Wall After the Earthquake

2010 ◽  
Vol 5 (4) ◽  
pp. 361-368
Author(s):  
Keiji Sekine ◽  
◽  
Yoshinari Munakata ◽  
Osamu Kontani ◽  
Koji Oishi ◽  
...  
Keyword(s):  
Reinforced Concrete ◽  
Nuclear Power ◽  
Radiation Shielding ◽  
Crack Model ◽  
Concrete Wall ◽  
Concrete Walls ◽  
Reinforced Concrete Walls ◽  
Earthquake Resistant ◽  
Reinforced Concrete Wall ◽  
Large Earthquakes

The structure in which radioactive substances are stored and handled must be earthquake resistant. We will be confirming radioactive shielding performance of reinforced concrete walls when cracks occur due to large earthquakes. In this study, we performed horizontal loading experiments to evaluate shielding performance of earthquake resisting walls and constructed a safety crack model. Next, the shielding calculation was done by using the crack model, and the shielding performance of the earthquake resisting wall was evaluated. As a result, if the structure is designed according to the standards outline for nuclear power related facility, and an earthquake causes cracks in an earthquake resisting wall, it was shown that if the thickness of the earthquake resisting wall was less than 80 cm, the decrease in shielding performance was very small, and that the radiation exposure on the general public and the employee was negligible.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Hybrid heat pipe based passive in-core cooling system for advanced nuclear power plant

2015 ◽  
Vol 90 ◽  
pp. 609-618 ◽  
Author(s):  
Yeong Shin Jeong ◽  
Kyung Mo Kim ◽  
In Guk Kim ◽  
In Cheol Bang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Pipe ◽  
Nuclear Power ◽  
Cooling System ◽  
Core Cooling

scholarly journals Soft Missile Impact on Shear Reinforced Concrete Wall

2010 ◽  
Vol 5 (4) ◽  
pp. 426-436 ◽  
Author(s):  
Arja Saarenheimo ◽  
◽  
Kim Calonius ◽  
Markku Tuomala ◽  
Ilkka Hakola ◽  
...  
Keyword(s):  
Reinforced Concrete ◽  
Transverse Shear ◽  
Shear Failure ◽  
Three Dimensional ◽  
Concrete Wall ◽  
Concrete Walls ◽  
Shell Elements ◽  
Bending Mode ◽  
Reinforced Concrete Wall ◽  
Wall Deformation

In developing numerical approaches for predicting the response of reinforced concrete structures impacted on by deformable projectiles, we predict structural behavior collapse and damage using simple analysis and extensive nonlinear finite element (FE)models. To verify their accuracy, we compared numerical results to experimental data and observations on impact-loaded concrete walls with bending and transverse shear reinforcement. Different models prove adequate for different cases and are sensitive to different variables, making it important to rely on more than a single model alone. For wall deformation in bending mode, deflection is predicted reasonably well by simple four-node shell elements. Where punching dominates, transverse shear behavior must be considered. Formation of a shear failure cone is modeled using three-dimensional solid elements.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Analysis of Sound Insulation Performance of Reinforce Concrete Walls between Households According to Wall Thickness Criteria in Apartments

2017 ◽  
Vol 873 ◽  
pp. 237-242
Author(s):  
Hye Kyung Shin ◽  
Kyoung Woo Kim ◽  
A Yeong Jeong ◽  
Kwan Seop Yang
Keyword(s):  
Reinforced Concrete ◽  
Wall Thickness ◽  
Field Measurements ◽  
Reinforce Concrete ◽  
Sound Insulation ◽  
Concrete Wall ◽  
Minimum Thickness ◽  
Concrete Walls ◽  
Reinforced Concrete Wall ◽  
Insulation Performance

Sound insulation between households is properly ensured to provide a quiet residential environment in apartments. The legal requirements for sound insulation in apartments in Korea are set to meet the wall’s minimum thickness or sound insulation performance. When construction companies choose the walls that satisfy thethickness in the standards of boundary walls between households, it is difficult to know the sound insulation performance. In this study, the sound insulation performance of reinforced concrete walls is predicted according to the wall thickness criteria and analyzed through field measurements. In newly built apartments, the reinforced concrete wall’s sound insulation performance(R'w) is 56 – 66 dB, which is a similar level of the international criterion. And the sound insulation performance of the reinforced concrete wall according to thickness standards is similar to sound insulation performance standardsof Korea.

scholarly journals Floors number influence on the instability parameter of reinforced concrete wall- or core-braced buildings

2013 ◽  
Vol 6 (5) ◽  
pp. 783-796 ◽  
Author(s):  
R. J. Ellwanger
Keyword(s):  
Reinforced Concrete ◽  
Analytical Study ◽  
Concrete Structures ◽  
Concrete Wall ◽  
Concrete Walls ◽  
Order Analysis ◽  
Reinforced Concrete Walls ◽  
Continuous Models ◽  
Reinforced Concrete Wall ◽  
Instability Parameter

This work aims to investigate the floors number influence on the instability parameter limit α1 of buildings braced by reinforced concrete walls and/or cores. Initially, it is showed how the Beck and König discrete and continuous models are utilized in order to define when a second order analysis is needed. The treatment given to this subject by the Brazilian code for concrete structures design (NBR 6118) is also presented. It follows a detailed analytical study that led to the derivation of equations for the limit α1 as functions of the floors number; a series of examples is presented to check their accuracy. Results are analyzed, showing the precision degree achieved and topics for continuity of research in this field are indicated.

Optimizing Maintenance Strategy of a Reactor Pressure Vessel Using 3D-CFD and FEM Based Probabilistic Pressurized Thermal Shock Analysis

2017 ◽  
Author(s):  
Xiaoyong Ruan ◽  
Toshiki Nakasuji ◽  
Kazunori Morishita
Keyword(s):  
Spatial Distribution ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Cooling System ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Core Cooling ◽  
Reactor Pressure

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

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