scholarly journals Study of Fast Transient Pressure Drop in VVER-1000 Nuclear Reactor Using Acoustic Phenomenon

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Soroush Heidari Sangestani ◽  
Mohammad Rahgoshay ◽  
Naser Vosoughi ◽  
Mitra Athari Allaf
Keyword(s):  
Pressure Drop ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Computer Code ◽  
Accurate Method ◽  
Transient Simulation ◽  
Transient Pressure ◽  
Pressurized Water ◽  
Thermal Feedback ◽  
Acoustic Phenomenon

This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR). Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.

scholarly journals Gas-cooled nuclear reactor core shaping using heat exchange intensifiers

2019 ◽  
Vol 5 (1) ◽  
pp. 75-80
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Heat Exchange ◽  
Pressure Drop ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Temperature Fields ◽  
Ansys Fluent ◽  
Safety Requirements ◽  
Cooling Channels ◽  
The Core ◽  
Nuclear Reactor Core

The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping. The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed. The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields. The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels. Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop. The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

scholarly journals ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

2016 ◽  
Vol 18 (3) ◽  
pp. 127 ◽  
Author(s):  
Setiyanto Setiyanto ◽  
Tukiran Surbakti
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Empty Space ◽  
Computer Code ◽  
Safety Analysis ◽  
Design Calculation ◽  
Triga Reactor ◽  
Calculation Results ◽  
Fuel Elements ◽  
Irradiation Facilities

ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g), but very low value for Lazy Susan position (lest then 0,11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung, telah dilakukan kajian penggunaan bahan bakar jenis pelat seperti yang digunakan oleh RSG-GAS. Berbagai langkah analisis telah disiapkan, termasuk perhitungan desain teras, dan sistem keselamatannya. Penggunaan elemen bakar tipe pelat menghasilkkan reaktor dapat dioperasikan hanya dengan 20 elemen bakar. Dibandingkan teras aslinya, nampak bahwa teras baru menjadi lebih kecil dan kompak, rapat dayanya naik, tetapi menyisakan beberapa ruang kosong yang dimungkinkan untuk menempatkan fasilitas iradiasi di teras. Dengan adanya fasilitas iradiasi di dalam teras, maka pembangkitan panas gamma di teras menjadi faktor baru yang harus diperhatikan. Untuk alasan ini, telah dilakukan perhitungan pembangkitan panas gamma teras reaktor Triga 2000 Bandung mengunakan program Gamset. Perhitungan didasarkan pada persamaan atenuasi liner, sumber garis dan arah perambatan tiga dimensi. Selain panas gamma di teras, akan dihitung juga panas gamma di reflektor (Lazy Susan), dan di CIP untuk berbagai jenis bahan. Diperoleh hasil bahwa panas gamma di CIP cukup signifikan (0,87 w/g), tetapi di posisi Lazy Susan relatif kecil, rata-rata hanya 0,11 w/g. Dari hasil tersebut dapat disimpulkan bahwa penggunaan CIP untuk iradiasi perlu mempertimbangkan panas gamma dalam perhitungan LAK nya. Kata kunci: panas gamma, reaktor nuklir, reaktor penelitian, keselamatan reaktor 

scholarly journals Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

2011 ◽  
Vol 32 (4) ◽  
pp. 67-79
Author(s):  
Tomasz Bury
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Computer Code ◽  
Hydrogen Combustion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Lumped Parameter ◽  
Loss Of Coolant

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

Influence of PIG Mass, Launching Time and Turbulence Model on 3-D CFD Transient Simulation of PIG Motion

2016 ◽  
Author(s):  
Diego Jaimes Parilli ◽  
Armando Blanco ◽  
Janneth García
Keyword(s):  
Pressure Drop ◽  
Numerical Models ◽  
Control Volume ◽  
Asymptotic Relation ◽  
Fluid Pressure ◽  
Turbulence Models ◽  
Transient State ◽  
Terminal Velocity ◽  
Transient Simulation ◽  
Transient Pressure

Pigging procedures are common maintenance operations used to perform cleaning, draining and pipeline inspection in order to improve flow efficiency and operation cost. Despite these procedures are commonly used, questions still remain regarding the flow and the PIG motion features due to the complex interaction among pig, wall and flow, and the changes in internal fluid pressure and local fluid density. Currently, the PIG dynamic predictions are based on experimental data from short scale laboratory experiments and numerical models founded on physical simplification. So far, the transient of PIG motion calculated by methods that combine CFD and fluid-structure interaction in a 3D model and the influence of the physic and numerical features over the pig dynamics has not been analyzed yet. To provide a better understanding of pigging runs, this paper proposes a CFD methodology to obtain a 3D transient simulation of PIG motion. A moving control volume attached to the PIG let to solve the governing equation in a stationary mesh. This methodology is used to obtain the transient simulation of a PIG launched in a straight water pipeline for different PIG mass, launching time and turbulence models in order to study its influence over the PIG dynamics. The numerical results show a linear relation between the mass and the pressure drop in the transient state, but with no influence over the final stationary state. Also, an asymptotic relation between the transient pressure drop and the launching time was observed with no influence over the PIG terminal velocity. Besides, it is exposed the influence of the turbulence models (κ-ε, SST and BSL Reynolds Stress) in the results of pig motion; appreciable difference between the drop pressure of Omega-Based Stress Models (SST and BSL) and κ-ε turbulent model at steady state is shown, and, finally, a comparison of the velocity profiles at the interstice for each model was developed, this one shows an inaccuracy of the κ-ε model to describe the velocity profile in the walls proximities.

FAVOR Version 16.1: A Computer Code for Fracture Mechanics Analyses of Nuclear Reactor Pressure Vessels

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Computer Code ◽  
The United States ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Current Status ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Reactor Pressure

This paper describes the current status of the Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code which has been under development at Oak Ridge National Laboratory (ORNL), with funding by the United States Nuclear Regulatory Commission (NRC), for over twenty-five years. Including this most recent release, v16.1, FAVOR has been applied by analysts from the nuclear industry and regulators at the NRC to perform deterministic and probabilistic fracture mechanics analyses to review / assess / update regulations designed to insure that the structural integrity of aging, and increasingly embrittled, nuclear reactor pressure vessels (RPVs) is maintained throughout the vessel’s operational service life. Early releases of FAVOR were developed primarily to address the pressurized thermal shock (PTS) issue; therefore, they were limited to applications involving pressurized water reactors (PWRs) subjected to cool-down transients with thermal and pressure loading applied to the inner surface of the RPV wall. These early versions of FAVOR were applied in the PTS Re-evaluation Project to successfully establish a technical foundation that served to better inform the basis of the then-existent PTS regulations to the original PTS Rule (Title 10 of the Code of Federal Regulations, Chapter I, Part 50, Section 50.61, 10CFR 50.61). A later version of FAVOR resulting from this project (version 06.1 - released in 2006) played a major role in the development of the Alternative PTS Rule (10 CFR 50.61.a). This paper describes recent ORNL developments of the FAVOR code; a brief history of verification studies of the code is also included. The 12.1 version (released in 2012) of FAVOR represented a significant generalization over previous releases insofar as it included the ability to encompass a broader range of transients (heat-up and cool-down) and vessel geometries, addressing both PWR and boiling water reactor (BWR) RPVs. The most recent public release of FAVOR, v16.1, includes improvements in the consistency and accuracy of the calculation of fracture mechanics stress-intensity factors for internal surface-breaking flaws; special attention was given to the analysis of shallow flaws. Those improvements were realized in part through implementation of the ASME Section XI, Appendix A, A-3000 curve fits into FAVOR; an overview of the implementation of those ASME curve fits is provided herein. Recent results from an extensive verification benchmarking project are presented that focus on comparisons of solutions from FAVOR versions 16.1 and 12.1 referenced to baseline solutions generated with the commercial ABAQUS code. The verifications studies presented herein indicate that solutions from FAVOR v16.1 exhibit an improvement in predictive accuracy relative to FAVOR v12.1, particularly for shallow flaws.

scholarly journals Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling

2019 ◽  
Vol 5 (4) ◽  
pp. 297-303
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Temperature Field ◽  
Mass Flow ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Computer Code ◽  
Cfd Modeling ◽  
Problem Solution ◽  
Conductivity Equation ◽  
The Core ◽  
Significant Difference

Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.

scholarly journals Powering Down

2003 ◽  
Vol 125 (04) ◽  
pp. 46-48
Author(s):  
Harry Hutchinson
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
The United States ◽  
Nuclear Industry ◽  
Significant Feature ◽  
Pressurized Water ◽  
Safety Testing ◽  
Fuel Assemblies ◽  
Enriched Uranium

This article reviews that after a half century of safety testing for the nuclear industry, a key heat-transfer lab is losing its home. Columbia University’s Heat Transfer Research Facility has been the only place to go for key safety testing. Since the days of the Atoms for Peace program during the Eisenhower years, the lab has tested generations of nuclear reactor fuel assemblies. The lab’s clients over the years have included all the designers of pressurized water reactors in the United States and others from much of the world. The tests are primarily concerned with one small, but significant feature of a reactor core. A core contains as many as 3000 fuel assemblies, bundles of long, slender rods containing enriched uranium. Controlled fission among the bundles heats water to begin the series of heat-transfer cycles that send steam to the turbines that will drive generators.

scholarly journals Physical-chemical Background of the Potential Phase Transitions during Loss of Coolant Accidents in the Supercritical Water Loops of Various Generation IV Nuclear Reactor Types

2019 ◽  
Vol 63 (2) ◽  
pp. 333-339
Author(s):  
Gábor Györke ◽  
Attila R. Imre
Keyword(s):  
Pressure Drop ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Chemical Properties ◽  
Initial Pressure ◽  
Pressurized Water ◽  
Physical Chemical ◽  
Loss Of Coolant ◽  
Flash Boiling ◽  
Water Reactors

Loss of coolant accidents (LOCA) are a serious type of accidents for nuclear reactors, when the integrity of the liquid-loop breaks. While in traditional pressurized water reactors, pressure drop can cause flash boiling, in Supercritical-Water Cooled reactors, the pressure drop can be terminated by processes with fast phase transition (flash boiling or steam collapse) causing pressure surge or the expansion can go smoothly to the dry steam region. Modelling the pressure drop of big and small LOCAs as isentropic and isenthalpic processes and replacing the existing reactor designs with a simplified supercritical loop, limiting temperatures for various outcomes will be given for 24.5 and 25 MPa initial pressure. Using the proposed method, similar accidents for chemical reactors and other equipment using supercritical fluids can be also analyzed, using only physical-chemical properties of the given supercritical fluid.

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