Nodalization of NEK in APROS and Steady State Simulation

2014 ◽  
Author(s):  
Tadeja Polach ◽  
Klemen Debelak ◽  
Ivica Bašić ◽  
Luka Štrubelj
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Building Blocks ◽  
Reactor Pressure Vessel ◽  
Primary Circuit ◽  
Physical Parameters ◽  
Detailed Model ◽  
Reactor Pressure

A model of the primary circuit and part of the secondary circuit of the Slovenian Krško NPP – NEK was built using APROS - Advanced PROcess Simulation environment. The data used to describe the properties of the system modelled in APROS, were the data describing Krško NPP and its operational properties after the uprating and the introduction of the 18-month cycle. Basis for data collections, nodalization, structure and simplifications was NEK RELAP5\MOD3.3 Engineering Handbook and the 23rd cycle. In order to build a model describing all the important parameters, the available elements in APROS environment were used as building blocks for each system. The goal was to create a detailed model nodalization, which would give accurate results and would run on reasonable processing power. Each submodel was checked to verify that the partial results are within the allowable limits and that the description of the physical parameters is consistent with the real components. The model includes reactor pressure vessel, reactor coolant pumps and primary piping, steam generator, part of main steam, part of feedwater, pressurizer and reactor core kinetics. The regulation of pressurizer level and pressure, steam generator level and control rod is also modelled. The model consists of more than 400 thermal hydraulic volumes. The aim of building this model was a through thermal hydraulic analysis of the PWR systems present in the NPP Krško. Several simulations of the steady states at different power levels were performed. The resulting data describing the flow rates in steam generator feedwater, reactor pressure vessel, including bypass flows, heat transfer in reactor core and steam generator, thermal losses to containment, liquid level in pressurizer and steam generator, pressure drops in primary circuit and other parameters were then compared to the results of different types of calculation and to the testing data obtained from Krško NPP. The next step was to identify variations in results and determine whether they are consequence of wrong parameters, measurement deviation or numerical error. In that manner the model was verified and validated (in the sense of comparison with available system surveillance plant test results) to ensure the correct setup, initial and boundary conditions were applied in order to get reliable steady state results.

scholarly journals Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Alejandro Nuñez-Carrera ◽  
Raúl Camargo-Camargo ◽  
Gilberto Espinosa-Paredes ◽  
Adrián López-García
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Detailed Model ◽  
Lower Head ◽  
Reactor Pressure

The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

Inadvertent Blowdown Event in a Nuclear Boiling Water Reactor

2011 ◽  
Author(s):  
Robert Engel
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Service Level ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Point Of View ◽  
Primary Circuit ◽  
Event Scale ◽  
The Impact ◽  
Reactor Pressure

On March 6th 2007, the Leibstadt Nuclear Power Plant in Switzerland experienced an automatic blowdown of eight safety/relief valves installed on the main steam lines caused by a faulty electrical manipulation while performing planned maintenance during full power operation. Due to the temperature measurements inside the reactor recirculation system and the reactor pressure vessel this event, at a first glance, appeared to be Event No. 23 (Automatic Blowdown event) as an Emergency (Service Level C) Condition in accordance with the relevant reactor pressure vessel Thermal Cycle Diagram. According to the ASME Code Section III, Service Level C limits permit large deformations in areas of structural discontinuity which may necessitate the removal of a component from service for inspection or repair. This paper presents a summary of thermal-hydraulic, stress, fatigue, and fracture mechanical evaluations as well as plant inspections performed to demonstrate the impact of the event on the reactor pressure vessel and associated components and to fulfill the requirements of the Swiss Federal Nuclear Safety Inspectorate. It is shown that the primary circuit of the plant was not inadmissibly stressed by the event and that it was acceptable from a safety-related point of view to return the plant to service. Corresponding to the 7-level International Nuclear and Radiological Event Scale this event was rated afterwards as level 1 (anomaly) by the Swiss Federal Nuclear Safety Inspectorate.

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

2020 ◽  
Author(s):  
Matthew Walter ◽  
Minghao Qin ◽  
Daniel Sommerville
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Structural Evaluation ◽  
Loss Of Coolant Accident ◽  
Specific Analysis ◽  
Reactor Pressure

Abstract As part of the license basis of a nuclear boiling water reactor pressure vessel, a sudden loss of coolant accident (LOCA) event needs to be analyzed. One of the loads that results from this event is a sudden depressurization of the recirculation line. This leads to an acoustic wave that propagates through the reactor coolant and impacts several structures inside the reactor pressure vessel (RPV). The authors have previously published a PVP paper (PVP2015-45769) which provides a survey of LOCA acoustic loads on boiling water reactor core shrouds. Acoustic loads are required for structural evaluation of core shrouds; therefore, a defensible load is required. The previous research compiled plant-specific data that was available at the time. Since then, additional data has become available which will add to the robustness of the bounding load methodology that was developed. Investigations are also made regarding the shroud support to RPV weld, which was neglected from the previous study. This will allow a practitioner a convenient method to calculate bounding acoustic loads on all shroud and shroud support welds in the absence of a plant-specific analysis.

scholarly journals Comparative Analysis of Severe Accident at WWER-1000 NPP with MELCOR 1.8.5 and 2.1 Code Versions

2020 ◽  
pp. 30-40
Author(s):  
O. Kotsuba ◽  
Yu. Vorobyov ◽  
O. Zhabin ◽  
D. Gumenyuk
Keyword(s):  
Comparative Analysis ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Computer Code ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Bottom Part ◽  
Control Volumes ◽  
Vessel Bottom ◽  
Reactor Pressure

An overview of the main improvements in updated version 2.1 of MELCOR computer code related to more representative mathematical modeling of complex thermohydraulic severe accident processes of core degradation, transfer of molten fragments to the bottom of the reactor, heating and failure of the bottom of the reactor pressure vessel is presented. The elements of WWER-1000 NPP computer model for the MELCOR 1.8.5 (control volumes, thermal structures and structures of the reactor core) that are reproduced for a reactor with the primary side, the secondary side and the containment are described. The changes implemented in WWER-1000 NPP model for MELCOR 1.8.5 to convert it to MELCOR 2.1 version that are mainly related to more detailed modeling of the reactor core and reactor pressure vessel bottom are provided. The paper presents the results of comparative analysis of severe accident scenario of total station blackout at WWER-1000 NPP with MELCOR 1.8.5 and 2.1. The comparison demonstrates good agreement between the main parameters’ results (pressure and temperature in hydraulic elements of the primary, secondary sides and the containment, temperature of core elements, the mass of the generated non-condensed gases and their concentration in the containment) obtained with these code versions for severe accident in-vessel phase. The identified differences in the time of core structures degradation and reactor vessel bottom failure are insignificantly affected by the behavior of the parameters in the primary side and the containment in the in-vessel phase of the severe accident and are related to more detailed modelling of the reactor core and bottom part of the reactor in MELCOR 2.1.

Development of Technique for Controlling Natural Circulation Flow by Using Helium Gas

2014 ◽  
Author(s):  
Hirofumi Hatori ◽  
Naoto Yanagawa ◽  
Tetsuaki Takeda ◽  
Shumpei Funatani
Keyword(s):  
Steady State ◽  
Pressure Vessel ◽  
Control Method ◽  
Natural Circulation ◽  
Vertical Channel ◽  
Vertical Tube ◽  
Reactor Pressure Vessel ◽  
Circulation Flow ◽  
Helium Gas ◽  
Reactor Pressure

The purpose of this study is to investigate a control method of natural circulation flow of air by injection of helium gas. A depressurization accident by the primary pipe rupture is one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the double coaxial duct connecting between a reactor core and an intermediate heat exchanger (IHX) module breaks, air is expected to enter the reactor pressure vessel from the breach and oxidize in-core graphite structures. Then, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In such condition, injection of helium gas into the channel by a passive method can prevent occurrence of the natural circulation flow of air in the reactor pressure vessel. Therefore, it is thought that oxidation of in-core graphite structures by air ingress can be prevented by establishing this method. The experiment has been carried out regarding the natural circulation flow using a circular tube consisting of a reverse U-shaped type. The vertical channel consists of one side heated tube and the other side cooled tube. The experimental procedure is as follows. Firstly, the apparatus is filled with air and one vertical tube is heated. Then, natural circulation of air will be produced in the channel. After the steady state is established, a small amount of helium gas is injected from the top of the channel. The velocity, mole fraction, temperature of gas, and temperature of the tube wall are measured during the experiment. The results were obtained as follows. When the temperature difference between the both vertical tubes was kept at about 60K, the velocity of the natural circulation flow of air was measured about 0.17m/s. During a steady state, a small amount of helium gas was injected into the channel. When the volume of injected helium gas is about 5.7% of the total volume of the channel, the velocity of the natural circulation flow of air became around zero. After 810 seconds elapsed, the natural circulation flow suddenly produced again. The natural circulation flow of air can be controlled by injecting of helium gas.

3D modeling of the primary circuit in the reactor pressure vessel of a PHWR

2013 ◽  
Vol 265 ◽  
pp. 356-365 ◽  
Author(s):  
Damian Ramajo ◽  
Santiago Corzo ◽  
Nicolas Schiliuk ◽  
Norberto Nigro
Keyword(s):  
3D Modeling ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Primary Circuit ◽  
Reactor Pressure

Evaluation of Mechanical Properties of Primary Circuit Components Using SPT Technique

2013 ◽  
Author(s):  
Martin Březina ◽  
Jana Petzová ◽  
Ľudovít Kupča
Keyword(s):  
Mechanical Properties ◽  
Pressure Vessel ◽  
Power Plants ◽  
Electrical Discharge Machining ◽  
Reactor Pressure Vessel ◽  
Primary Circuit ◽  
Surface Sampling ◽  
Primary Coolant ◽  
Term Operation ◽  
Reactor Pressure

The paper deals with the evaluation of mechanical properties of safety-related components of the primary circuit of nuclear power plants (NPPs). During a long-term operation of NPPs, changes of mechanical properties occur. To ensure the safe operation of NPP it is necessary to monitor and evaluate these changes. One possibility how to solve this problem is a direct sampling and the assessment of the actual mechanical properties using the small punch test (SPT) technique. By the SPT technique it is possible to evaluate the basic tensile properties such as the ultimate tensile strength and the yield stress of the tested materials as well as the transition temperature. The authors describe model examples of surface sampling using the special Rolls Royce equipment SSam™-2. The SPT specimens were prepared from the removed samples by application of the wire electrical discharge machining and fine grinding. The prepared specimens were tested and the obtained data were evaluated. Original pieces of primary tubes with diameter 500 mm made of stainless steel type 08Ch18N10T (AISI 321) were used for these model examples. Another experimental material used was a block of the reactor pressure vessel wall cut from NPP Greifswald Unit 7 which contains a weld joint. Those experiments were a part of the preparatory activities for the planned sampling and the assessment of the actual mechanical properties of primary components such as the main primary coolant piping, the steam generator shell and the reactor pressure vessel.

Study on Structural Integrity Assessment of Reactor Pressure Vessel Based on Three-Dimensional Thermal-Hydraulics and Structural Analyses

2014 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Tadashi Watanabe ◽  
Yutaka Nishiyama
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Coolant Water ◽  
Reactor Pressure

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.

Application of Small Punch Testing Methods for Thermal Ageing Monitoring at Primary Circuit Components in Nuclear Power Plant

2015 ◽  
Author(s):  
Jana Petzová ◽  
Martin Březina ◽  
Michal Kapusňák ◽  
Ľudovít Kupča
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Thermal Ageing ◽  
Reactor Pressure Vessel ◽  
Actual State ◽  
Primary Circuit ◽  
Surface Sampling ◽  
Small Punch ◽  
Circuit Components ◽  
Reactor Pressure

This paper deals with the evaluation of material properties of safety-related components of the primary circuit of nuclear power plants (NPP) such as reactor pressure vessel (RPV), primary piping (PP) or steam generator (SG). The main degradation mechanism of NPP’s components is radiation damage, but these processes are situated only in the core region of the reactor pressure vessel (RPV). The main mechanical loading of all individual parts of NPP’s primary circuit are the influence of high pressure at elevated temperature till 300°C. These processes lead to the fatigue damage of structural materials, which is characterized as the thermal fatigue or thermal ageing. The reactors in Slovakia have been behind several decades in operation. Therefore, to ensure the safe and reliable operation of NPP, it is necessary to monitor and evaluate these changes. While the monitoring of the radiation degradation is standardly performed using surveillance programs during all plant life operation, the monitoring of thermal ageing at primary circuit components was realized only after few years of all NPP units operation. The concept of the thermal ageing assessment was divided into two possible approaches to evaluation. The first is long-term exposition of surveillance specimens. And the second approach is a direct surface sampling of heavy components after several years of operation. The Small Punch Test (SPT) methods are mainly used for evaluation of materials actual state. By the SPT technique it is possible to evaluate the basic tensile properties as the ultimate tensile strength and the yield stress of the tested materials from a very small amount of obtained material. The details of the original VUJE design program of thermal ageing monitoring of NPP primary circuit materials in Slovakia is described in this paper.

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