VVER-440 Containment Thermal Hydraulic Analyses With MELCOR and CONTAIN Codes

2002 ◽  
Author(s):  
Gregory Gromov ◽  
Igor Lola ◽  
Stanislav Sholomitsky ◽  
Dmitry Gumenyuk ◽  
Valery Shikhabutinov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Sar Model ◽  
Release Data ◽  
Main Steam Line ◽  
Safety Features ◽  
Main Steam ◽  
Containment Pressure

In support of the analyses for the Rivne Nuclear Power Plant (RNPP) VVER-440/213 (Ukraine) Safety Analysis Report (SAR), detailed MELCOR and CONTAIN models of the containment were developed. The RNPP containment features a bubble condenser tower with air locks and active and passive spray systems. Code input models were developed to accurately represent the containment volumes, room interconnections, structural masses, and the engineering safety features. Although MELCOR 1.8.3 [1] was the primary tool for the SAR containment analysis, comparison calculations were performed using CONTAIN Version 1.12 [2]. Consequently, both the response of the VVER-440 containment to limiting design conditions as well as a comparison of the two codes is presented. In the context of SAR requirements, the present application was performed for design basis accidents with conservative assumptions to compare the containment temperature and pressure with design criteria. The peak containment pressure and temperature were evaluated using the most intensive release of the primary and secondary coolant into the hermetic compartments, in particular, for the large break loss of coolant accident and main steam line break. Conservative coolant release data were evaluated using the RELAP5/Mod3.2 SAR model. The selection of the accident scenario, initial and boundary conditions, and the major results are presented. The results of the analyses will be included in the design basis accident analysis chapter of the RNPP SAR.

LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 128-142
Author(s):  
J.-J. Huang ◽  
S.-W. Chen ◽  
J.-R. Wang ◽  
C. Shih ◽  
H.-T. Lin
Keyword(s):  
Nuclear Power ◽  
Coupled Model ◽  
Short Term ◽  
Containment System ◽  
Loss Of Coolant Accident ◽  
Main Steam Line ◽  
Main Steam ◽  
Temperature Responses ◽  
Term Analysis

Abstract This study established an RCS-Containment coupled model that integrates the reactor coolant system (RCS) and the containment system by using the TRACE code. The coupled model was used in both short-term and long-term loss of coolant accident (LOCA) analyses. Besides, the RELAP5/CONTAN model that only contains the containment system was also developed for comparison. For short-term analysis, three kinds of LOCA scenarios were investigated: the recirculation line break (RCLB), the main steam line break (MSLB), and the feedwater line break (FWLB). For long-term analysis, the dry-well and suppression pool temperature responses of the RCLB were studied. The analysis results of RELAP5/CONTAN and TRACE models are benchmarked with those of FSAR and RELAP5/GOTHIC models, and it appears that the results of the above four models are consistent in general trends.

Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China

2021 ◽  
Author(s):  
Haiying Chen ◽  
Shaowei Wang ◽  
Xinlu Tian ◽  
Fudong Liu
Keyword(s):  
Radiation Dose ◽  
Effective Dose ◽  
Half Life ◽  
Nuclear Power ◽  
Outer Boundary ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Typical Design ◽  
Total Effective Dose

Abstract The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.

The Sensitivity of the Addition of Low-Low Pressurizer Pressure in Coincidence With High Containment Pressure on M310 Type Nuclear Power Plant

2016 ◽  
Author(s):  
Qiu Yanfei
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Deterministic Analysis ◽  
Limit Value ◽  
Loss Of Coolant Accident ◽  
Stop Time ◽  
Design Basis Accident ◽  
Operator Action ◽  
Containment Pressure

Due to the new security system that the operator intervention is assumed to occur in 20 minutes is not acceptable, for the current M310 type nuclear power plant. The loss of coolant accident with Intermediate breaks in primary loop is the only one of design basis accident which need operator action in 20 minutes. For certain size break, the consequences are very sensitive to the pump stop time. According to deterministic analysis that for a certain size break, if stop the pump in 20 minutes after accident, the peak cladding temperature will exceed the limit value of 1204°C. Therefore, it is necessary to add low-low pressurizer pressure in coincidence with high containment pressure signal to stop pump automatically on M310 type nuclear power plant.

Analysis of ABWR Containment Pressure-Temperature Loads and Suppression Pool Hydrodynamics Following Loss of Coolant Accidents

2011 ◽  
Author(s):  
P. Sawant ◽  
M. Khatib-Rahbar ◽  
F. Moody ◽  
A. Drozd
Keyword(s):  
Computer Code ◽  
Behavior Modeling ◽  
Model Parameters ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Design Basis Accident ◽  
Temperature Loads ◽  
Suppression Pool ◽  
Containment Pressure

This paper focuses on the assessment of Advanced Boiling Water Reactor (ABWR) containment pressure-temperature and suppression pool hydrodynamics under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a phenomena identification and ranking table (PIRT) applicable to the ABWR containment response behavior, modeling of pressure-temperature loads using the MELCOR computer code, and analysis of suppression pool hydrodynamics parameters based on a mechanistic one-dimensional hydrodynamics model. A MELCOR 1.8.6 model with detailed nodalization of the ABWR containment is used to perform the containment pressure-temperature calculations following a design basis accident. The best estimate and several sensitivity calculations are performed for the ABWR containment using the suppression pool swell model. The sensitivity calculations demonstrate the influence of key model parameters and assumptions on the suppression pool hydrodynamics response. The comparison of containment pressure-temperature and the suppression pool swell analyses results to those reported in the ABWR licensing calculations showed reasonable agreement.

Environmental Qualification of an Actuator Torque Switch

2006 ◽  
Author(s):  
A. M. Chan ◽  
S. L. Barreca ◽  
T. Kostela
Keyword(s):  
Nuclear Power ◽  
Normal Operation ◽  
Normal Contact ◽  
Electric Actuator ◽  
Loss Of Coolant Accident ◽  
Accident Simulation ◽  
Qualification Testing ◽  
Service Conditions ◽  
Main Steam Line ◽  
Main Steam

Environmental qualification testing was performed on a modified Limitorque torque switch for the torque switch safety functions in the Limitorque type SMB actuators located inside and outside containment in a nuclear power plant. The torque switch specimen was installed in a Limitorque SMB-1 electric actuator mounted on an 8” Velan gate valve and operated with a customized programmable logic controller to allow normal torque switch behaviour to be observed. The present paper describes the qualification testing performed. The modified torque switch was aged to a 30-year service life at the normal service conditions for both inside and outside containment. Aging included radiation, thermal and cycle aging. A seismic test and then a combined Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) steam accident simulation were followed. After each stage of aging, functional tests were done to confirm normal insulation resistance, normal contact resistance and normal operation.

Single Passive Condenser Failure During a Large-Break LOCA in the SBWR-1200 Reactor

2003 ◽  
Author(s):  
S. T. Revankar ◽  
M. Ishii ◽  
Y. Xu ◽  
H. J. Yoon ◽  
L. Cheng
Keyword(s):  
Safety System ◽  
Integral Test ◽  
Cooling Systems ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Accident Conditions ◽  
Main Steam Line ◽  
Main Steam ◽  
Core Cooling ◽  
Safety Systems

The performance of the safety systems of a new design of the 1200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated when one of three passive condenser system failed. The accident considered was a break in the main steam line which is the major design basis accident. The integral test was performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The loss of single condenser system did not affect the LOCA transient indicating that the existing safety system will adequately handle the large break LOCA with single condenser failure. The details of the safety system behavior are presented.

Visual and Thermal Studies of Single Tube Reflood Under Typical PWR LB-LOCA Conditions

2010 ◽  
Author(s):  
Yiyun Jessy Zeng ◽  
Colin P. Hale ◽  
Simon P. Walker ◽  
Geoffrey F. Hewitt
Keyword(s):  
Nuclear Power ◽  
Thermal Studies ◽  
Experimental Investigations ◽  
Primary Coolant ◽  
Single Tube ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Droplet Production ◽  
Fuel Pin ◽  
Design Basis Accident

The common approach to safety in a nuclear power plant is to design the system to respond safely to a large postulated accident, the so-called design basis accident. Accidents more severe than the design basis accident (“severe accidents”) are assessed but the system is not designed to withstand them; they are considered too unlikely to require specific design actions. For the pressurised-water reactor (PWR), the design basis accident (DBA) is the Large Break Loss-of-Coolant Accident (LB-LOCA), in which it is assumed that one of the large inlet coolant pipes from the circulating pump to the reactor vessel is completely broken and moves apart to allow free discharge of the primary coolant from both broken ends. For this type of break total coolant loss will occur in 100s or less. Although the reactor is by this time sub-critical so that little power is produced from fission, a large amount of decay power exists and causes the fuel rod claddings to have a temperature in the region of 600–800 °C. This paper describes experimental investigations designed and performed in order to provide detailed information about the macroscopic behaviour of the steam-water flow occurring during the reflood phase following a PWR LB-LOCA. Specifically, a bottom-up rewetting process was studied, in which water droplets may be entrained in the vapour flow and contribute to cooling of the hot fuel pin before it is quenched. In these experiments the test section is initially preheated to temperatures up to 600 °C and then quenched by introducing water at the bottom of the tube at atmospheric pressure. During the course of this transient process axial temperature and heat flux profiles will be recorded, extending the existing databank of cases for code validation. Simultaneously, an axial viewing technique will be applied to observe the quench front, and any pre-cursory droplet production, occurring during these singletube reflood experiments. As part of the preliminary validation of this novel technique, a series of air-water vertical upflow conditions have been examined. The results of these preliminary studies provide detailed visualisation of typical entrainment processes likely to be encountered during single-tube reflood.

Application of Advanced Model of Zr-Based Cladding Oxidation in Steam-Oxygen-Nitrogen Gas Mixtures to Separate Effects Tests and Integral Experiments

2015 ◽  
Author(s):  
Alexander Vasiliev
Keyword(s):  
Diffusion Coefficient ◽  
Nuclear Power ◽  
Temperature Oxidation ◽  
Parabolic Kinetics ◽  
Design Basis ◽  
Design Basis Accident ◽  
Air Ingress ◽  
Pure Steam ◽  
Modeling Code ◽  
Advanced Model

During postulated design-basis or beyond-design-basis accident at nuclear power plant with PWR or BWR, the high temperature oxidation of Zr-based fuel claddings in H2O-O2-N2 gas atmosphere could take place. Recent experimental observations showed that the oxidation of those claddings in the air (or, more generally, in oxygen-nitrogen and steam-nitrogen mixtures) behaves in much more aggressive way (linear or enhanced parabolic kinetics) compared to oxidation in pure steam (standard parabolic kinetics). This is why an advanced model of Zr-based cladding oxidation was developed. For calculations of cladding oxidation in oxygen-nitrogen and steam-nitrogen mixtures, the effective oxygen diffusion coefficient in ZrO2+ZrN layer formed in cladding is used. The diffusion coefficient enhancement factor depends on ZrN content in ZrO2+ZrN layer. A numerical scheme was realized to determine ZrO2+ZrN/α-Zr(O) and α-Zr(O)/β-Zr layers boundaries relocation and layers transformations in claddings. The model was implemented to the SOCRAT best estimate computer modeling code. The SOCRAT code with advanced model of oxidation was successfully used for calculations of separate effects tests and air ingress integral experiments QUENCH-10, QUENCH-16 and PARAMETER-SF4.

scholarly journals Analysis of Containment Pressure and Temperature Changes Following Loss of Coolant Accident (LOCA)

2015 ◽  
Vol 5 (4) ◽  
pp. 1-8
Author(s):  
Van Thai Nguyen ◽  
Ngoc Dung Kieu
Keyword(s):  
High Pressure ◽  
Cooling System ◽  
Primary System ◽  
Similar Work ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Water ◽  
Main Steam Line ◽  
Main Steam ◽  
Work Done

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories.

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