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.

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.

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.

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.

Determination of Thermal-Hydraulic Loads on Reactor Internals in a DBA-Situation

2006 ◽  
Author(s):  
Ville Lestinen ◽  
Timo Toppila ◽  
Antti Timperi ◽  
Timo Pa¨ttikangas ◽  
Markku Ha¨nninen
Keyword(s):  
Boundary Condition ◽  
Nuclear Power ◽  
Break Point ◽  
Mechanical Analysis ◽  
Element Analysis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Cfd Analyses ◽  
Reactor Internals

According to Finnish regulatory requirements, reactor internals have to stay intact in design basis accident (DBA) situations, so that control rods can always penetrate into the core. This is the basic motivation to study and develop more detailed methods for analyses of thermal-hydraulic loads on reactor internals during the DBA situation in the Loviisa Nuclear Power Plant (NPP) in Finland. In this work, the studied accident situation is Large Break Loss of Coolant Accident (LBLOCA). The objective of this work is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. In the present model, the downcomer of a PWR is only included and the reactor internals will be added later. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. Both thermal-hydraulic and mechanical aspects are discussed in this paper. Firstly, the pressure boundary condition in the pipe break point was calculated with the system code. In the second step, CFD analyses were made. Finally, the full fluid-structure interaction coupling between the CFD and FEA codes was used. The codes used for development and simulations are APROS system code for boundary condition calculations, STAR-CD and FLUENT for CFD calculations and ABAQUS for FEA calculations.

Severe Fuel Damage Experiments With Advanced Cladding Materials to be Performed in the QUENCH Facility (QUENCH-ACM)

2008 ◽  
Author(s):  
L. Sepold ◽  
M. Große ◽  
M. Steinbru¨ck ◽  
J. Stuckert
Keyword(s):  
Nuclear Power ◽  
Cooling Water ◽  
Test Series ◽  
Experimental Program ◽  
Limited Data ◽  
Standard Type ◽  
Operational Conditions ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Accident Conditions

The QUENCH out-of-pile experiments are part of the Severe Fuel Damage (SFD) program at the Karlsruhe Research Center. They are to investigate the hydrogen source term that results from reflooding an uncovered core of a Light-Water Reactor (LWR) with emergency cooling water. In the QUENCH experimental program Zircaloy-4 was used as standard-type material for rod cladding and grid spacer. Up to the end of 2007, 12 QUENCH experiments have been performed with this type of cladding; two test bundles contained B4C and one AgInCd absorber. One experiment (QUENCH-12) was conducted with Zr1%Nb cladding (VVER-type). Due to the niobium-bearing cladding, the VVER-type test QUENCH-12 could be regarded as a precursor for the upcoming program “QUENCH-ACM” with advanced cladding materials, i.e. M5, Duplex, ZIRLO, to be tested under SFD or BDBA (beyond design basis accident) conditions. These materials were developed for longer operation times in nuclear power reactors and extended burnup. They are optimized regarding their corrosion behavior under operational conditions and were also tested for LOCA (loss of coolant accident) and RIA (reactivity-initiated accident) conditions by the manufacturers. However, there are only very limited data available on the behavior of the new alloys in the SFD/BDBA temperature range, i.e. above 1500 K. The QUENCH-ACM test series has been defined with three experiments, i.e. QUENCH-14 through QUENCH-16. As in the Zircaloy-4 experiments, fuel is represented by ZrO2 pellets. Also, the test section instrumentation will be as usual with thermocouples attached to the cladding, shroud, and cooling jacket at elevations between −50 mm and 1350 mm. The QUENCH-ACM test series is scheduled to be performed in the period of 2008–2010. Test matrix and test bundle arrangements are presented in this paper.

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