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.

Posttest Calculations of PARAMETER-SF4 Experiment With Air Ingress and Water Reflood Using SOCRAT/V3 Computer Modeling Code

2011 ◽  
Author(s):  
Alexander Vasiliev
Keyword(s):  
Experimental Data ◽  
Computer Modeling ◽  
Nuclear Power ◽  
Cooling System ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Vver Fuel ◽  
Design Basis ◽  
Air Ingress ◽  
Modeling Code

The PARAMETER-SF4 test conditions simulated a severe LOCA (Loss of Coolant Accident) NPP (nuclear power plant) sequence in which the overheated up to 1700–2300K core would be reflooded from the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in July 21, 2009, and was the fourth of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH” (scientific and industrial association 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. Heating was carried out electrically using 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 VVER (water-water energetic reactor, Russian type of pressurized water reactor). After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF4 test, the bottom flooding was initiated. The important feature of PARAMETER-SF4 test was the air ingress phase during which the air was supplied to the working section of experimental installation. It is known that zirconium oxidation in the air proceeds in a different way in comparison to oxidation in the steam. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the calculation of PARAMETER-SF4 experiment. Thermal hydraulics in PARAMETER-SF4 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V3 were compared with experimental data concerning different aspects of air ingress phase and thermal hydraulics behavior 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-SF4 test.

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.

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