scholarly journals Analysis of a severe beyond design basis accident for the EGP-6 reactor facility at Bilibino NPP. Release source term formation

2018 ◽  
Vol 2018 (1) ◽  
pp. 99-111 ◽  
Author(s):  
Leonid Mikhailovich Parafilo ◽  
Ruben Ildarovich Mukhamadeev ◽  
Yury Dmitrievich Baranaev ◽  
Albert Petrovich Suvorov
2018 ◽  
Vol 4 (2) ◽  
pp. 135-142
Author(s):  
Ruben Mukhamadeev ◽  
Leonid Parafilo ◽  
Yury Baranaev ◽  
Albert Suvorov

Analysis was performed of dynamic phase of severe accident of the EGP-6 reactor of the Bilibino NPP, due to uncontrolled reactivity insertion initiated by withdrawal of two pare of automatic control rods with followed by full failure of reactor emergency protection system. This initial event leads to promt increasing of reactor core power up to 450% of nominal value with short period, coupled with rise of temperature of fuel, pressure and temperature of coolant. These factors lead to crisis of heat exchange with subsequent ruptures tubes of fuel assemblies and coolant blow down into graphite stack. All its lead to rise of pressure in reactor shell and damage of it, outflow of steam-water mixture through up-reactor area to ventilation system, communication corridors and reactor hall and further – to atmospheric release. Transient processes were calculated using code RELAP5/Mod3.2. It was considered stages of processes of fuel damage and evaluated dynamic of a number and degree of damaged fuel assembles. They were grouped on burn-up and for each group it was performed analysis of dynamic of damage values. Further it was considered processes of yield of fission products from damaged fuel with models, based on experimental data on yield of fission products from fuel material, used in assembles of Bilibino NPP fuel type (fuel tubes with steel cladding, where fuel material is grits of uranium dioxide in magnesium), under condition of severe accident, especially performed in SSC IPPE. Transport of fission products with steam and air up to release points was evaluated with models, based on experimental data of fission product transport through graphite stack under conditions of severe accident, also especially performed in SSC IPPE. Evaluation of source term was performed in accordance with accident dynamic and assumed modes of release for conservative and most possible approaches. It was noted good self-protection property of EGP-6 reactor under severe beyond design basis accident condition.


2021 ◽  
Author(s):  
Haiying Chen ◽  
Shaowei Wang ◽  
Xinlu Tian ◽  
Fudong Liu

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.


2015 ◽  
Author(s):  
Alexander Vasiliev

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.


2015 ◽  
Vol 98-99 ◽  
pp. 2235-2238 ◽  
Author(s):  
Massimo Zucchetti ◽  
Bruno Coppi ◽  
Maria Teresa Porfiri ◽  
Marco Riva

Kerntechnik ◽  
2015 ◽  
Vol 80 (4) ◽  
pp. 373-378
Author(s):  
A. Keresztúri ◽  
Gy. Hegyi ◽  
Cs. Maráczy ◽  
I. Trosztel ◽  
Á. Tóta ◽  
...  

Author(s):  
Jun Wang ◽  
Wenxi Tian ◽  
Jianan Lu ◽  
Yingying Ma ◽  
Guanghui Su ◽  
...  

Beyond-design basis accidents in the AP1000 may result in reactor core melting and are therefore termed core melt accidents. The aim of this work is to develop a code to calculate and analyze the oxidation of a single fuel rod with total failures of engineered safeguard systems under a certain beyond-design basis accident such as a gigantic earthquake which can result in station blackout and then total loss of coolant flow. Using the code, the responses of the most dangerous fuel rod in the AP1000 were calculated under the accident. A discussion involving fuel pellets melting, cladding rupture and oxidation, and hydrogen production then was carried out, focused on DNBR during coolant pump coastdown, the cladding intactness under different flow rates in natural circulation, and the delay effect on cladding rupture due to cladding oxidation. By the analysis of calculated results, several suggestions on guaranteeing the security of fuel rods were provided.


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