scholarly journals DOSE ASSESSMENT FOR TYPICAL NPP-2006 STAFF FOR DESIGN BASIS ACCIDENT TAKING INTO ACCOUNT SITE INFRASTRUCTURE

2018 ◽  
Vol 63 (1) ◽  
pp. 111-118 ◽  
Author(s):  
V. I. Orlovskaya ◽  
A. G. Trifonov
Keyword(s):  
Dose Assessment ◽  
Design Basis ◽  
Design Basis Accident

scholarly journals Dose assessment for NPP staff for beyond design basis accident taking into account typical site infrastructure

2020 ◽  
Vol 64 (4) ◽  
pp. 485-490
Author(s):  
V. I. Orlovskaya ◽  
I. G. Trifonov
Keyword(s):  
Effective Dose ◽  
Radiation Effect ◽  
Dose Assessment ◽  
Internal Exposure ◽  
External Exposure ◽  
Respiratory Protection ◽  
Annual Dose ◽  
Design Basis ◽  
Design Basis Accident ◽  
Total Effective Dose

Assessment of radiation effect on nuclear power plant staff was made for beyond design basis accident (4 hours period). The considered accident scenario includes emergency radionuclide emission through containment bypass. Assessment of radiation effect on NPP staff was done on the basis of radionuclide concentration distribution on site considering typical infrastructure. Concentration mapping was calculated by developed program module for COMSOL 3.5a application. The obtained data included average volume radionuclide activities in lower air layer, total inhalation dose, effective dose of external exposure, equivalent and effective dose in thyroid and total effective dose for NPP staff during beyond design basis accident. Doses from radioactive cloud (external exposure) and from inhalation (internal exposure) were estimated for the following radionuclides: 137Cs, 134Cs, 131I, 133I, 90Sr. In the case of selected beyond design basis accident the total effective dose of staff is 61,98 mSv for the first 4 hours after the accident beginning. This number is slightly above the threshold of the allowable annual dose limit for personnel in emergency situations (50 mSv). Taking into account that short-lived iodine radionuclides 131I и133I give the main contribution in the dose (50.23 mSv including 27.23 mSv for thyroid), such emergency actions as respiratory protection and iodine prophylaxis for the staff can significantly decrease the received doses.

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.

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 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
Keyword(s):  
Source Term ◽  
Reactor Facility ◽  
Design Basis ◽  
Design Basis Accident

Design basis accident analysis for the Ignitor experiment

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

Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios

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

The Research on Core Melting Process: Oxidation

2013 ◽  
Author(s):  
Jun Wang ◽  
Wenxi Tian ◽  
Jianan Lu ◽  
Yingying Ma ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Reactor Core ◽  
Melting Process ◽  
Coolant Pump ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Design Basis ◽  
Loss Of Coolant ◽  
Design Basis Accident ◽  
Station Blackout

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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