Optimization of power microfield distribution in JA profiles RK3+ fuel assemblies with 4.68% average enrichment for VVER-440 prospective fuel cycles

Kerntechnik ◽  
2020 ◽  
Vol 85 (4) ◽  
pp. 213-225
Author(s):  
A. A. Gagarinskiy ◽  
Zh. Yu. Liventseva ◽  
D. R. Kireeva ◽  
D. A. Oleksyuk ◽  
Yu. P. Kalinin ◽  
...  
Keyword(s):  
Kerntechnik ◽  
2018 ◽  
Vol 83 (4) ◽  
pp. 307-313 ◽  
Author(s):  
A. Gagarinskiy ◽  
E. Osipova ◽  
Yu. Kalinin
Keyword(s):  

Atomic Energy ◽  
2010 ◽  
Vol 108 (3) ◽  
pp. 155-164 ◽  
Author(s):  
A. A. Gagarinskii ◽  
V. V. Saprykin ◽  
M. P. Lizorkin ◽  
V. N. Proselkov

Author(s):  
Liao Yi ◽  
Wang Cong ◽  
Chen Lei

Modular boiling water reactor (MBWR) can be considered as a small sized economic simplified boiling water reactor (ESBWR). It has the advantage of easier fabrication, transportation and construction. In this paper, a 65MWe MBWR core was designed with natural circulation, passive safety features, high power density and an 18 months fuel cycle. The MBWR core consists of 104 fuel assemblies with 4.6 w/o U-235, the assemblies were divided into 3 batches based on the depletion level, the batches shuffled at the end of each cycle. The core converged to equilibrium after 8 fuel cycles. A steady-state equilibrium fuel cycle depletion analysis was performed over a 540 day cycle using the HELIOS and PARCS software. The control blades insertion patterns were chosen to minimize axial and radial power peaking and provide uniform burnup throughout the cycle. At the end of the equilibrium cycle, 16% of total control blade worth remained inserted and the average assembly burnup is 21.318 GWd/MTHM. Thermal hydraulic analysis was also performed to insure the core’s safety feature.


2021 ◽  
Vol 247 ◽  
pp. 10013
Author(s):  
O.V. Vilkhivskaya ◽  
I.A. Evdokimov ◽  
V.V. Likhanskii ◽  
E.Yu. Afanasieva

The present work continues the series of papers on the revision of the conventional technique for evaluation of leaking fuel burnup during reactor operation at nuclear power plants (NPPs). The focus was made on reduction of uncertainties in evaluation of leaking fuel burnup in modern fuel cycles at WWER-1000 power units. A set of models was proposed for express calculation of the build-up of caesium isotopes in fuel and to relate 134Cs/137Cs activity ratio with fuel burnup for each rod in the core. These models are based on routine neutronic calculations of pin-by-pin linear heat generation rates which are performed at NPPs for each particular fuel cycle with particular core loading pattern (however, these calculations do not provide data on caesium inventory in fuel). Previously, the proposed models have been validated against several practical cases. This latest validation study relied on the analysis of the most recent fuel cycles at two NPPs that reported spike-events and identified the leaking fuel assemblies (LFAs) after the reactor shutdown. The calculated 134Cs/137Cs activity ratios in the fuel of the LFAs were compared to the NPPs data on the activity measurements, and to the post-irradiation examination (PIE) data provided for one FA. A reasonable agreement between the model predictions and the experimental data on 134Cs/137Cs activity ratios in the fuel as a function of its burnup is shown for the advanced FA designs in modern fuel cycles.


2010 ◽  
pp. 50-56
Author(s):  
Pablo R. Rubiolo ◽  
Guy Chaigne ◽  
Pierre Peturand ◽  
Jérôme Bigot ◽  
Jean-François Desseignes ◽  
...  

2019 ◽  
Vol 12 (4) ◽  
pp. 62-70
Author(s):  
K.N. Proskuryakov ◽  
A.V. Anikeev ◽  
E. Afshar ◽  
D.A. Pisareva
Keyword(s):  

Materials ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 1818
Author(s):  
Di-Si Wang ◽  
Bo Liu ◽  
Sheng Yang ◽  
Bin Xi ◽  
Long Gu ◽  
...  

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.


2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.


Sign in / Sign up

Export Citation Format

Share Document