scholarly journals On Burnup Modelling Issues Associated with VVER–440 Fuels

2019 ◽  
Vol 9 (2) ◽  
pp. 1-9
Author(s):  
Branislav Vrban ◽  
Štefan Čerba ◽  
Jakub Lüley ◽  
Filip Osuský ◽  
Mikuláš Vorobeľ ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Computational Modelling ◽  
Special Treatment ◽  
Spent Fuel ◽  
Power Normalization ◽  
Cooling Period ◽  
Special Part ◽  
Full Power ◽  
Irradiated Fuel ◽  
Good Agreement

The paper investigates various computational modelling issues associated with VVER-440 fuel depletion, relevant to burnup credit. The SCALE system and the TRITON sequence are used for the calculations. The effects of variations in depletion parameters and used calculation methods on the isotopic vectors are investigated. The burnup behaviour of Gadolinium is quite important in actual core analysis, but its behaviour is somewhat complicated, requiring special treatment in numerical modelling and calculations. Therefore, a special part of the paper is devoted to the treatment of Gadolinium-bearing fuels. Moreover, some discussions on power normalization are also included. To assess the acquired modelling experience used to predict the VVER-440 spent fuel nuclide composition, the measured compositions of Novovoronezh NPP irradiated fuel assembly are compared to data calculated by TRITON sequence. The samples of fuel assembly with 3.6 wt. % U-235 enrichment underwent 4-cycle campaign of totally 1109 effective full power days in the core and cooling period of 1-13 years. Calculated concentrations are compared to measured values burdened with their experimental uncertainties for totally 47 nuclides. The calculated results show overall a good agreement for all nuclides, differences from measured are pointed out and discussed in the paper.

Preliminary Design of Unloading Device for MNSR LEU Conversion

2018 ◽  
Author(s):  
Hao Qian ◽  
Li Yiguo ◽  
Peng Dan ◽  
Wu Xiaobo ◽  
Lu Jin ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Reactor Vessel ◽  
Cover Plate ◽  
Spent Fuel ◽  
Long Distance ◽  
Control Rod ◽  
Small Cover ◽  
Sealing Performance ◽  
Safety And Reliability ◽  
Application Requirements

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.

scholarly journals Cooling Period Calculation of Evolutionary Power Reactor Spent Fuel for Dry Management Safety

2016 ◽  
pp. 17-21
Author(s):  
M. I. Youssef ◽  
G. F. Sultan ◽  
F. Morsi Hassan
Keyword(s):  
Heat Load ◽  
Power Reactor ◽  
Computer Code ◽  
Normal Operation ◽  
Spent Fuel ◽  
Cooling Period ◽  
Decay Heat ◽  
Safety Margins ◽  
Epr Parameters

The calculation of the evolutionary power reactor (EPR) spent fuel (SF) cooling period (CP) was performed. The CP was determined by comparing the heat load of a cask with the calculated value of EPR decay heat (DH). The EPR DH was calculated by the ORIGEN computer code based on the EPR parameters. For conservatively study, the EPR and ORIGEN parameters that lead to higher DH values were selected and safety margins were considered. The fitting tool was utilized in the calculation of CP to overcome the ORIGEN limitation. The resultant values of CP will maintain the peak cladding temperature (PCT) of SF lower than 400°C during storage, transport, and disposal. The results show that -for normal operation- the SF of EPR should stay in the pool at least 4.75 years before it is loaded to the passively cooled dry casks.

scholarly journals An Automated SeaFAST ICP-DRC-MS Method for the Determination of 90Sr in Spent Nuclear Fuel Leachates

Molecules ◽  
2020 ◽  
Vol 25 (6) ◽  
pp. 1429 ◽  
Author(s):  
Víctor Vicente Vilas ◽  
Sylvain Millet ◽  
Miguel Sandow ◽  
Luis Iglesias Pérez ◽  
Daniel Serrano-Purroy ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Analytical Approach ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Chemical Purification ◽  
Radiochemical Method ◽  
Good Agreement

To reduce uncertainties in determining the source term and evolving condition of spent nuclear fuel is fundamental to the safety assessment. ß-emitting nuclides pose a challenging task for reliable, quantitative determination because both radiometric and mass spectrometric methodologies require prior chemical purification for the removal of interfering activity and isobars, respectively. A method for the determination of 90Sr at trace levels in nuclear spent fuel leachate samples without sophisticated and time-consuming procedures has been established. The analytical approach uses a commercially available automated pre-concentration device (SeaFAST) coupled to an ICP-DRC-MS. The method shows good performances with regard to reproducibility, precision, and LOD reducing the total time of analysis for each sample to 12.5 min. The comparison between the developed method and the classical radiochemical method shows a good agreement when taking into account the associated uncertainties.

scholarly journals Data Analysis and Calibration of Radiocarbon Dating Results from the Cemetery of the Marquises of Jin

Radiocarbon ◽  
2001 ◽  
Vol 43 (1) ◽  
pp. 55-62 ◽  
Author(s):  
Xiangyang Lu ◽  
Zhiyu Guo ◽  
Hongji Ma ◽  
Sixun Yuan ◽  
Xiaohong Wu
Keyword(s):  
Data Analysis ◽  
High Precision ◽  
Radiocarbon Dating ◽  
Careful Analysis ◽  
Historical Record ◽  
Calibration Result ◽  
Special Part ◽  
Peking University ◽  
Very High ◽  
Good Agreement

The chronology study of the cemetery of Marquises of Jin is valuable to improving the chronological table of Marquis of Jin family. It is also helpful for improving the chronological table of the Zhou Dynasty. The samples were measured at Peking University (PKUAMS). We also made an interlaboratory check with Isotrace to ensure the accuracy. By careful analysis of archaeological information, we built different models and calibrated by OxCal. The calibration results, both sampling contexts and estimations, are in very good agreement with the historical record. Because the dates of some events correspond to the special part of the curve, the calibration gets very high precision. The calibration result of tomb M93 suggests that its host is Marquis Shangshu instead of Marquis Wen.

Coprecipitation Studies from Simfuel Solutions in 5 m NaCl Media

1996 ◽  
Vol 465 ◽  
Author(s):  
P. Diaz-Arocas ◽  
J. Garcia-Serrano
Keyword(s):  
Fission Products ◽  
Solution Concentration ◽  
Laser Fluorescence ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Minor Elements ◽  
Icp Ms ◽  
Upper Limits ◽  
Ph Range ◽  
Irradiated Fuel

ABSTRACTExtensive Research is performed in many countries in order to evaluate the spent fuel behaviour under repository conditions. Several aspects as the control of the oxidative spent fuel dissolution by secondary phases formation are not yet clear.Coprecipitation experiments from SIMFUEL solutions are performed to study if minor elements will influence the formation of secondary phases. Therefore, coprecipitation studies from SIMFUEL solutions aims at identification of stable phases of significant simulated fission products. These experiments provide upper limits for solution concentration and distribution ratios of simulate fission products at several pH values. SIMFUEL pellets, which simulate an irradiated fuel with burnup of 50 GWd/tU were provided by AECL Research Laboratories, Canada. Experiments were carried out by addition of an aliquot of the initial SIMFUEL solution to 5 m NaCI free of carbonates solution. The selected pH was maintained constant during the experiments. The pH range considered was from 5.5 to 9.3. Analyses of the solutions were performed for uranium by Laser fluorescence and for the minor elements by ICP-MS. Solid phases formed at pH 5.5 were dissolved and analysed by ICP-MS. Results of the evolution in solution vs. pH of simulated fission products concentrations are shown in this paper.

Numerical Analysis of Automobile Environment During the Cooling Period in Summer

2008 ◽  
Author(s):  
Kum-Bae Lee ◽  
Md. Faisal Kader ◽  
Young-Muk Youn
Keyword(s):  
Three Dimensional ◽  
Considerable Change ◽  
Practical Significance ◽  
Instrument Panel ◽  
Shelf Area ◽  
Cooling Period ◽  
Initial Stage ◽  
Temperature And Humidity ◽  
Humidity Profiles ◽  
Good Agreement

Understanding the temperature and humidity profiles inside a vehicle when the A/C is on during summer is of paramount practical significance to maintain comfortable environment. In this paper, the airflow patterns, temperature and humidity distributions are predicted utilizing a three-dimensional finite volume numerical method and measured experimentally inside the automobile. Numerical analyses of the three-dimensional geometry predict a detailed description of fluid flow patterns. The velocity vectors from the side instrument panel registers impinge on the front occupant’s chest to head then travel forward and finally reach to the rear parcel shelf area. The undesirable returning flow passing by the driver’s head and neck has been observed. The velocity vectors from the central instrument panel registers travel directly to the rear compartment. A recirculation has been observed near the occupant’s knee area of the front compartment and rear compartment. The Temperature comes down to a comfortable range almost linearly at the initial stage. After that no considerable change has been observed. A good agreement has been found between the numerical and experimental results.

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