BOUNDARY CONDITION MODELING EFFECT ON THE SPENT FUEL CHARACTERIZATION AND FINAL DECAY HEAT PREDICTION FROM A PWR ASSEMBLY

In this paper, two main exercises have been carried out to describe the effect that varying an albedo boundary condition has in the computation of observables such as decay heat, neutron emission rate and nuclide inventory from a PWR fuel assembly (or a configuration of assemblies) during a depletion scenario. The SERPENT2 code was then employed to emphasize the importance of modeling a proper boundary condition for such purposes. Moreover, the effect of taking into account more than a single fuel-pin region for depletion studies while varying the type of boundary condition, was also accounted for. The first exercise has the main objective of comparing in a single fuel assembly the albedo variations ranging from 1.1 up to full vacuum conditions. By comparing to the reference assembly (considered to be the case of full reflective conditions), relative differences up to +17% were observed in decay heat and up to almost -30% in neutron emissions. Also, a clear dependence on the albedo was detected if more than one depletable zone was considered while computing the integral value of observables of interest. Regarding the second exercise, where a 3 × 3 configuration of fuel assemblies is being now considered with a reflector section in the middle, a negligible effect on the observables was observed for the single fuel pin zone case; instead, an effect in the 244Cm computation when analyzing two fuel pin-zones produced a change in the neutron emission rate during cooling time up to 2.5% (while comparing it to the reference single assembly case).

Download Full-text

Study on Pool Boiling Behavior of the Spent Fuel Under Accident Conditions

Volume 9: Student Paper Competition ◽

10.1115/icone25-66616 ◽

2017 ◽

Author(s):

Shang Gao ◽

Daogang Lu ◽

Han Wang ◽

Yuhang Zhong

Keyword(s):

Heat Transfer ◽

Heat Transfer Coefficient ◽

Transfer Coefficient ◽

Water Temperature ◽

Fuel Assembly ◽

Saturation Temperature ◽

Spent Fuel ◽

Decay Heat ◽

Phase Calculation ◽

Spent Fuel Pool

Under normal refueling and emergency full-core offload condition, the fuel assembly is removed to the Spent Fuel Pool (SFP). Decay heat produced by the spent fuel is carried out by the cooling system. Active cooling method is adopted by the traditional PWR nuclear power plants, which means decay heat is taken away depending on forced circulation of the pump. However, the spent fuel pool, under accident condition, will lost the forced circulating cooling capacity, which will be a threat of for the fuel building safety. To study the thermal-hydraulic characteristics in the SFP missing the forced cooling, through CFX methodology and experiment, change of temperature and heat transfer coefficient of the wall of the heating tube at different heights were discussed, meanwhile the streamline chart and temperature contour were obtained as well. The present result indicated that under different power conditions, different height of water temperature increased at first and then trend to stable at saturation temperature. For a single 9*9 spent fuel assembly, water temperature at the higher height is higher than the lower at the same time, and water temperature at higher location reached a stable value more quickly. In addition, power value had a significant impact on the time of reaching saturation temperature, for example, 7000s is needed to reach saturation under 8.68KW condition while only 3000s under 16.12KW, which illustrates that fuel unload power is crucial to the SFP safety. Based on the experiment data and single phase calculation, heat transfer coefficient at different height of the heating tube decreased slowly at first, and then increased. Especially, heat transfer coefficient at the highest test point rapidly decreased at one point because of boiling crisis.

Download Full-text

Analysis for the ARIANE GU3 sample: nuclide inventory and decay heat

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2021013 ◽

2021 ◽

Vol 7 ◽

pp. 14

Author(s):

Dimitri Rochman ◽

Alexander Vasiliev ◽

Hakim Ferroukhi ◽

Mathieu Hursin ◽

Raphaelle Ichou ◽

...

Keyword(s):

Predictive Power ◽

Nuclear Data ◽

Operating Conditions ◽

Two Dimensional ◽

Spent Fuel ◽

Dimensional Lattice ◽

Decay Heat ◽

Code Validation ◽

Manufacturing Tolerances ◽

Fuel Characterization

This study presents an analysis of the ARIANE GU3 sample, in terms of nuclide inventory, as well as sample rod and assembly decay heat. The validation of a number of CASMO5 and library versions are performed with regards to the measured nuclide inventory, taking into account two dimensional lattice simulations. Uncertainties due to various sources (nuclear data, operating conditions and manufacturing tolerances) are also provided, and are combined with biases into expanded uncertainties. This study is similar to a previous one on the GU1 sample and fit in the framework of code validation, as well as in the estimation of code predictive power for spent fuel characterization.

Download Full-text

Monticello BWR spent fuel assembly decay heat predictions and measurements

10.2172/5595925 ◽

1986 ◽

Cited By ~ 1

Author(s):

M.A. McKinnon ◽

J.W. Doman ◽

C.M. Heeb ◽

J.M. Creer

Keyword(s):

Fuel Assembly ◽

Spent Fuel ◽

Decay Heat

Download Full-text

Machine Learning of LWR Spent Nuclear Fuel Assembly Decay Heat Measurements

Nuclear Engineering and Technology ◽

10.1016/j.net.2021.05.037 ◽

2021 ◽

Author(s):

Bamidele Ebiwonjumi ◽

Alexey Cherezov ◽

Siarhei Dzianisau ◽

Deokjung Lee

Keyword(s):

Machine Learning ◽

Fuel Assembly ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Decay Heat ◽

Nuclear Fuel Assembly

Download Full-text

Preliminary Design of Unloading Device for MNSR LEU Conversion

Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation ◽

10.1115/icone26-81520 ◽

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.

Download Full-text

Strengthening the SSTC NRS Scientific and Technical Potential through Participation in the IAEA Coordinated Research Projects

Nuclear and Radiation Safety ◽

10.32918/nrs.2021.1(89).01 ◽

2021 ◽

pp. 5-13

Author(s):

Yu. Balashevska ◽

D. Gumenyuk ◽

Iu. Ovdiienko ◽

O. Pecherytsia ◽

I. Shevchenko ◽

...

Keyword(s):

Best Practices ◽

Safety Assessment ◽

Radiation Safety ◽

Research Projects ◽

Spent Fuel ◽

Research Activities ◽

Technical Potential ◽

Technical Issues ◽

Fuel Characterization ◽

Definition Of

The State Scientific and Technical Center for Nuclear and Radiation Safety (SSTC NRS), a Ukrainian enterprise with a 29-year experience in the area of scientific and technical support to the national nuclear regulator (SNRIU), has been actively involved in international research activities. Participation in the IAEA coordinated research activities is among the SSTC NRS priorities. In the period of 2018–2020, the IAEA accepted four SSTC NRS proposals for participation in respective Coordinated Research Projects (CRPs). These CRPs address scientific and technical issues in different areas such as: 1) performance of probabilistic safety assessment for multi-unit/multi-reactor sites; 2) use of dose projection tools to ensure preparedness and response to nuclear and radiological emergencies; 3) phenomena related to in-vessel melt retention; 4) spent fuel characterization. This article presents a brief overview of the abovementioned projects with definition of scientific contributions by the SSTC NRS (participation in benchmarks, development of methodological documents on implementing research stages and of IAEA technical documents (TECDOC) for demonstration of best practices and results of research carried out by international teams).

Download Full-text

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

Nuclear and Radiation Safety ◽

10.32918/nrs.2016.2(70).04 ◽

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.

Download Full-text

Characterization of the New Vertical Neutron Camera Designed for the Low Neutron Emission Rate Plasma in Large Helical Device

Plasma and Fusion Research ◽

10.1585/pfr.16.1402039 ◽

2021 ◽

Vol 16 (0) ◽

pp. 1402039-1402039

Author(s):

Siriyaporn SANGAROON ◽

Kunihiro OGAWA ◽

Mitsutaka ISOBE ◽

Yutaka FUJIWARA ◽

Hiroyuki YAMAGUCHI ◽

...

Keyword(s):

Emission Rate ◽

Neutron Emission

Download Full-text

ABSOLUTE STANDARDIZATION OF RADIOACTIVE NEUTRON SOURCES: II. THE USE OF THE F19(α, n)Na22 REACTION

Canadian Journal of Physics ◽

10.1139/p59-062 ◽

1959 ◽

Vol 37 (5) ◽

pp. 550-556 ◽

Cited By ~ 7

Author(s):

K. W. Geiger

Keyword(s):

Nuclear Reactions ◽

Emission Rate ◽

Neutron Emission ◽

National Research Council ◽

Neutron Output ◽

Neutron Sources ◽

Neutron Thermalization ◽

Γ Rays ◽

Absolute Standardization ◽

Good Agreement

Fluorine has only one stable isotope, F19. If neutrons are produced by the F19(α, n)Na22 reaction the neutron output can be calculated from the yield of the resulting radioactive Na22. The growth of Na22 (half-life, 2.58 years) has been measured in a neutron source consisting originally of 1.6 curies Po210 mixed with CaF2 powder. Since Na22 is a positron emitter, discrimination against γ-rays from Po210 and from nuclear reactions could be achieved by detecting the two positron annihilation quanta in coincidence. The Na22 growth has been followed over 20 months and is in agreement with the theoretical growth curve. Comparison with a calibrated Na22 source yielded a neutron emission rate of (10.70 ± 0.25) × 104 sec−1. This resulted in a neutron emission rate of (3.16 ± 0.10) × 106 sec−1 for the Ra-α-Be source of the National Research Council, in good agreement with (3.22 ± 0.05) × 106 sec−1 obtained by a neutron thermalization method.

Download Full-text