The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

2021 ◽  
Author(s):  
Riku Tuominen ◽  
Ville Valtavirta
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
Photon Emission ◽  
Relative Difference ◽  
Fission Rate ◽  
Decay Heat ◽  
Probability Table ◽  
Zone Division ◽  
Energy Dependent ◽  
Fuel Source

Abstract The estimation of spent nuclear fuel source term (decay heat, reactivity, nuclide inventory etc.) has several sources of uncertainty such as uncertainties in nuclear data, uncertainties in the operation history, choice of calculation parameters etc. In this work the effect of calculation parameters is studied by estimating the source term with the built-in burnup capability of Serpent. The effect of the following parameters is considered: depletion zone division, burnup steps, unresolved resonance probability table sampling, Doppler-Broadening Rejection Correction (DBRC) and energy dependent branching ratios. As a test case a 2D BWR fuel assembly was modelled by first running a burnup calculation followed by a decay calculation. The following source term components were considered when investigating the effect of the studied parameters: total decay heat, photon emission rate and spontaneous fission rate. In general the differences resulting from the use of different parameter variations were small for all three studied source term components. For the decay heat largest absolute relative difference was approximately 0.6 % and for the photon emission rate approximately 1.1 %. For the spontaneous fission rate maximum absolute relative difference of nearly 8 % was observed. For all three components the variation of the depletion zone division resulted in the largest relative differences. Clear differences were also observed for burnup step length and DBRC variations. The use of unresolved resonance probability table sampling and energy dependent branching ratios had an insignificant effect on the studied source term components.

Impact of Approximations in Operating History Data on Spent Fuel Properties with Serpent 2

2021 ◽  
Author(s):  
Silja Häkkinen
Keyword(s):  
Spontaneous Fission ◽  
Spent Nuclear Fuel ◽  
Emission Rate ◽  
Photon Emission ◽  
Fuel Properties ◽  
Fission Rate ◽  
Decay Heat ◽  
Activity Decay ◽  
Operating History ◽  
Assembly Activity

Abstract In this work, the effect of averaging operating history parameters such as power history, boron concentration and coolant density and temperature on spent nuclear fuel properties was investigated. The examined properties were assembly activity, decay heat, photon emission rate, spontaneous fission rate and the concentration of some mobile nuclides and fissile nuclides. Calculations were performed on two similar VVER-440 fuel assemblies irradiated in different positions of the core using Serpent 2. Averaging power history over the entire irradiation history had a significant effect on assembly activity, decay heat and photon emission rate overestimating these properties approximately 70 % right after irradiation. However, the effect quickly died out and after 10 years of cooling the effect was less than 1 %. If the last cycle (3rd cycle) was modelled accurately and the power density of only the first two cycles were averaged, the differences remained always below 1 %. The effect of operating history approximations on spontaneous fission rate and the nuclide concentrations was much smaller reamaining mostly below 1.5 %. The sensitivity of nuclide concentrations to approximations in individual operating history parameters was dependent on the nuclide in question and no trend applying to all studied nuclides could be observed.

scholarly journals FEATURES OF THE SOURCE TERM CALCULATION CAPABILITY OF THE NECP-X CODE

2021 ◽  
Vol 247 ◽  
pp. 10024
Author(s):  
Xingjian Wen ◽  
Zhouyu Liu ◽  
Kai Huang ◽  
Liangzhi Cao
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
High Fidelity ◽  
Memory Requirement ◽  
Compression Method ◽  
Heat Measurement ◽  
Neutron Spectra ◽  
Decay Heat ◽  
Heat Calculation ◽  
Full Activation

The source term calculation capability is developed for the high-fidelity neutronics code NECP-X. Generally, a full activation library is used, but the memory requirement is unacceptable for the high-fidelity calculation. In order to minimize the memory requirement during the calculation with very strict conditions, a new generalized activation chain compressed method is proposed based on the influence qualification. One basic compression element is a reaction channel or an isotope, and the influence of every compression element to the final results are qualified. To enlarge the range of application of the new compressed library, an effective method to determine representative problems, which utilizes the neutron spectra and neutron flux, is developed and analyzed. Based on the ENDF-VII.0, EAF-2010 evaluated nuclear library and the influence qualification-based activation library compression method, a new compressed activation library is generated. The VERA-3A problem and the KAIST problem are used to assess the accuracy and the efficiency of the new activation library. 85 measurements of decay heat from decay heat measurement facilities GE-Morris and CLAB are used to validate the decay heat calculation in NECP-X. The results show good accuracy of NECP-X in predicting radiation source term of the spent nuclear fuel and significant memory saving when using new compressed activation library.

scholarly journals VERIFICATION AND VALIDATION OF BACK-END CYCLE SOURCE TERM CALCULATION OF THE NODAL CODE RAST-K

2021 ◽  
Vol 247 ◽  
pp. 10025
Author(s):  
Jaerim Jang ◽  
Bamidele Ebiwonjumi ◽  
Wonkyeong Kim ◽  
Jinsu Park ◽  
Deokjung Lee
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
Interpolation Method ◽  
Prediction Method ◽  
Computation Time ◽  
Verification And Validation ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Decay Heat ◽  
Fuel Assemblies

Verification and validation (V&V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper. An isotope inventory prediction method is presented in this work which is implemented with RAST-K and the lattice code STREAM. STREAM generates cross-section and provides number density information by history branch calculations. RAST-K supplies the power history and three history indexes (boron concentration, moderator temperature and fuel temperature). The main feature of the newly implemented spent nuclear fuel (SNF) characterization is the direct consideration of three-dimensional (3D) core simulation conditions by using operation history information. As a result of this, it could reduce the computation time. The implemented SNF analysis capability have two main functions. The first is to predict isotope inventory by Lagrange non-linear interpolation method, using power history correction factors. The second is to calculate the radiological response activity, decay heat, and neutron/gamma source strengths. The V&V of these two functions are thus presented herein. The isotope inventory prediction is validated with measured data from ten SNF samples of Takahama-3 and six samples of Calvert Cliffs-1 pressurized water reactors (PWR). Eighteen decay heat measurements of Ringhals Unit 3 PWR fuel assemblies are then employed to validate the decay heat calculation results. In addition, STREAM is employed in a code-to-code comparison for verification. The fuel assemblies cover the burnup range 14.3 - 47.25 GWd/tU, initial enrichment of 2.1 - 4.11 235U w/o and cooling time of 3.96 to 20.01 years. The comparison to STREAM shows the accuracy of the RAST-K SNF and prediction of the decay heat is within 4%. Overall, this paper demonstrates that RAST-K SNF calculation can be applied to the back-end cycle source term analysis.

scholarly journals Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

2017 ◽  
Vol 153 ◽  
pp. 07035 ◽  
Author(s):  
Mikhail Ternovykh ◽  
Georgy Tikhomirov ◽  
Ivan Saldikov ◽  
Alexander Gerasimov
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Heat Power ◽  
Decay Heat ◽  
Long Term Storage ◽  
High Burnup ◽  
Term Storage

scholarly journals Prediction of the maximum temperature inside container with spent nuclear fuel

2018 ◽  
pp. 31-35
Author(s):  
S. Alyokhina ◽  
О. Dybach ◽  
A. Kostikov ◽  
D. Dimitriieva
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Storage Facility ◽  
Storage Container ◽  
Atmospheric Air ◽  
Decay Heat ◽  
Fuel Storage ◽  
Maximum Temperatures ◽  
Cooling Air

The definition of the thermal state of containers with spent nuclear fuel is important part of the ensuring of its safe storage during all period of storage facility operation. The this work all investigations are carried out for the storage containers of spent nuclear fuel of WWER-1000 reactors, which are operated in the Dry Spent Nuclear Fuel Storage Facility in Zaporizhska NPP. The analysis of existing investigations in the world nuclear engineering science concerning to the prediction of maximum temperatures in spent nuclear fuel storage container is carried out. The absence of studies in this field is detected and the necessity of the dependence for the maximum temperature in the storage container and temperature of cooling air on the exit of ventilation duct from variated temperatures of atmospheric air and decay heat formulation is pointed out. With usage of numerical simulation by solving of the conjugate heat transfer problems, the dependence of maximum temperatures in storage container with spent nuclear fuel from atmospheric temperature and decay heat is detected. The verification of used calculation method by comparison of measured air temperature on exit of ventilation channels and calculated temperature of cooling air was carried out. By regression analysis of numerical results of studies the dependence of ventilation air temperature from the temperature of atmospheric air and the decay heat of spent nuclear fuel was formulated. For the obtained dependence the statistical analysis was carried out and confidence interval with 95% of confidence is calculated. The obtained dependences are expediently to use under maximum temperature level estimation at specified operation conditions of spent nuclear fuel storage containers and for the control of correctness of thermal monitoring system work.

scholarly journals Modeling of decay heat removal from CONSTOR RBMK-1500 casks during long-term storage of spent nuclear fuel

Energy ◽  
2019 ◽  
Vol 170 ◽  
pp. 978-985 ◽  
Author(s):  
R. Poškas ◽  
V. Šimonis ◽  
H. Jouhara ◽  
P. Poškas
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Decay Heat ◽  
Long Term Storage ◽  
Term Storage

Air Cooling Temperature Analysis of Spent Fuel

2005 ◽  
Author(s):  
A. L. Laursen ◽  
F. J. Moody ◽  
J. C. Law
Keyword(s):  
Nuclear Reactor ◽  
Spent Nuclear Fuel ◽  
Natural Circulation ◽  
Maximum Temperature ◽  
Air Cooling ◽  
Spent Fuel ◽  
Decay Heat ◽  
Spent Fuel Pool ◽  
Large Water ◽  
The Relationship

Spent nuclear fuel is currently being stored at nuclear reactor sites. The spent fuel removed from the reactor is first placed in a large water pool to remove the initial decay heat. After several years, when the decay heat has dropped below a set level, the fuel is moved into concrete storage casks where natural circulation continues the cooling process. The purpose of this report is to predict, using a simplified analysis, how hot the fuel rods get when cooled by air in the cask. The increase in temperature and the decrease in density cause a chimney effect in the cask. This paper presents an analytical method of obtaining maximum fuel clad temperature in the cask. A non-dimensional model is derived, which is used to calculate the entrance and exit air velocities of the cask. The relationship between these velocities and the temperature used to obtain the maximum fuel clad temperature. A numerical scheme used to predict the maximum temperature is presented here and the results are compared to the analytical model. Both methods yielded corroborating results for fuel placed in the casks after spending similar amounts of time in a spent fuel pool.

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.

A Heat Dissipation Simulation Study in a CFD Way on Spent Nuclear Fuel Dry Stored in Underground Vertical Shaft

2017 ◽  
Author(s):  
Zhang Hong-jian ◽  
Yu Ren ◽  
Liu Xiao-fan
Keyword(s):  
Heat Dissipation ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Vertical Shaft ◽  
Thermal Safety ◽  
Spent Fuel ◽  
Ansys Fluent ◽  
Measuring Point ◽  
Decay Heat ◽  
Current Heat

In order to ensure the thermal safety of the spent fuel stored in an underground vertical shaft, an air current heat dissipation is simulated in CFD way using ANSYS FLUENT code. Forced convection heat dissipation is focused in the research. The air current in the shaft and the temperature distribution on the surface of the spent fuel canister are calculated. The result confirms the reliability and security of the spent fuel dry storage. Finally, based on the calculating result, a support structure is designed, and the storage position of the spent fuel canister in vertical shaft is discussed, to optimize the decay heat removal of the spent fuel, and to ensure the temperature measuring point is set in a reasonable position.

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