Uncertainty quantification in decay heat calculation of spent nuclear fuel by STREAM/RAST-K

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

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

EPJ Web of Conferences ◽

10.1051/epjconf/201715307035 ◽

2017 ◽

Vol 153 ◽

pp. 07035 ◽

Cited By ~ 3

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

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Prediction of the maximum temperature inside container with spent nuclear fuel

Nuclear and Radiation Safety ◽

10.32918/nrs.2018.2(78).05 ◽

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.

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

Energy ◽

10.1016/j.energy.2018.12.217 ◽

2019 ◽

Vol 170 ◽

pp. 978-985 ◽

Cited By ~ 3

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

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Heat Transfer Modeling of Spent Nuclear Fuel Using Uncertainty Quantification and Polynomial Chaos Expansion

Journal of Heat Transfer ◽

10.1115/1.4037501 ◽

2017 ◽

Vol 140 (2) ◽

Cited By ~ 2

Author(s):

Imane Khalil ◽

Quinn Pratt ◽

Harrison Schmachtenberger ◽

Roger Ghanem

Keyword(s):

Heat Transfer ◽

Fuel Assembly ◽

Uncertainty Quantification ◽

Nuclear Fuel ◽

Polynomial Chaos ◽

Spent Nuclear Fuel ◽

Polynomial Chaos Expansion ◽

Complex Case ◽

Chaos Expansion ◽

Input Parameters

A novel method that incorporates uncertainty quantification (UQ) into numerical simulations of heat transfer for a 9 × 9 square array of spent nuclear fuel (SNF) assemblies in a boiling water reactor (BWR) is presented in this paper. The results predict the maximum mean temperature at the center of the 9 × 9 BWR fuel assembly to be 462 K using a range of fuel burn-up power. Current related modeling techniques used to predict the heat transfer and the maximum temperature inside SNF assemblies rely on commercial codes and address the uncertainty in the input parameters by running separate simulations for different input parameters. The utility of leveraging polynomial chaos expansion (PCE) to develop a surrogate model that permits the efficient evaluation of the distribution of temperature and heat transfer while accounting for all uncertain input parameters to the model is explored and validated for a complex case of heat transfer that could be substituted with other problems of intricacy. UQ computational methods generated results that are encompassing continuous ranges of variable parameters that also served to conduct sensitivity analysis on heat transfer simulations of SNF assemblies with respect to physically relevant parameters. A two-dimensional (2D) model is used to describe the physical processes within the fuel assembly, and a second-order PCE is used to characterize the dependence of center temperature on ten input parameters.

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

EPJ Web of Conferences ◽

10.1051/epjconf/202124710024 ◽

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.

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