scholarly journals Analyzing Design Considerations for Disassembly of Spent Nuclear Fuel during Head-End Process of Pyroprocessing

2020 ◽  
Vol 2020 ◽  
pp. 1-13
Author(s):  
Young-Hwan Kim ◽  
Yung-Zun Cho
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Scale Up ◽  
High Heat ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Design Considerations ◽  
Nuclear Fuel Assembly ◽  
Cutting Processes

We have developed a practical-scale dry disassembling process to dismantle PWR (Pressurized Water Reactor) spent nuclear fuel assembly in the order of several tens of kilograms of heavy metal/batch to supply rod-cuts (cladding tube and UO2 pellets) for mechanical decladding process. Dry head-end disassembling process has advantages over the wet head-end process because of the lower risk of proliferation and treatment of spent fuel with relatively high heat and radioactivity. This study describes the main design considerations for the disassembling process of the spent nuclear fuel assembly during the dry head-end process. The down-ender, dismantling, extraction, and cutting technologies are analyzed and models have been designed for testing. The purpose of dry head-end disassembly process is to test the main device performance and to obtain scale-up data for practical-scale disassembling. With this in mind, design considerations were analyzed based on remoteness, and basic verification tests were performed. However, the authors used simulated fuel, instead of the actual spent fuel, owing to a lack of joint determination. In addition, in the present study, we did not consider the heat generated from minor actinides or the radioactivity of the fission product; these aspects will be considered in a future study. During the basic test performed in this study, a simulated assembly was completely disassembled using new methods, such as dismantling, extraction, and cutting processes. The practical-scale dry disassembling technology can be tested using scale-up data for reuse of the spent fuel.

Comparative Stress Analyses of Dropped Spent Nuclear Fuel Assembly in a Prototypal Cask

2021 ◽  
Author(s):  
SAEHANSOL KANG ◽  
Donghyun Kim ◽  
Yoon-suk Chang ◽  
Sanghwan Lee
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Nuclear Fuel Assembly ◽  
Stress Analyses

scholarly journals Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

Materials ◽  
2019 ◽  
Vol 12 (3) ◽  
pp. 494
Author(s):  
Alexander Vasiliev ◽  
Jose Herrero ◽  
Marco Pecchia ◽  
Dimitri Rochman ◽  
Hakim Ferroukhi ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Design Parameters ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
National Cooperative ◽  
Criticality Safety

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, keff, of the canister would comply with the imposed criticality safety criterion.

scholarly journals Analysis of Some Computer Codes Used for Evaluation of Spent Nuclear Fuel Radiation Parameters

2019 ◽  
pp. 44-50
Author(s):  
A. Smaizys ◽  
E. Narkunas ◽  
V. Rudychev ◽  
Y. Rudychev
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Input Data ◽  
Spent Nuclear Fuel ◽  
Operation Time ◽  
Computer Code ◽  
Chemical Compositions ◽  
Spent Fuel ◽  
Management Options ◽  
Computer Codes

The radiation parameters such as radionuclide content and activities, fluxes and energy spectrum of gamma and neutrons of spent nuclear fuel are essential when planning further spent fuel management options – interim wet or dry storage or disposal into a geological repository. Radiation parameters determine the design of a storage or disposal facility, what materials, structures and thicknesses of structures should be used to provide adequate biological shielding. Experimental measurements of spent fuel radiation parameters are rather complicated and expensive, therefore numerical methods are widely used. Various computer codes (APOLLO, BOXER, CASMO, FISPACT, ORIGEN-S, WIMS, etc.) have been developed to simulate the irradiation processes of nuclear fuel and to obtain resulting radiation parameters. Irrespective of the used computer code, the input data firstly must be entered into that code. When simulating nuclear fuel irradiation and burn-up in the reactor core, the geometrical parameters of the fuel assembly, materials’ data (chemical compositions, densities), the operating parameters of the reactor (power, operation time, coolant parameters, etc.) shall be entered into the program as initial data. Fairly often approximations of the input data are performed, for example, fuel rods in a fuel assembly are homogenized and geometrically described as a solid cylinder, the reactor operation time is assumed as continuous and at constant power. The particularity of the input data and accepted assumptions depend on what initial information is available and on the capabilities of the computer code. The modelled spent fuel radiation parameters depend not only on the input data and assumptions, but also on the cross-section databases that are used in computer codes. Computer codes TRITON, ORIGEN-S and FISPACT have been used to model the concentration of actinides and fission products in the spent fuel from the RBMK-1000 reactor. The obtained results are compared and possible reasons for the differences in the modelling results are discussed.

scholarly journals SENSITIVITY ANALYSIS OF PWR SPENT FUEL DUE TO MODELLING PARAMETER UNCERTAINTIES USING SURROGATE MODELS

2021 ◽  
Vol 247 ◽  
pp. 15009
Author(s):  
Bamidele Ebiwonjumi ◽  
Peng Zhang ◽  
Deokjung Lee
Keyword(s):  
Sensitivity Analysis ◽  
Fuel Assembly ◽  
Spent Nuclear Fuel ◽  
Nuclear Data ◽  
Surrogate Models ◽  
Source Terms ◽  
Spent Fuel ◽  
Parameter Uncertainties ◽  
Pressurized Water ◽  
Sensitivity Indices

In the BEPU (Best Estimate Plus Uncertainty) framework, uncertainty quantification (UQ) is a requirement to improve confidence and reliability of code predictions. Over the years, a lot of works have been done to quantify uncertainties in code predictions of spent nuclear fuel (SNF) characteristics due to nuclear data uncertainties. The purpose of this study is to quantify the uncertainty in pressurized water reactor (PWR) fuel assembly radiation source terms (isotopic inventory, activity, decay heat, neutron and gamma source) due to uncertainties in modeling parameters. The deterministic code STREAM is used to predict the source terms of a typical PWR fuel assembly following realistic and detailed irradiation history. For the sensitivity analysis (SA) and UQ, surrogate models are developed based on polynomial chaos expansion (PCE) and variance-based global sensitivity indices (i.e., Sobol’ indices) are employed. The global SA identifies the less important uncertain parameters, showing that the number of uncertain input parameters can be reduced. The surrogate model offers a significantly reduced computational burden even with large number of samples required for the SA/UQ of the model response.

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