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.

Synthesis of Uranium Phosphate Phases and Potential Retardation Effects on Spent Fuel Radionuclide Migration

2002 ◽  
Vol 757 ◽  
Author(s):  
A. S. Turner ◽  
D. J. Wronkiewicz
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Potassium Phosphate ◽  
Fission Products ◽  
Chemical Compositions ◽  
Spent Fuel ◽  
Variable Chemical ◽  
Calcium Phosphorus ◽  
Uranyl Phosphates ◽  
Uranium Phosphate

ABSTRACTThe UO2 in spent nuclear fuel is unstable in the oxidizing conditions within the volcanic tuffs at the proposed nuclear repository at Yucca Mountain, Nevada. Over time, the UO2 will oxidize and corrode, releasing actinides and fission products to the surrounding environment. However, uranyl (U6+) phosphates (autunite, phurcalite, sodium autunite, etc.) are stable in such an oxidizing environment. The mobility of released radionuclides may be greatly retarded if they can be incorporated into these naturally stable phosphate phases, while the complex structures, variable chemical compositions, and natural analogue occurrences of the uranyl phosphates suggests such a process is favorable. Current tests have focused on synthesizing such phases by reacting uranium oxynitrate or UO3 with a calcium, sodium, or potassium phosphate and a base (if necessary) in a Teflon reaction vessel. Excess water is added, and the solution is heated at 90°C for 7, 35, or 182 days. SEM analyses have confirmed that various uranium phosphate crystalline solids have formed. XRD results indicate that tests using two different calcium phosphorus source materials, Ca2P2O7 and Ca10(OH)2(PO4)6, have both created synthetic phosphuranylite, Ca(UO2)[(UO2)3(OH)2(PO4)2] 2*12H2O. The formation of this phase is appears to be kinetically favored over other similar phases. Results utilizing sodium phosphates, NaH2(PO4)*H2O and Na2HPO4, have produced sodium autunite (NaUO2PO4*2H2O), but other phases are probably also present. Test results utilizing potassium phosphate, K3PO4, were inconclusive. Experiments using surrogate radionuclides are currently being performed in order to determine whether radionuclides, such as 239Pu, 137Cs, and 99Tc, released from corroded spent nuclear fuel can become incorporated into the crystalline structure of specific uranium phosphate phases, effectively limiting any further migration.

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.

Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel

2006 ◽  
Vol 985 ◽  
Author(s):  
Jeffrey A. Fortner ◽  
A. Jeremy Kropf ◽  
James L. Jerden ◽  
James C. Cunnane
Keyword(s):  
Transition Zone ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Aqueous Environment ◽  
Spent Fuel ◽  
Aqueous Corrosion ◽  
Reduction Potentials ◽  
Chemical Effects ◽  
Before And After ◽  
X Ray Absorption

AbstractPerformance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

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