scholarly journals Post-Closure Safety Calculations for the Disposal of Spent Nuclear Fuel in a Generic Horizontal Drillhole Repository

Energies ◽  
2020 ◽  
Vol 13 (10) ◽  
pp. 2599 ◽  
Author(s):  
Stefan Finsterle ◽  
Richard A. Muller ◽  
John Grimsich ◽  
John Apps ◽  
Rod Baltzer
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Uncertainty Propagation ◽  
Radioactive Decay ◽  
Transport Processes ◽  
Sensitivity Analyses ◽  
Annual Dose ◽  
Performance Metric ◽  
Safety Function ◽  
Wide Range

The post-closure performance of a generic horizontal drillhole repository for the disposal of spent nuclear fuel (SNF) is quantitatively evaluated using a physics-based numerical model that accounts for coupled thermal-hydrological flow and radionuclide transport processes. The model incorporates most subcomponents of the repository system, from individual waste canisters to the geological far field. The main performance metric is the maximum annual dose to an individual drinking potentially contaminated water taken from a well located above the center of the repository. Safety is evaluated for a wide range of conditions and alternative system evolutions, using deterministic simulations, sensitivity analyses, and a sampling-based uncertainty propagation analysis. These analyses show that the estimated maximum annual dose is low (on the order of 10−4 mSv yr−1, which is 1000 times smaller than a typical dose standard), and that the conclusions drawn from this dose estimate remain valid even if considerable changes are made to key assumptions and property values. The depth of the repository and the attributes of its configuration provide the main safety function of isolation from the accessible environment. Long-term confinement of radionuclides in the waste matrix and slow, diffusion-dominated transport leading to long migration times allow for radioactive decay to occur within the repository system. These preliminary calculations suggest that SNF can be safely disposed in an appropriately sited and carefully constructed and sealed horizontal drillhole repository.

Inservice Inspection of Extended Dry Storage of Spent Nuclear Fuel, Part II: NDE/Sensor Technology Development and Codification

2021 ◽  
Author(s):  
Ryan M. Meyer ◽  
Jeremy Renshaw ◽  
Kenn Hunter ◽  
Mike Orihuela ◽  
Jim Stadler ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Technology Development ◽  
Storage System ◽  
Sensor Technology ◽  
Nondestructive Examination ◽  
Dry Storage ◽  
Safety Function ◽  
Majority Population ◽  
The Usa

Abstract This paper describes development and demonstration of nondestructive examination (NDE) technologies to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA to verify continued safe operation and that the canister confinement is intact and performing its intended safety function. Specifically, this work relates to NDE technology development for “canister” based DSS systems, which form the majority population of DSSs in the USA for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond the initial license or certified term. Physical access to the canister surface is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics related to inspections. Several activities to demonstrate NDE technologies for the inspections of different DSS systems are summarized.

Environmental Safety Aspects of the New Spent Nuclear Fuel Management and Storage System at Ignalina NPP

2007 ◽  
Author(s):  
P. Poskas ◽  
V. Ragaisis ◽  
J. E. Adomaitis
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Design Solution ◽  
Fuel Management ◽  
Nuclear Facilities ◽  
Annual Dose ◽  
Management Activity ◽  
Dose Constraint

In the framework of the preparation for the decommissioning of the Ignalina Nuclear Power Plant (INPP) a new Interim Spent Nuclear Fuel Storage Facility (ISFSF) will be built in the existing sanitary protection zone (SPZ) of INPP. In addition to the ISFSF, the new spent nuclear fuel management activity will include all necessary spent nuclear fuel retrieval and packaging operations at the Reactor Units, transfer of storage casks to the ISFSF, and other activities appropriate to the chosen design solution and required for the safe removal of the existing spent nuclear fuel from storage pools and insertion into the new ISFSF. The Republic of Lithuania regulations require that the average annual dose to the critical group members of population due to operation of nuclear facility shall not exceed dose constraint. If several nuclear facilities are located in the same SPZ, the same dose constraint shall envelope radiological impacts from all operating and planned nuclear facilities. The paper discusses radiological safety assessment aspects as relevant for the new nuclear activity to be implemented in the SPZ of INPP considering specificity of Lithuanian regulatory requirements. The safety assessment methodology aspects, results and conclusions as concern public exposure are outlined and discussed.

Mechanical Shock and Vibration Analysis of Spent Nuclear Fuel Carried by the Atlas Railcar

2020 ◽  
Author(s):  
Nicholas Klymyshyn ◽  
Kevin Kadooka ◽  
Pavlo Ivanusa ◽  
Casey Spitz
Keyword(s):  
Nuclear Fuel ◽  
Dynamic Models ◽  
Spent Nuclear Fuel ◽  
Pacific Northwest National Laboratory ◽  
Department Of Energy ◽  
Mechanical Shock ◽  
Structural Dynamic ◽  
Multimodal Transportation ◽  
Wide Range ◽  
Dynamics Models

Abstract Researchers at Pacific Northwest National Laboratory have completed a structural-dynamic analysis of spent nuclear fuel subjected to the mechanical shock and vibration environment that is anticipated during normal conditions of transport in casks carried by the Atlas railcar. The Atlas railcar is a new railcar design that is being developed specifically for the purpose of carrying spent nuclear fuel casks. The analysis used best-estimate railcar dynamics models of the Atlas railcar and considered 17 different spent nuclear fuel transportation cask systems, representing the current fleet of cask options. This work used NUCARS, a specialized railcar dynamics explicit finite element code to calculate railcar dynamic response to prescribed speeds and track configurations. The railcar dynamics models provided cask transient motion for a wide range of speeds and track conditions, generating a relatively large database of potential cask motion. All of the cask motion transients were then applied as loading conditions to LS-DYNA structural-dynamic models of a single fuel rod. The analyses predict that the Equipos Nucleares S.A./U.S. Department of Energy (ENSA/DOE) multimodal transportation test of 2017 provided a relatively stronger vibration environment than is expected from the Atlas railcar. This paper describes the analysis methods, the analysis results, and compares the results of the Atlas transportation analysis to the test results and analyses of the ENSA/DOE multimodal transportation test of 2017.

Site Descriptive Modeling as a Part of Site Characterization in Sweden: Concluding the Surface Based Investigations

2007 ◽  
Author(s):  
Johan Andersson ◽  
Kristina Skagius ◽  
Anders Winberg ◽  
Anders Stro¨m ◽  
Tobias Lindborg
Keyword(s):  
Nuclear Fuel ◽  
Rock Mechanics ◽  
Spent Nuclear Fuel ◽  
Three Dimensional ◽  
Site Investigation ◽  
Three Dimensional Modeling ◽  
Dimensional Modeling ◽  
Wide Range ◽  
A Site ◽  
Parameter Values

The Swedish Nuclear Fuel and Waste Management Co., SKB, is currently finalizing its surface based site investigations for the final repository for spent nuclear fuel in the municipalities of O¨sthammar (the Forsmark area) and Oskarshamn (the Simpevar/Laxemar area). The investigation data are assessed into a Site Descriptive Model, constituting a synthesis of geology, rock mechanics, thermal properties, hydrogeology, hydrogeochemistry, transport properties and a surface system description. Site data constitute a wide range of different measurement results. These data both need to be checked for consistency and to be interpreted into a format more amenable for three-dimensional modeling. The three-dimensional modeling (i.e. estimating the distribution of parameter values in space) is made in a sequence where the geometrical framework is taken from the geological models and in turn used by the rock mechanics, thermal and hydrogeological modeling. These disciplines in turn are partly interrelated, and also provide feedback to the geological modeling, especially if the geological description appears unreasonable when assessed together with the other data. Procedures for assessing the uncertainties and the confidence in the modeling have been developed during the course of the site modeling. These assessments also provide key input to the completion of the site investigation program.

scholarly journals Neutron Activation of Structural Materials of a Dry Storage System for Spent Nuclear Fuel and Implications for Radioactive Waste Management

Energies ◽  
2020 ◽  
Vol 13 (20) ◽  
pp. 5325
Author(s):  
Se Geun Lee ◽  
Jae Hak Cheong
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Storage Systems ◽  
Storage System ◽  
Residual Activity ◽  
Sensitivity Analyses ◽  
Dry Storage ◽  
Residual Radioactivity

In order to estimate the radiological characteristics of disused dry storage systems for spent nuclear fuel, a stepwise framework to calculate neutron sources (ORIGEN-ARP), incident neutron flux and reaction rate (MCNPX), effective cross-section (hand calculation), and residual activity (ORIGEN-2) was established. Applicability of the framework was demonstrated by comparing the residual activity of a commercialized storage system, HI-STORM 100, listed in the safety analysis report and calculated in this study. For a reference case assuming an impurity-free storage system, the modified effective cross-sections were theoretically interpreted and the need for managing disused components as a radioactive waste for at least four years was demonstrated. Sensitivity analyses showed that the higher burnup induces the higher residual radioactivity, and the impurity 59Co may extend the minimum decay-in-storage period up to 51 years within the reported range of 59Co content in stainless steel. The extended long-term storage over 100 years, however, caused no significant increase in residual radioactivity. Impurity control together with appropriate decay-in-storage was proposed as an effective approach to minimize the secondary radioactive waste arising from disused dry storage systems. The results of this study could be used to optimize the decommissioning and waste management plan regarding interim storage of spent fuel.

scholarly journals The Three-­Dimensional Neutron-­Physical Model of Spent Nuclear Fuel Storage System

2021 ◽  
Vol 20 ◽  
pp. 51-59
Author(s):  
О. R. Trofymenko ◽  
◽  
І. M. Romanenko ◽  
М. І. Holiuk ◽  
C. V. Hrytsiuk ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Storage Facility ◽  
Spent Fuel ◽  
Wide Range ◽  
Fuel Assemblies ◽  
Fuel Storage

The management of spent nuclear fuel is one of the most pressing problems of Ukraine’s nuclear energy. To solve this problem, as well as to increase Ukraine’s energy independence, the construction of a centralized spent nuclear fuel storage facility is being completed in the Chornobyl exclusion zone, where the spent fuel of Khmelnytsky, Rivne and South Ukrainian nuclear power plants will be stored for the next 100 years. The technology of centralized storage of spent nuclear fuel is based on the storage of fuel assemblies in ventilated HI-STORM concrete containers manufactured by Holtec International. Long-term operation of a spent nuclear fuel storage facility requires a clear understanding of all processes (thermohydraulic, neutron-physical, aging processes, etc.) occurring in HI-STORM containers. And this cannot be achieved without modeling these processes using modern specialized programs. Modeling of neutron and photon transfer makes it possible to analyze the level of protective properties of the container against radiation, optimize the loading of MPC assemblies of different manufacturers and different levels of combustion and evaluate biological protection against neutron and gamma radiation. In the future it will allow to estimate the change in the isotopic composition of the materials of the container, which will be used for the management of aging processes at the centralized storage of spent nuclear fuel. The article is devoted to the development of the three-dimensional model of the HI-STORM storage system. The model was developed using the modern Monte Carlo code Serpent. The presented model consists of models of 31 spent fuel assemblies 438MT manufactured by TVEL company, model MPC-31 and model HISTORM 190. The model allows to perform a wide range of scientific tasks required in the operation of centralized storage of spent nuclear fuel.

A Trade Study for Waste Concepts to Minimize HLW Volume

2008 ◽  
Vol 1124 ◽  
Author(s):  
Dirk Gombert ◽  
Joe Carter ◽  
Bill Ebert ◽  
Steve Piet ◽  
Tim Trickel ◽  
...  
Keyword(s):  
Waste Management ◽  
Nuclear Fuel ◽  
Heat Dissipation ◽  
Spent Nuclear Fuel ◽  
Management Strategies ◽  
Radioactive Decay ◽  
Fission Products ◽  
Purex Process ◽  
High Level Waste ◽  
Glass Waste

AbstractAdvanced nuclear fuel reprocessing can partition wastes into groups of common chemistry. This enables new waste management strategies not possible with the plutonium, uranium extraction (PUREX) process alone. Combining all of the metallic fission products in an alloy and the balance as oxides in glass minimizes high level waste (HLW) volume. Implementing a waste management strategy using state-of-the-art combined waste forms and storage to allow radioactive decay and heat dissipation prior to placement in a repository makes it possible to place almost 10x the HLW equivalent of spent nuclear fuel (SNF) in the same repository space. However, using generic costs based on preliminary studies for waste stabilization facilities and separations modules, this analysis shows that combining the non-actinide wastes and using only one glass waste form is the most cost-effective.

Influence of Defect Interactions on Diffusion Processes in UO2+x: a Key Issue for Understanding the Behaviour of Spent Nuclear Fuel.

2004 ◽  
Vol 824 ◽  
Author(s):  
Georgette Petot-Ervas ◽  
Gianguido Baldinozzi ◽  
Pascal Ruello ◽  
Lionel Desgranges ◽  
Georgeta Chirlesan ◽  
...  
Keyword(s):  
Electrical Conductivity ◽  
Diffusion Coefficient ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Diffusion Processes ◽  
Transport Processes ◽  
Accurate Information ◽  
Chemical Diffusion Coefficient ◽  
Large Defect ◽  
Matter Transport

AbstractThe transformation of UO2 into U3O8 is of technological and academical interest because of the severe consequences on the spent nuclear fuel management. The structural mechanism responsible for the isothermal transformation of UO2 into U3O8 seems still unclear. Several phases (UO2+x, U4O9, β-U3O7, α-U3O7, U3O8 were reported but their true structures, phase boundaries between their existence domains and matter transport processes are still a matter of debate. Gathering accurate information on the behaviour of uranium oxide is a key issue for understanding the behaviour of spent nuclear fuel. The chemical diffusion coefficient ( ~ D) of UO2+x was determined by electrical conductivity experiments. Measurements were performed in transient state for departure from stoichiometry in the range 0<x<0.17 (10-11<P(O2)<10-8 atm.)and for 973<T<1673 K. We have found that ~ D is a decreasing function of the departure from stoichiometry x. This behaviour was attributed to the presence of singly charged (2:2:2) Willis defects as suggested by equilibrium conductivity measurements. The decrease of Dchim can be explained by transport processes occurring via a dynamic exchange between isolated mobile defects and complex defects frozen in clusters or domains. At higher P(O2), near U4O9, the time to reach an equilibrium electrical conductivity value becomes increasingly long. This suggests the presence either of large defect aggregates or of complex defects arranged into domains. Furthermore, the analysis of the transport processes in non equilibrium conditions has allowed us to show that the results of ~ D are consistent with those of the oxygen diffusion coefficient within the P(O2) and temperature range of stability of the [2:2:2] clusters.

Calculation and Optimization of Plasma-Based Utilization Process of Inflammable Wastes after Reprocessing of Spent Nuclear Fuel of Closed Nuclear Cycle

2015 ◽  
Vol 1084 ◽  
pp. 178-182 ◽  
Author(s):  
Alexander Karengin ◽  
Alexey Karengin ◽  
Ivan Novoselov ◽  
Nikolay Tundeshev
Keyword(s):  
Nitric Acid ◽  
Acid Solution ◽  
Nuclear Fuel ◽  
Nitric Acid Solution ◽  
Spent Nuclear Fuel ◽  
Practical Implementation ◽  
Air Plasma ◽  
Wide Range ◽  
Mass Fractions ◽  
Processing Wastes

This work demonstrates the results of modeling the joint utilization process of the spent solutions of tributylphosphate with hexachlorinebutadiene for extracting uranium and plutonium from a nitric acid solution of spent nuclear fuel and its processing wastes. Calculations are made for a wide range of temperature and mass fractions of air plasma coolant. Also, optimal inflammable water-organic compositions and work regimes for the practical implementation of this process in air plasma were determined.

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