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Applied Sciences
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Nuclear Wastes Management
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Nuclear Wastes Management

A special issue of Applied Sciences (ISSN 2076-3417). This special issue belongs to the section "Applied Physics General".

Deadline for manuscript submissions: closed (20 February 2022) | Viewed by 42579

Special Issue Editors

Prof. Bruno Merk

E-Mail Website
Guest Editor
School of Engineering, University of Liverpool, Liverpool L69 3BX, UK
Interests: strategic development of nuclear; nuclear waste management; partitioning and transmutation; advanced nuclear systems; core physics for nuclear reactor operation, design, analysis, and safety; development of modern modelling and simulation methods; reactor physics experiments
Special Issues, Collections and Topics in MDPI journals
Dr. Andreas Wilden

E-Mail Website
Guest Editor
Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, 52428 Jülich, Germany
Interests: actinide chemistry; actinides; centrifugal contactors; inorganic chemistry; lanthanide chemistry; lanthanide complexes; lanthanides; metallurgy & metallurgical engineering; nuclear chemistry; nuclear science & technology
Special Issues, Collections and Topics in MDPI journals
Dr. Alexander Stanculescu

E-Mail Website
Guest Editor
Retired, International Atomic Energy Agency, 1400 Vienna, Austria
Interests: reactor design and engineering; methods development; fuel cycle analysis; strategy and economics studies; non-electric applications of nuclear energy
Special Issues, Collections and Topics in MDPI journals

Special Issue Information

Dear Colleagues,

“Nuclear Partitioning and Transmutation is one of the most promising fields of nuclear technology. We expect that partitioning and transmutation technology would contribute to the enhancement of the efficiency of high-level waste disposal and the utilization of resources in the spent fuel. We believe that the basic research and development effort in this field would be beneficial for the future generations although it is not quite an alternative to the present back-end policy” [opening address of the first Information Exchange Meeting on Actinide and Fission Product Separation and Transmutation, 1990, T. Yamamoto].

Nuclear waste management technology of has made considerable progress since the 1990s, including the demonstration of the feasibility of separation and burning of plutonium and minor actinides at laboratory scale in order to make nuclear technology a more viable option in the clean energy mix. However, the technologies are now at a crossroads, requiring a major leap from laboratory-scale demonstration to industrial demonstration and application. This requires a serious rethinking of the approach to identify the ideal next steps and substantial research to progress into the next level.

We welcome for publication any kind of work related to nuclear waste management, from separation technologies to fuel production and reactor studies to burn nuclear waste, as well as systematic studies of possible innovative strategies and advanced experimental work and highly innovative reactor development.     

Prof. Bruno Merk
Dr. Andreas Wilden
Dr. Alexander Stanculescu
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Applied Sciences is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2400 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • Nuclear waste management
  • Partitioning
  • transmutation
  • Advanced reactors
  • Reprocessing
  • Actinide separation
  • Waste burners
  • Fast reactors
  • Accelerator-driven systems
  • Molten salt reactors
  • Fuel cycle scenario
  • National and international programmes
  • Advanced fuel cycle
  • Advanced fuels
  • Closed fuel cycle
  • Recycling
  • Waste management strategies
  • Re-use of actinides
  • Waste storage
  • Nuclear energy
  • Final disposal
  • Geological disposal
  • Waste conditioning

Published Papers (20 papers)

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Research

18 pages, 2813 KiB  
Article
A HELIOS-Based Dynamic Salt Clean-Up Study for iMAGINE
by Bruno Merk, Anna Detkina, Dzianis Litskevich, Omid Noori-Kalkhoran and Gregory Cartland-Glover
Appl. Sci. 2022, 12(17), 8748; https://doi.org/10.3390/app12178748 - 31 Aug 2022
Cited by 4 | Viewed by 1087
Abstract
Nuclear technologies have the potential to play a unique role in delivering low carbon energy for a future net-zero society. However, for long-term success, nuclear technologies will need to deliver innovative solutions as proposed in iMAGINE. One of the key challenges for the [...] Read more.
Nuclear technologies have the potential to play a unique role in delivering low carbon energy for a future net-zero society. However, for long-term success, nuclear technologies will need to deliver innovative solutions as proposed in iMAGINE. One of the key challenges for the envisaged highly integrated nuclear energy system is the need for a demand-driven salt clean-up system. The work described provides an insight into the interplay between a potential salt clean-up system and the reactor operation in a dynamic approach. The results provided will help to optimise the parameters for the salt clean-up process by delivering a dynamically calculated priority list, identifying the elements with great influence on reactor operation. The integrated model is used to investigate the ideal time for the initiation of the clean-up as well as the effect of different throughputs through the clean-up system on criticality as well as on the concentration of the elements in the reactor salt. Finally, a staggered approach is proposed with the idea to phase in the chemical clean-up processes step by step to keep the reactor critical. The results provide an essential step for the progress of iMAGINE as well as a basis for the interdisciplinary work required to bring iMAGINE into real operation. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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14 pages, 5044 KiB  
Article
Conceptual Design, Development, and Preliminary Safety Evaluation of a PWR Dry Storage Module for Spent Nuclear Fuel
by Taehyeon Kim, Kiyoung Kim, Donghee Lee, Taehyung Na, Sunghwan Chung and Yongdeog Kim
Appl. Sci. 2022, 12(9), 4587; https://doi.org/10.3390/app12094587 - 30 Apr 2022
Viewed by 1903
Abstract
Dry storage systems are one of the storage methods for spent nuclear fuel used in many countries that operate nuclear power plants. To ensure the safe storage of spent nuclear fuel, dry storage systems are designed to maintain radiation shielding, thermal management, and [...] Read more.
Dry storage systems are one of the storage methods for spent nuclear fuel used in many countries that operate nuclear power plants. To ensure the safe storage of spent nuclear fuel, dry storage systems are designed to maintain radiation shielding, thermal management, and subcritical and mechanical integrity. In addition, these systems must be able to withstand earthquakes, tornadoes, floods, extreme temperatures, and other operating and design-based accident conditions. In order to develop a model with safety and economic feasibility by analyzing various dry storage systems, a vertically dry storage module was proposed, and evaluations were performed on safety evaluation along with design requirements. As a result of the evaluation, all of the safety design requirements were met. This evaluation’s results can be used as basic data for the detailed design of the dry storage module to proceed with further research, including the preparation of a safety analysis report and experimental verification for licensing applications. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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15 pages, 2447 KiB  
Article
Defining the Challenges—Identifying the Key Poisoning Elements to Be Separated in a Future Integrated Molten Salt Fast Reactor Clean-Up System for iMAGINE
by Bruno Merk, Anna Detkina, Dzianis Litskevich, Michael Drury, Omid Noori-kalkhoran, Gregory Cartland-Glover, Leon Petit, Stefano Rolfo, Justin P. Elliott and Andrew R. Mount
Appl. Sci. 2022, 12(9), 4124; https://doi.org/10.3390/app12094124 - 19 Apr 2022
Cited by 8 | Viewed by 2165
Abstract
Nuclear fission technologies have the potential to play a significant role in the energy mix of a net-zero and sustainable society. However, to achieve the sustainability goal two significant challenges remain: efficient and sustainable fuel usage and the minimization of long-term nuclear waste. [...] Read more.
Nuclear fission technologies have the potential to play a significant role in the energy mix of a net-zero and sustainable society. However, to achieve the sustainability goal two significant challenges remain: efficient and sustainable fuel usage and the minimization of long-term nuclear waste. Civil nuclear molten salt systems and technologies offer the opportunity to address both, delivering future reactors at scale for efficient and effective power production and nuclear waste burnup. Potentially, both objectives could be fulfilled in one reactor system, which could significantly improve sustainability indices. The key to this innovation is demand driven development of a significantly reduced fuel cycle with enhanced proliferation resistance which offers further potential for improvement. To achieve these goals, a transformative approach for salt clean-up during molten salt reactor operation is proposed, by concentrating on the detection and removal of key neutron poisoning elements which prevent the reactor from long-term operation. To enable this highly innovative development work, a novel analysis of the evolving elementary fuel composition, their concentrations, and their criticality influence is now provided in this work. This, combined with consideration of the oxidation states of each of these elements then provides the basis for the selection of these key poisons and the development of advanced separation processes and process monitoring. This work also discusses the importance of the effective integration of physics and chemistry when systems modelling in achieving these system development goals. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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19 pages, 3771 KiB  
Article
Thermally Constrained Conceptual Deep Geological Repository Design under Spacing and Placing Uncertainties
by Jeremy Leong, Kumaraswamy Ponnambalam, Jeff Binns and Ali Elkamel
Appl. Sci. 2021, 11(24), 11874; https://doi.org/10.3390/app112411874 - 14 Dec 2021
Cited by 2 | Viewed by 1733
Abstract
The temperature evolution within a deep geological repository (DGR) is a key design consideration for the safe and permanent storage of the high-level radioactive waste contained inside used nuclear fuel containers (UFCs). Due to the material limitations of engineered components with respect to [...] Read more.
The temperature evolution within a deep geological repository (DGR) is a key design consideration for the safe and permanent storage of the high-level radioactive waste contained inside used nuclear fuel containers (UFCs). Due to the material limitations of engineered components with respect to high temperature tolerance, the Nuclear Waste Management Organization of Canada requires the maximum temperature within a future Canadian DGR to be less than 100 °C. Densely placing UFCs within a DGR is economically ideal, but greater UFC placement density will increase the maximum temperature reached in the repository. This paper was aimed to optimize (i) the separation between UFCs, (ii) the distance between container placement rooms, and (iii) the locations of the age-dependent UFCs in the placement rooms for a conceptual DGR constructed in crystalline rock. Surrogate-based optimization reduced the amount of computationally expensive evaluations of a COMSOL Multiphysics model used to study the temperature evolution within the conceptual DGR and determined optimal repository design points. Via yield optimization, nominal design points that considered uncertainties in the design process were observed. As more information becomes available during the design process for the Canadian DGR, the methods employed in this paper can be revisited to aid in selecting a UFC placement plan and to mitigate risks that may cause repository failure. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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22 pages, 1148 KiB  
Article
Appraisal of Nuclear Energy as an Alternative Option in South Africa’s Energy Scenario: A Multicriteria Analysis
by Solomon Eghosa Uhunamure, Ephraim Bonah Agyekum, Olatunde Samod Durowoju, Karabo Shale, Nthaduleni Samuel Nethengwe, Georges-Ivo Ekosse Ekosse and Tomiwa Sunday Adebayo
Appl. Sci. 2021, 11(21), 10349; https://doi.org/10.3390/app112110349 - 03 Nov 2021
Cited by 7 | Viewed by 3682
Abstract
South Africa is being confronted with an irregular power supply, leading to persistent load shedding due to aged and unreliable coal-fired power plants. Connected with coal as a generating source for electricity from fossil fuels are environmental concerns such as emissions of greenhouse [...] Read more.
South Africa is being confronted with an irregular power supply, leading to persistent load shedding due to aged and unreliable coal-fired power plants. Connected with coal as a generating source for electricity from fossil fuels are environmental concerns such as emissions of greenhouse gases and climate change impacts. Nuclear energy can allay the country’s dependence on coal as a source of energy. This article, therefore, reviews the feasibility of nuclear energy using a multicriteria analysis technique. A combination of Strengths, weaknesses, Opportunities, and Threats (SWOT) analysis and Analytical Hierarchy Process (AHP) was used to evaluate the external and internal factors that could either positively or negatively affect the country’s nuclear energy expansion drive. From the analysis, the country’s enabling laws and regulatory framework recorded the highest score of 39.2% under the strengths for the sector. In the case of the weaknesses, the high cost of construction and long construction framework recorded the highest weight, of 50.47%. Energy export and demand under the opportunities recorded a weight of 52.09%, ranking it as the highest opportunity for the sector. Seismic events were identified as the biggest threat for nuclear power expansion in the country, and the experts assigned a weight of 42.5% to this factor. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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19 pages, 1451 KiB  
Article
Comparative Safety Analysis of Accelerator Driven Subcritical Systems and Critical Nuclear Energy Systems
by Run Luo, Shripad T. Revankar and Fuyu Zhao
Appl. Sci. 2021, 11(17), 8179; https://doi.org/10.3390/app11178179 - 03 Sep 2021
Cited by 2 | Viewed by 1706
Abstract
The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly [...] Read more.
The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly different from those of the critical nuclear energy system (CNES). In this work, the safety characteristics of ADS are analyzed and compared with CNES by a developed neutronics and thermal-hydraulics coupled code named ARTAP. Three typical accidents are carried out in both ADS and CNES, including reactivity insertion, loss of flow, and loss of heat sink. The comparison results show that the power and the temperatures of fuel, cladding, and coolant of the CNES reactor are much higher than those of the ADS reactor during the reactivity insertion accident, which means ADS has a better safety advantage than CNES. However, due to the subcriticality of the ADS core and its low sensitivity to negative reactivity feedback, the simulation results indicate that the inherent safety characteristics of CNES are better than those of ADS under loss of flow accident, and the protection system of ADS would be quickly activated to achieve an emergency shutdown after the accident occurs. For the loss of heat sink, it is found that the peak temperatures of the cladding in the ADS and CNES reactors are lower than the safety limit, which imply these two reactors have good safety performance against loss of heat sink accidents. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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26 pages, 7615 KiB  
Article
The Effect of Porosity Change in Bentonite Caused by Decay Heat on Radionuclide Transport through Buffer Material
by Suu-Yan Liang, Wen-Sheng Lin, Gwo-Fong Lin, Chen-Wuing Liu and Chihhao Fan
Appl. Sci. 2021, 11(17), 7933; https://doi.org/10.3390/app11177933 - 27 Aug 2021
Cited by 2 | Viewed by 2084
Abstract
Bentonite is used as a buffer material in most high-level radioactive waste (HLW) repository designs. Smectite clay is the main mineral component of bentonite and plays a key role in controlling the buffer’s physical and chemical behaviors. Moreover, the long-term functions of buffer [...] Read more.
Bentonite is used as a buffer material in most high-level radioactive waste (HLW) repository designs. Smectite clay is the main mineral component of bentonite and plays a key role in controlling the buffer’s physical and chemical behaviors. Moreover, the long-term functions of buffer clay could be lost through smectite dehydration under the prevailing temperature stemming from the heat of waste decay. Therefore, the influence of waste decay temperatures on bentonite performance needs to be studied. However, seldom addressed is the influence of the thermo-hydro-chemical (T-H-C) processes on buffer material degradation in the engineered barrier system (EBS) of HLW disposal repositories as related to smectite clay dehydration. Therefore, we adopted the chemical kinetic model of smectite dehydration to calculate the amount of water expelled from smectite clay minerals caused by the higher temperatures of waste decay heat. We determined that the temperature peak of about 91.3 °C occurred at the junction of the canister and buffer material in the sixth year. After approximately 20,000 years, the thermal caused by the release of the canister had dispersed and the temperature had reduced close to the geothermal background level. The modified porosity of bentonite due to the temperature evolution in the buffer zone between 0 and 0.01 m near the canister was 0.321 (1–2 years), 0.435 (3–10 years), and 0.321 (11–20,000 years). In the buffer zone of 0.01–0.35 m, the porosity was 0.321 (1–20,000 years). In the simulation results of near-field radionuclide transport, we determined that the concentration of radionuclides released from the buffer material for the porosity of 0.321 was higher than that for the unmodified porosity of 0.435. It occurs after 1, 1671, 63, and 172 years for the I-129, Ni-59, Sr-90, and Cs137 radionuclides, respectively. The porosity correction model proposed herein can afford a more conservative concentration and approach to the real release concentration of radionuclides, which can be used for the safety assessment of the repository. Smectite clay could cause volume shrinkage because of the interlayer water loss in smectite and cause bentonite buffer compression. Investigation of the expansion pressure of smectite and the confining stress of the surrounding host rock can further elucidate the compression and volume expansion of bentonite. Within 10,000 years, the proportion of smectite transformed to illite is less than 0.05%. The decay heat temperature in the buffer material should be lower than 100 °C, which is a very important EBS design condition for radioactive waste disposal. The results of this study may be used in advanced research on the evolution of bentonite degradation for both performance assessments and safety analyses of final HLW disposal. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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15 pages, 4214 KiB  
Article
Lifetimes of Used Nuclear Fuel Containers Affected by Sulphate-Reducing Bacteria Reactions inside the Canadian Deep Geological Repository
by Jorge A. Garcia-Hernandez, Kumaraswamy Ponnambalam and Mythreyi Sivaraman
Appl. Sci. 2021, 11(17), 7806; https://doi.org/10.3390/app11177806 - 25 Aug 2021
Cited by 1 | Viewed by 1176
Abstract
The present work aims at approximating the reduction of sulphate to sulphide caused by sulphate-reducing bacteria (SRB) inside the Canadian deep geological repository in order to calculate the expected lifetime of used nuclear fuel containers (UFCs). Previous studies have assumed a conservative constant [...] Read more.
The present work aims at approximating the reduction of sulphate to sulphide caused by sulphate-reducing bacteria (SRB) inside the Canadian deep geological repository in order to calculate the expected lifetime of used nuclear fuel containers (UFCs). Previous studies have assumed a conservative constant concentration of sulphide at the host rock interface. The novelty of this study resides in the use of first-order kinetics to explicitly account for the SRB-induced sulphide production. This reaction term is developed following an empirical approach using published results on actual sulphate reduction by SRB and included in a coupled reaction-diffusion system. Lifetimes of UFCs are subsequently calculated following the conditions of two scenarios: having SRB active only at the region closest to the host rock and having SRB active at the host rock and throughout the bentonite clay. This study shows that the mean lifetimes of UFCs in both cases are above one million years. However, more accurate results would require the characterization of the host rock and groundwater of the prospective emplacement, as well as additional experiments on growth and sulphide production by the microbial communities from the site. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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21 pages, 5592 KiB  
Article
Evaluating Reactivity Control Options for a Chloride Salt-Based Molten Salt Zero-Power Reactor
by Bruno Merk, Anna Detkina, Seddon Atkinson, Dzianis Litskevich and Gregory Cartland-Glover
Appl. Sci. 2021, 11(16), 7447; https://doi.org/10.3390/app11167447 - 13 Aug 2021
Cited by 2 | Viewed by 1613
Abstract
Molten salt reactors have gained substantial interest in recent years due to their flexibility and their potential for simplified closed fuel cycle operation for massive net-zero energy production. However, a zero-power reactor experiment will be an essential first step in the process of [...] Read more.
Molten salt reactors have gained substantial interest in recent years due to their flexibility and their potential for simplified closed fuel cycle operation for massive net-zero energy production. However, a zero-power reactor experiment will be an essential first step in the process of delivering this technology. The topic of the control and shutdown for a zero-power reactor is, for the first time, introduced through a literature review and a reduction in the control approaches to a limited number of basic functions with different variations. In the following, the requirements for the control and shutdown systems for a reactor experiment are formulated, and based on these assessments, an approach for the shutdown, i.e., splitting the lower part of the core with a reflector, and an approach for the control, i.e., a vertically movable radial reflector, are proposed. Both systems will be usable for a zero-power system with a liquid as well as a solid core, and even more importantly, both systems somehow work at the integral system level without disturbing the central part of the core which will be the essential area for the experimental measurements. Both approaches were investigated as a singular system, in addition to their interactions with one another and the sensitivity of the control system. This study demonstrates that both proposed systems are able to deliver the required characteristics with a sufficient shutdown margin and a sufficiently wide control span. The interaction of the system is shown to be manageable, and the sensitivity is at a very good level. The multi-group Monte Carlo approach was cross-evaluated by a continuous energy test, leading to good results, but they also demonstrate that there is room for improvement. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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10 pages, 3050 KiB  
Article
Comparison of Gamma and Neutron Dose Rate Variations with Time for Cast Iron and Metal–Concrete Casks Used for RBMK-1500 Spent Fuel Storage
by Arturas Smaizys, Ernestas Narkunas, Gintautas Poskas and Povilas Poskas
Appl. Sci. 2021, 11(16), 7362; https://doi.org/10.3390/app11167362 - 10 Aug 2021
Cited by 1 | Viewed by 1585
Abstract
The present SF management concept in Lithuania envisages that spent RBMK-1500 fuel will be stored in dry storage containers for 50 years, before being disposed of in a deep geological repository. However, the risk that a deep geological repository will not be constructed [...] Read more.
The present SF management concept in Lithuania envisages that spent RBMK-1500 fuel will be stored in dry storage containers for 50 years, before being disposed of in a deep geological repository. However, the risk that a deep geological repository will not be constructed at the planned time should be taken into account, and the extension of SF storage over 50 years should be considered. This paper presents a comparison of gamma and neutron dose rate distributions and variations with planned and extended storage times for cast iron and metal–concrete containers loaded with RBMK-1500 SF. All calculations were performed using the SCALE computer codes system. The modeling results show that the overall shielding properties of the CONSTOR® RBMK-1500 container containing the same neutron and gamma sources are better than those of the CASTOR® RBMK-1500 container. During an extended storage period (from 50 to 300 years), the total dose rate would decrease considerably and the dose rate due to neutrons would become dominant for both containers. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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12 pages, 813 KiB  
Article
Impact of Updated OECD/NEA Thermodynamic Database on the Safety Assessment of Radioactive Waste Repository Studied Using RESRAD-OFFSITE Code
by Jun-Yeop Lee, Sang June Park and Seokyoung Ahn
Appl. Sci. 2021, 11(16), 7269; https://doi.org/10.3390/app11167269 - 06 Aug 2021
Cited by 1 | Viewed by 1549
Abstract
A RESRAD-OFFISTE computational code for the safety assessment model of a radioactive waste repository was utilized to evaluate the influence of the updated OECD/NEA thermodynamic database on the safety assessment model in terms of exposure dose. The solubility data as the input parameter [...] Read more.
A RESRAD-OFFISTE computational code for the safety assessment model of a radioactive waste repository was utilized to evaluate the influence of the updated OECD/NEA thermodynamic database on the safety assessment model in terms of exposure dose. The solubility data as the input parameter for the RESRAD-OFFSITE code obtained with two different sets of chemical thermodynamic databases such as JAEA-TDB and amended JAEA-TDB reflecting the updates of the OECD/NEA thermodynamic database were calculated and compared with each other. As a result, almost identical exposure doses were obtained due to the remarkable similarity between the solubility data of various radionuclides for both chemical thermodynamic databases. In contrast, dramatic changes in exposure dose were observed with varying distribution coefficients. Thermodynamic calculations indicated that the aqueous species distribution can be significantly changed by the selection of a chemical thermodynamic database and thus the relevant distribution coefficient can also be influenced as a consequence. Accordingly, the result obtained in the present work indicated that (i) the impact of the updated chemical thermodynamic data was somewhat minor from the viewpoint of the solubility and (ii) the distribution coefficient, which can be sensitively influenced by the predominant chemical species, produced a remarkable change in the exposure dose. This work provided an insight into the precise exposure dose calculation in terms of the reliable estimation of the distribution coefficient by means of a surface complexation model, which can predict the distribution coefficient as a function of groundwater composition coupled with a chemical speciation calculation based on up to date chemical thermodynamic data. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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22 pages, 11810 KiB  
Article
Innovative Investigation of Reflector Options for the Control of a Chloride-Based Molten Salt Zero-Power Reactor
by Bruno Merk, Anna Detkina, Seddon Atkinson, Dzianis Litskevich and Gregory Cartland-Glover
Appl. Sci. 2021, 11(15), 6795; https://doi.org/10.3390/app11156795 - 23 Jul 2021
Cited by 2 | Viewed by 1836
Abstract
Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operations for massive net-zero energy production. However, a zero-power reactor experiment will be an essential first step into the process [...] Read more.
Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operations for massive net-zero energy production. However, a zero-power reactor experiment will be an essential first step into the process delivering this technology. The choice of the optimal reflector material is one of the key issues for such experiments since, on the one hand, it offers huge cost savings potential due to reduced fuel demand; on the other hand, an improper choice of the reflector material can have negative effects on the quality of the experiments. The choice of the reflector material is, for the first time, introduced through a literature review and a discussion of potential roles of the reflector. The 2D study of different potential reflector materials has delivered a first down-selection with SS304 as the representative for stainless steel, lead, copper, graphite, and beryllium oxide. A deeper look identified, in addition, iron-based material with a high Si content. The following evaluation of the power distribution has shown the strong influence of the moderating reflectors, creating a massively disturbed power distribution with a peak at the core boundary. This effect has been confirmed through a deeper analysis of the 2D multi-group flux distribution, which led to the exclusion of the BeO and the graphite reflector. The most promising materials identified were SS304, lead, and copper. The final 3D Monte Carlo study demonstrated that all three materials have the potential to reduce the required amount of fuel by up to 60% compared with NaCl, which has been used in previous studies and is now taken as the reference. An initial cost analysis has identified the SS304 reflector as the most attractive solution. The results of the 2D multi-group deterministic study and the 3D multi-group Monte Carlo study have been confirmed through a continuous energy Monte Carlo reference calculation, showing only minor differences. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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20 pages, 4855 KiB  
Article
On the Dimensions Required for a Molten Salt Zero Power Reactor Operating on Chloride Salts
by Bruno Merk, Anna Detkina, Seddon Atkinson, Dzianis Litskevich and Gregory Cartland-Glover
Appl. Sci. 2021, 11(15), 6673; https://doi.org/10.3390/app11156673 - 21 Jul 2021
Cited by 5 | Viewed by 2238
Abstract
Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of [...] Read more.
Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of a zero-power reactor for the process of developing a new, innovative rector concept, such as that required for the molten salt fast reactor based on iMAGINE technology, which operates directly on spent nuclear fuel, is described here. It is based on historical developments as well as the current demand for experimental results and key factors that are relevant to the success of the next step in the development process of all innovative reactor types. In the systematic modelling and simulation of a zero-power molten salt reactor, the radius and the feedback effects are studied for a eutectic based system, while a heavy metal rich chloride-based system are studied depending on the uranium enrichment accompanied with the effects on neutron flux spectrum and spatial distribution. These results are used to support the relevant decision for the narrowing down of the configurations supported by considerations on cost and proliferation for the follow up 3-D analysis. The results provide for the first time a systematic modelling and simulation approach for a new reactor physics experiment for an advanced technology. The expected core volumes for these configurations have been studied using multi-group and continuous energy Monte-Carlo simulations identifying the 35% enriched systems as the most attractive. This finally leads to the choice of heavy metal rich compositions with 35% enrichment as the reference system for future studies of the next steps in the zero power reactor investigation. An alternative could be the eutectic system in the case the increased core diameter is manageable. The inter-comparison of the different applied codes and approaches available in the SCALE package has delivered a very good agreement between the results, creating trust into the developed and used models and methods. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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19 pages, 2069 KiB  
Article
Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal
by Matthias Frankl, Mathieu Hursin, Dimitri Rochman, Alexander Vasiliev and Hakim Ferroukhi
Appl. Sci. 2021, 11(14), 6499; https://doi.org/10.3390/app11146499 - 15 Jul 2021
Cited by 4 | Viewed by 2235
Abstract
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. [...] Read more.
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. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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10 pages, 1763 KiB  
Article
Separation of Zr from Zr-2.5Nb by Electrorefining in LiCl-KCl for Volumetric Decontamination of CANDU Pressure Tube
by Jungho Hur, Seongjin Jeong, Sungjune Sohn, Jaeyeong Park and Il Soon Hwang
Appl. Sci. 2021, 11(9), 3790; https://doi.org/10.3390/app11093790 - 22 Apr 2021
Cited by 3 | Viewed by 1758
Abstract
This study presents an experimental investigation on Zr separation from Zr-2.5Nb by anode potentiostatic electrorefining in LiCl-KCl-ZrCl4 0.5 wt. % at 773 K for irradiated CANDU pressure tube decontamination. By the ORIGEN-2 code calculation, radioactive characteristics were investigated to show that Nb-94 [...] Read more.
This study presents an experimental investigation on Zr separation from Zr-2.5Nb by anode potentiostatic electrorefining in LiCl-KCl-ZrCl4 0.5 wt. % at 773 K for irradiated CANDU pressure tube decontamination. By the ORIGEN-2 code calculation, radioactive characteristics were investigated to show that Nb-94 was the most significant radionuclide with an aspect of waste level reduction by electrorefining. Three electrorefining tests were performed by fixing the applied potential as −0.9 V (vs. Ag/AgCl 1 wt. %) at the anode to dissolve only Zr. A cathode basket was installed to collect detached deposits from the cathode. Electrorefining results showed Zr was deposited on the cathode with a small amount of Nb and other alloying elements. The chemical form of the cathode deposits was shown to be only Zr metal or a mixture of Zr metal and ZrCl, depending on the experimental conditions related to the surface area ratio of the cathode to the anode. It was determined that the Zr metal reduction at the cathode was attributed to the two-step reduction reaction of Zr4+/ZrCl and ZrCl/Zr. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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20 pages, 24344 KiB  
Article
High-Temperature Corrosion Behaviors of Structural Materials for Lead-Alloy-Cooled Fast Reactor Application
by Seung Gi Lee, Yong-Hoon Shin, Jaeyeong Park and Il Soon Hwang
Appl. Sci. 2021, 11(5), 2349; https://doi.org/10.3390/app11052349 - 06 Mar 2021
Cited by 14 | Viewed by 2601
Abstract
The corrosion of nuclear-grade steels in lead–bismuth eutectic (LBE) complicates the realization of high coolant temperatures. Corrosion tests of T91, HT9, and SS316L were performed in static cells at 600 °C for 2000 h at an oxygen level of 10−6 wt.%. The [...] Read more.
The corrosion of nuclear-grade steels in lead–bismuth eutectic (LBE) complicates the realization of high coolant temperatures. Corrosion tests of T91, HT9, and SS316L were performed in static cells at 600 °C for 2000 h at an oxygen level of 10−6 wt.%. The obtained corrosion surfaces of post-processed samples were characterized by several microscopy methods. Up to 1000 h, all the alloys exhibited an evolution of duplex oxide layers, which were spalled until 2000 h due to their increased thickness and decreased integrity. Following the spallation, a thin internal Cr-rich oxide layer was formed above the Cr-depleted zone for T91 and HT9. SS316L was penetrated by LBE down to 300 μm in severe cases. A comparison on the corrosion depths of the materials with regard to the parabolic oxidation law with abundant literature data suggests that it may lose its validity once the duplex layer is destroyed as it allows LBE to penetrate the metal substrate. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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20 pages, 4855 KiB  
Article
Criticality Analysis for BWR Spent Fuel Based on the Burnup Credit Evaluation from Full Core Simulations
by Anna Detkina, Dzianis Litskevitch, Aiden Peakman and Bruno Merk
Appl. Sci. 2021, 11(4), 1498; https://doi.org/10.3390/app11041498 - 07 Feb 2021
Cited by 3 | Viewed by 2378
Abstract
This study performed criticality analysis for the GBC-68 storage cask loaded with boiling water reactor (BWR) spent fuel at the discharged burnups obtained from the full-core simulations. The analysis was conducted for: (1) different reloading scenarios; (2) target burnups; and (3) two fuel [...] Read more.
This study performed criticality analysis for the GBC-68 storage cask loaded with boiling water reactor (BWR) spent fuel at the discharged burnups obtained from the full-core simulations. The analysis was conducted for: (1) different reloading scenarios; (2) target burnups; and (3) two fuel assembly types—GE14 and SVEA100—to estimate the impact each of the three factors has on the cask reactivity. The BWR spent fuel composition was estimated using the results of the nodal analysis for the advanced boiling water reactor (ABWR) core model developed in this study. The nodal calculations provided realistic operating data and axial burnup and coolant density profiles, for each fuel assembly in the reactor core. The estimated cask’s keff were compared with the fresh fuel and peak reactivity standards to identify the benefit of the burnup credit method applied to the BWR spent fuel at their potential discharge burnups. The analysis identified the significant cask criticality reduction from employing the burnup credit approach compared to the conventional fresh fuel approach. However, the criticality reduction was small compared to the peak reactivity approach, and could even disappear for low burnt fuel assemblies from non-optimal reloading patterns. In terms of cask manufacturing, the potential financial benefit from using the burnup credit approach was estimated to be USD 3.3 million per reactor cycle. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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15 pages, 5112 KiB  
Article
System Studies on the Fusion-Fission Hybrid Systems and Its Fuel Cycle
by Mikhail Shlenskii and Boris Kuteev
Appl. Sci. 2020, 10(23), 8417; https://doi.org/10.3390/app10238417 - 26 Nov 2020
Cited by 2 | Viewed by 2528
Abstract
This paper is devoted to applications of fusion-fission hybrid systems (FFHS) as a powerful neutron source implementing transmutation of minor actinides (MA: Np, Am, Cm) extracted from the spent nuclear fuel (SNF) of nuclear reactors. Calculations which simulated nuclide kinetics for the metallic [...] Read more.
This paper is devoted to applications of fusion-fission hybrid systems (FFHS) as a powerful neutron source implementing transmutation of minor actinides (MA: Np, Am, Cm) extracted from the spent nuclear fuel (SNF) of nuclear reactors. Calculations which simulated nuclide kinetics for the metallic fuel containing MA and neutron transport were performed for particular facilities. Three FFHS with fusion power equal to 40 MW are considered in this study: demo, pilot-industrial and industrial reactors. In addition, needs for a fleet of such reactors are assessed as well as future FFHSs’ impact on Russian Nuclear Power System. A system analysis of nuclear energy development in Russia was also performed with the participation of the FFHSs, with the help of the model created at AO “Proryv”. The quantity of MA that would be produced and transmuted in this scenario is estimated. This research shows that by the means of only one hybrid facility it is possible to reduce by 2130 the mass of MA in the Russian power system by about 28%. In the case of the absence of partitioning and transmutation of MA from SNF, 287 t of MA will accumulate in the Russian power system by 2130. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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19 pages, 5498 KiB  
Article
Burnup Credit Evaluation for BWR Spent Fuel from Full Core Calculations
by Anna Detkina, Dzianis Litskevitch, Aiden Peakman and Bruno Merk
Appl. Sci. 2020, 10(21), 7549; https://doi.org/10.3390/app10217549 - 27 Oct 2020
Cited by 3 | Viewed by 2473
Abstract
Due to the challenges of spent fuel accumulation, the nuclear industry is exploring more cost-effective solutions for spent fuel management. The burnup-credit method, in application for storage and transport of the spent fuel, gained traction over recent decades since it can remove the [...] Read more.
Due to the challenges of spent fuel accumulation, the nuclear industry is exploring more cost-effective solutions for spent fuel management. The burnup-credit method, in application for storage and transport of the spent fuel, gained traction over recent decades since it can remove the over-conservatism of the “fresh-fuel” approach. The presented research is focused on creating an innovative, best estimate approach of the burnup-credit method for boiling water reactor (BWR) spent fuel based on the results of neutronic/thermal-hydraulic coupled full core simulations. The analysis is performed using a Polaris/DYN3D sequence. Four different shuffling procedures were used to estimate the possible range of the BWR fuel discharged burnup variation. The results showed a strong influence of the shuffling procedure on the final burnup distribution. Moreover, a comparison of the 2D lattice and 3D coupled nodal approaches was conducted for the criticality estimation of single fuel assemblies. The analysis revealed substantial improvement in criticality curves obtained with the full-core model. Finally, it was shown that the benefit from the burnup-credit method is larger in the case of more optimal fuel utilisation-based shuffling procedures. The new approach developed here delivers a promising basis for future industrial optimisation procedures and thus cost optimisation. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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20 pages, 847 KiB  
Article
Countercurrent Actinide Lanthanide Separation Process (ALSEP) Demonstration Test with a Simulated PUREX Raffinate in Centrifugal Contactors on the Laboratory Scale
by Andreas Wilden, Fabian Kreft, Dimitri Schneider, Zaina Paparigas, Giuseppe Modolo, Gregg J. Lumetta, Artem V. Gelis, Jack D. Law and Andreas Geist
Appl. Sci. 2020, 10(20), 7217; https://doi.org/10.3390/app10207217 - 16 Oct 2020
Cited by 14 | Viewed by 2636
Abstract
An Actinide Lanthanide Separation Process (ALSEP) for the separation of trivalent actinides (An(III)) from simulated raffinate solution was successfully demonstrated using a 32-stage 1 cm annular centrifugal contactor setup. The ALSEP solvent was composed of a mixture of 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) [...] Read more.
An Actinide Lanthanide Separation Process (ALSEP) for the separation of trivalent actinides (An(III)) from simulated raffinate solution was successfully demonstrated using a 32-stage 1 cm annular centrifugal contactor setup. The ALSEP solvent was composed of a mixture of 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and N,N,N′,N′-tetra-(2-ethylhexyl)-diglycolamide (T2EHDGA) in n-dodecane. Flowsheet calculations and evaluation of the results were done using the Argonne’s Model for Universal Solvent Extraction (AMUSE) code using single-stage distribution data. The co-extraction of Zr(IV) and Pd(II) was prevented using CDTA (trans-1,2-diaminocyclohexane-N,N,N′,N′-tetraacetic acid) as a masking agent in the feed. For the scrubbing of co-extracted Mo; citrate-buffered acetohydroxamic acid was used. The separation of An(III) from the trivalent lanthanides (Ln(III)) was achieved using citrate-buffered diethylene-triamine-N,N,N′,N″,N″-pentaacetic acid (DTPA), and Ln(III) were efficiently back extracted using N,N,N′,N′-tetraethyl-diglycolamide (TEDGA). A clean An(III) product was obtained with a recovery of 95% americium and curium. The Ln(III) were efficiently stripped; but the Ln(III) product contained 5% of the co-stripped An(III). The carryover of Am and Cm into the Ln(III) product is attributed to too few actinide stripping stages, which was constrained by the number of centrifugal contactors available. Improved separation would be achieved by increasing the number of An strip stages. The heavier lanthanides (Pr, Nd, Sm, Eu, and Gd) and yttrium were mainly routed to the Ln product, whereas the lighter lanthanides (La and Ce) were mostly routed to the raffinate. Full article
(This article belongs to the Special Issue Nuclear Wastes Management)
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