Developments in Fabrication of Annular MOX Fuel Pellet for Indian Fast Reactor

Total Page:16

File Type:pdf, Size:1020Kb

Developments in Fabrication of Annular MOX Fuel Pellet for Indian Fast Reactor Developments in fabrication of annular MOX fuel pellet for Indian fast reactor A.K. Mishra1,#, B.K. Shelke1, M.K. Yadav1, Mohd. Afzal1, Arun Kumar2, G.J. Prasad2 1Advanced Fuel Fabrication Facility, 2Nuclear Fuels Group, Bhabha Atomic Research Centre, Tarapur- 401 502, India. #Email address: [email protected] Abstract Mechanical rotary presses along with adoption of core rod feature were inducted for fabrication of intricate annular Mixed Oxide (MOX) pellets for Prototype Fast Breeder Reactor (PFBR). In the existing tooling, bottom plungers contain core rod whereas top plungers contain a central hole for the entry of core rod during compaction. Frequent manual clean up of top plungers after few operations were required due to settling of powder in the annular hole of top plungers during compaction. Delay in cleaning can also result in breakage of tooling apart from increase in the dose to extremities of personnel. New design of tooling has been introduced to clean up the top plungers online during the operation of rotary press. It leads to increase in the productivity, reduces the spillage of valuable nuclear material and also reduces man-rem to operators significantly. The present paper describes the modification in tooling design and compaction sequence established for online cleaning of top plungers. ________________________________________________________________________________________________________ International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), 4–7 March 2013, Paris, France 1. Introduction Indian Nuclear Power Programme is based on closed nuclear fuel cycle for efficient utilization of its nuclear resources. This strategy also enables waste classification and minimizes long-lived waste disposal problem. The three stage nuclear programmes envisage indigenous PHWRs in the first stage, Fast Breeder Reactors in the second stage and thorium utilization in the third stage. Fast reactors are essential for India not only for its contribution to nuclear power but for extension of its modest resources. Commissioning of Fast Breeder Test Reactor (FBTR) at Kalpakkam represents our first attempt in this direction. FBTR was fuelled with Mark I mixed carbide fuel of composition (U0.3,Pu0.7)C for the initial core which was followed by Mark II fuel of (U0.45,Pu0.55)C for the extended core. The mixed carbide fuel has performed exceedingly well and its burn-up has exceeded 1, 00,000 MWd/Te without any fuel failure. (U-Pu) MOX fuel was selected as the driver fuel for our Prototype Fast Breeder Reactor (PFBR- 500) because of our good technology base for (U, Pu) MOX fuel manufacture for thermal reactors as well as industrial scale reprocessing experience with oxide fuels of PHWRs. The world-wide experience also indicates (U, Pu) MOX fuel has high burn-up potential, excellent safety response and overall acceptability at a higher level in all parts the fuel cycle. Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur has been fabricating MOX fuel of various type for thermal and fast reactors such as Boiling Water Reactors (BWRs), Pressurised Heavy Water Reactors (PHWRs) and Fast Breeder Test Reactor (FBTR). Presently, AFFF has taken up responsibility for fabrication of annular MOX fuel for first core of Prototype Fast Breeder Reactor (PFBR). The MOX fuel is being fabricated through powder metallurgical route involving cold compaction and sintering. Annular MOX fuel pellets of two compositions i.e. 21% and 28% PuO2 are being fabricated for PFBR. A large number of (U, Pu)O2 MOX pellets have to be fabricated for the first core of PFBR. The fabrication of MOX fuel is carried out in α-leak tight glove boxes line to comply with safety requirements. It creates lot of constraints like space restriction, accessibility, maintainability and handling of material during the fabrication of plutonium based mixed oxide fuel. Hence, consideration of all aspects is required prior to selection of equipment [1]. Fabrication of annular MOX pellets with outer diameter 5.55 mm and annuls of about 1.8 mm with inner diametrical tolerance of less than + 0.20 mm without grinding an inner surface is a challenging task. Fabrication of annular pellet demands incorporation of additional tooling i.e. core rod assembly that provides relative motion with respect to bottom plunger. Initially, hydraulic press was inducted with modified tooling for fabrication of intricate annular pellet [2]. The major restrictions in using hydraulic press for final compaction of annular PFBR MOX pellet was complex tooling design due to requirement of core rod, complicated tool changing and maintenance. Hence, it is generally preferred to adopt a press having easy core rod adoption facility. Rotary press has been inducted for final compaction of MOX pellet over hydraulic press due its high rate of production, simple design of tooling, easy way of tool changing, easy adoption of core rod feature and no chance of Oil leakage. This paper deals with technology developed for fabrication of annular MOX pellet. It also describes modification carried out in tooling design and compaction sequence established for instantaneous online cleaning of top plunger during operation of rotary press. ________________________________________________________________________________________________________ International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), 4–7 March 2013, Paris, France 2. Fabrication The flowsheet for fabrication of MOX fuel fabrication along with the quality control checks at different process steps is shown in Fig. 1. Uranium dioxide and plutonium dioxide powder along with organic binder as Polyethylene Glycol and lubricant as Oleic acid were first weighed and then milled together in an indigenously developed attritor to get the required enrichment and homogeneity. Proper choice of binder and lubricant has a considerable significance in achieving desired specification of pellets [3-5]. The milled material was pre-compacted and the pre-compacts were granulated in the size range of 500-1200 µm. These granules were used in final compaction to make the green pellets. Final compaction of MOX annular pellets was carried out in a rotary press. Sintering of green pellets was done in a batch o type resistance heating furnace under reducing atmosphere (mixed N2-7% H2 gas) at 1600 C for 4-6 h. Oversize pellets were ground to acceptable size by a dry centerless grinder. Physical and chemical specifications of sintered pellets for PFBR are given in Table 1. The inspected pellets were loaded into clad tubes for encapsulation. Table 1 Physical and Chemical Specification of PFBR MOX fuel. Outer Diameter of pellet 5.55 ± 0.05 mm Inner diameter of pellet 1.8 ± 0.2 mm Length of pellet 7.0 mm (nominal) Linear Mass of pellet 2.25 ± 0.15 g/cm PuO2 enrichment (nominal) 21 ± 1% & 28 ± 1 % Equivalent Hydrogen Content < 3 ppm Total Concentration of Impurities < 5000 ppm ________________________________________________________________________________________________________ International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), 4–7 March 2013, Paris, France U, O/M, S.A Pu, Am, Isotopes, UO2 PuO2 Impurities Impurities ATTRITOR MILLING NWCC PRECOMPACTION & GRANULATION FINAL COMPACTION GLM, DIA, HEIGHT SINTERING SLM, DIA, H, (U/Pu), Metallography CENTERLESS GRINDING VISUAL DIA CHECK INSPECTION REJECTS H, O/M, F, Cl, VACCUM DEGASSING N , Metallic 2 impurities STACK MAKING STACK REJECTS INSPECTION & RADIOMETRY PELLET LOADING Weld chemistry, FUEL ELEMENT WELDING Metallography DECONTAMINATION Visual, He leak, DECLADDING X-ray REJECT radiography, γ PELLETS autoradiograph WIRE WRAPPING Assembly inspection, Contaminatio PACKING AND TRANSPORT n check Fig.1 Flowsheet for fabrication of MOX fuel. ________________________________________________________________________________________________________ International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), 4–7 March 2013, Paris, France 3. Annular pellet fabrication technique for PFBR Depending on the feed material, die fill, final compact shape and dimension, compacting pressure, the type of press, the tooling is designed. Various techniques for the fabrication of annular pellet and the sequence of operation are available in literature [6]. Technology for fabrication of annular pellet using indigenous rotary press has been developed at Advanced Fuel Fabrication Facility (AFFF). Annular pellet fabrication demands incorporation of additional tooling i.e. core rod assembly that provides relative motion with respect to bottom plunger as compared to solid pellet fabrication. Core rod assembly is consists of core rod and core rod holder. Core rod is prone to failure due to misalignment or incomplete die filling, which causes eccentric loading. It is also subjected to wear during the operation. Hence proper selection of core rod material should be done to increase the life of the tooling. The core rod with sufficient toughness can be used to overcome the non- uniform stresses developed during compaction and should also be capable of resisting the ejection loads imposed. Core rod dimension play vital role in achieving inner diameter specification, since it is not easy grind inner surface of an annular pellet. 3.1 Constraint during fabrication of PFBR annular pellets 1. All the fabrication steps had to be carried out
Recommended publications
  • Advanced Nuclear Power and Fuel Cycle Technologies: Outlook and Policy Options
    Order Code RL34579 Advanced Nuclear Power and Fuel Cycle Technologies: Outlook and Policy Options July 11, 2008 Mark Holt Specialist in Energy Policy Resources, Science, and Industry Division Advanced Nuclear Power and Fuel Cycle Technologies: Outlook and Policy Options Summary Current U.S. nuclear energy policy focuses on the near-term construction of improved versions of existing nuclear power plants. All of today’s U.S. nuclear plants are light water reactors (LWRs), which are cooled by ordinary water. Under current policy, the highly radioactive spent nuclear fuel from LWRs is to be permanently disposed of in a deep underground repository. The Bush Administration is also promoting an aggressive U.S. effort to move beyond LWR technology into advanced reactors and fuel cycles. Specifically, the Global Nuclear Energy Partnership (GNEP), under the Department of Energy (DOE) is developing advanced reprocessing (or recycling) technologies to extract plutonium and uranium from spent nuclear fuel, as well as an advanced reactor that could fully destroy long-lived radioactive isotopes. DOE’s Generation IV Nuclear Energy Systems Initiative is developing other advanced reactor technologies that could be safer than LWRs and produce high-temperature heat to make hydrogen. DOE’s advanced nuclear technology programs date back to the early years of the Atomic Energy Commission in the 1940s and 1950s. In particular, it was widely believed that breeder reactors — designed to produce maximum amounts of plutonium from natural uranium — would be necessary for providing sufficient fuel for a large commercial nuclear power industry. Early research was also conducted on a wide variety of other power reactor concepts, some of which are still under active consideration.
    [Show full text]
  • MOX Fuel Program: Current Plans and Controversy
    MOX Fuel Program: Current Plans and Controversy The Mixed Oxide (MOX) Fuel Fabrication Facility at Savannah River, South Carolina is intended to manufacture nuclear fuel from surplus weapons-grade plutonium for use in commercial nuclear energy reactors. However, the project has faced serious delays and massive cost overruns – and currently has no customers for its proposed fuel. As a result, the President’s FY17 Budget Proposal requests $270 million to begin closing the project, while diluting the plutonium for transfer to the Waste Isolation Pilot Plant (WIPP) in New Mexico, a more cost-efficient option. What Is It? The MOX facility at Savannah River was designed to repurpose 3.5 tonnes of surplus weapons-grade plutonium yearly. This facility was intended to play a key role in the United States’ fulfillment of the 2000 Plutonium Management and Disposition Agreement (PMDA) between Russia and the U.S., which affirms each country’s commitment to dispose of 34 metric tonnes of plutonium, enough collectively for 17,000 nuclear weapons. Challenges: The MOX Fuel Fabrication Facility’s anticipated date of operation was 2007, with plutonium disposition set to end in 2020. Multiple delays in construction led to significant cost overruns, with beginning operations delayed until 2019. Initially valued at $2.898 billion (2016 dollars), the total cost of the project skyrocketed to $15.683 billion as a result of construction delays and program mismanagement. This estimate, however, assumes a steady rate of funding, and fluctuations in funding levels could exacerbate delays and cost overruns. Even if completed, the site currently boasts zero customers for MOX fuel.
    [Show full text]
  • Partitioning and Transmutation Annual Report 2009
    R-10-07 Partitioning and transmutation Annual report 2009 Emma Aneheim, Christian Ekberg, Anna Fermvik, Mark Foreman, Elin Löfström-Engdahl, Teodora Retegan, Gunnar Skarnemark, Irena Špendlíková Nuclear Chemistry Department of Chemical and Biological Engineering Chalmers University of Technology January 2010 Svensk Kärnbränslehantering AB Swedish Nuclear Fuel and Waste Management Co Box 250, SE-101 24 Stockholm Phone +46 8 459 84 00 R-10-07 CM Gruppen AB, Bromma, 2010 ISSN 1402-3091 Tänd ett lager: SKB Rapport R-10-07 P, R eller TR. Partitioning and transmutation Annual report 2009 Emma Aneheim, Christian Ekberg, Anna Fermvik, Mark Foreman, Elin Löfström-Engdahl, Teodora Retegan, Gunnar Skarnemark, Irena Špendlíková Nuclear Chemistry Department of Chemical and Biological Engineering Chalmers University of Technology January 2010 This report concerns a study which was conducted for SKB. The conclusions and viewpoints presented in the report are those of the authors. SKB may draw modified conclusions, based on additional literature sources and/or expert opinions. A pdf version of this document can be downloaded from www.skb.se. Abstract The long-lived elements in the spent nuclear fuels are mostly actinides, some fission products (79Se, 87Rb, 99Tc, 107Pd, 126Sn, 129I and 135Cs) and activation products (14C, 36Cl, 59Ni, 93Zr, 94Nb). To be able to destroy the long-lived elements in a transmutation process they must be separated from the rest of the spent nuclear fuel for different reasons. One being high neutron capture cross- sections for some elements, like the lanthanides. Other reasons may be the unintentional production of other long lived isotopes. The most difficult separations to make are those between different actinides but also between trivalent actinides and lanthanides, due to their relatively similar chemical properties.
    [Show full text]
  • Management of Reprocessed Uranium Current Status and Future Prospects
    IAEA-TECDOC-1529 Management of Reprocessed Uranium Current Status and Future Prospects February 2007 IAEA-TECDOC-1529 Management of Reprocessed Uranium Current Status and Future Prospects February 2007 The originating Section of this publication in the IAEA was: Nuclear Fuel Cycle and Materials Section International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria MANAGEMENT OF REPROCESSED URANIUM IAEA, VIENNA, 2007 IAEA-TECDOC-1529 ISBN 92–0–114506–3 ISSN 1011–4289 © IAEA, 2007 Printed by the IAEA in Austria February 2007 FOREWORD The International Atomic Energy Agency is giving continuous attention to the collection, analysis and exchange of information on issues of back-end of the nuclear fuel cycle, an important part of the nuclear fuel cycle. Reprocessing of spent fuel arising from nuclear power production is one of the strategies for the back end of the fuel cycle. As a major fraction of spent fuel is made up of uranium, chemical reprocessing of spent fuel would leave behind large quantities of separated uranium which is designated as reprocessed uranium (RepU). Reprocessing of spent fuel could form a crucial part of future fuel cycle methodologies, which currently aim to separate and recover plutonium and minor actinides. The use of reprocessed uranium (RepU) and plutonium reduces the overall environmental impact of the entire fuel cycle. Environmental considerations will be important in determining the future growth of nuclear energy. It should be emphasized that the recycling of fissile materials not only reduces the toxicity and volumes of waste from the back end of the fuel cycle; it also reduces requirements for fresh milling and mill tailings.
    [Show full text]
  • Appendix B Differences Between Thermal and Radiation Properties of Mixed Oxide and Low-Enriched Uranium Radioactive Materials
    APPENDIX B DIFFERENCES BETWEEN THERMAL AND RADIATION PROPERTIES OF MIXED OXIDE AND LOW-ENRICHED URANIUM RADIOACTIVE MATERIALS The contents considered in this Standard Review Plan (SRP) appendix are unirradiated mixed oxide (MOX) radioactive material (RAM), in the form of powder, pellets, fresh fuel rods, or fresh reactor fuel assemblies. Unirradiated MOX RAM will also be referred to in this appendix as MOX fresh fuel. This appendix summarizes the relative degree of differences between the thermal and radiation properties of the various MOX RAM contents relative to similar properties for analogous low-enriched uranium (LEU) RAM contents. MOX fresh fuel can be made with plutonium having various compositions of plutonium isotopes. The discussion in this appendix makes use of the 3013 Standard (DOE 2012), which specifies the typical grades of plutonium that are used to make the MOX fresh fuel. The actual plutonium compositions found in practice may not match these compositions exactly, but these grades can be considered typical for the purposes of this appendix. Table B–6 of the 3013 Standard gives weight percents for various plutonium isotopes in various grades of plutonium. They are reproduced in the following table (Table B–1) as representative values for typical grades of plutonium that might be used to fabricate MOX fresh fuel. Pure plutonium-239 has been included to contrast the effect of the other plutonium isotopes. Note that in addition to the isotopes identified in Table B–1, plutonium will contain plutonium-236 and americium-241 (from plutonium-241 decay). Initially, it is expected that MOX fresh fuel will be fabricated using weapons grade (WG) plutonium.
    [Show full text]
  • Framatome ANP MOX Fuel Design
    BAW-10238(NP) Revision 0 43-10238NP-00 MOX Fuel Design Report March 2002 Framatome ANP, Inc. U.S. Nuclear Regulatory Commission Report Disclaimer Important Notice Regarding the Contents and Use of This Document Please Read Carefully This technical report was derived through research and development programs sponsored by Framatome ANP, Inc. It is being submitted by Framatome ANP, Inc. to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Framatome ANP, Inc. fabricated reload fuel or technical services provided by Framatome ANP, Inc. for light water power reactors and it is true and correct to the best of Framatome ANP, Inc.'s knowledge, information, and belief. The information contained herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report and, under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Framatome ANP, Inc. in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations. Framatome ANP, Inc.'s warranties and representations concerning the subject matter of this document are those set forth in the agreement between Framatome ANP, Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Framatome ANP, Inc. nor any person acting on its behalf: a. makes any warranty, or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or b.
    [Show full text]
  • A Review of the Nuclear Fuel Cycle Strategies and the Spent Nuclear Fuel Management Technologies
    energies Review A Review of the Nuclear Fuel Cycle Strategies and the Spent Nuclear Fuel Management Technologies Laura Rodríguez-Penalonga * ID and B. Yolanda Moratilla Soria ID Cátedra Rafael Mariño de Nuevas Tecnologías Energéticas, Universidad Pontificia Comillas, 28015 Madrid, Spain; [email protected] * Correspondence: [email protected]; Tel.: +34-91-542-2800 (ext. 2481) Received: 19 June 2017; Accepted: 6 August 2017; Published: 21 August 2017 Abstract: Nuclear power has been questioned almost since its beginnings and one of the major issues concerning its social acceptability around the world is nuclear waste management. In recent years, these issues have led to a rise in public opposition in some countries and, thus, nuclear energy has been facing even more challenges. However, continuous efforts in R&D (research and development) are resulting in new spent nuclear fuel (SNF) management technologies that might be the pathway towards helping the environment and the sustainability of nuclear energy. Thus, reprocessing and recycling of SNF could be one of the key points to improve the social acceptability of nuclear energy. Therefore, the purpose of this paper is to review the state of the nuclear waste management technologies, its evolution through time and the future advanced techniques that are currently under research, in order to obtain a global vision of the nuclear fuel cycle strategies available, their advantages and disadvantages, and their expected evolution in the future. Keywords: nuclear energy; nuclear waste management; reprocessing; recycling 1. Introduction Nuclear energy is a mature technology that has been developing and improving since its beginnings in the 1940s. However, the fear of nuclear power has always existed and, for the last two decades, there has been a general discussion around the world about the future of nuclear power [1,2].
    [Show full text]
  • Nuclear Fuel Cycles Technology Assessment
    Clean Power Quadrennial Technology Review 2015 Chapter 4: Advancing Clean Electric Power Technologies Technology Assessments Advanced Plant Technologies Biopower Carbon Dioxide Capture and Storage Clean Power Value-Added Options Carbon Dioxide Capture for Natural Gas and Industrial Applications Carbon Dioxide Capture Technologies Carbon Dioxide Storage Technologies Crosscutting Technologies in Carbon Dioxide Capture and Storage Fast-spectrum Reactors Geothermal Power High Temperature Reactors Hybrid Nuclear-Renewable Energy Systems Hydropower Light Water Reactors Marine and Hydrokinetic Power Nuclear Fuel Cycles Solar Power Stationary Fuel Cells Supercritical Carbon Dioxide Brayton Cycle U.S. DEPARTMENT OF Wind Power ENERGY Clean Power Quadrennial Technology Review 2015 Nuclear Fuel Cycles Chapter 4: Technology Assessments Introduction and Background The Nuclear Fuel Cycle (NFC) is defined as the total set of operations required to produce fission energy and manage the associated nuclear materials. It can have different attributes, including the extension of natural resources, or the minimization of waste disposal requirements. The NFC, as depicted in Figure 4.O.1, is comprised of a set of operations that include the extraction of uranium (U) resources from the earth (and possibly from seawater), uranium enrichment and fuel fabrication, use of the fuel in reactors, interim storage of used nuclear fuel, the optional recycle of the used fuel, and the final disposition of used fuel and waste forms from the recycling processes. Thorium (Th) fuel cycles have been proposed also, but have not been commercially implemented). Figure 4.O.1 Schematic of the uranium based Nuclear Fuel Cycle 1 Quadrennial Technology Review 2015 Clean Power TA 4.O: Nuclear Fuel Cycles The nuclear fuel cycle is often grouped into three classical components (front-end, reactor, and back-end): Front End: The focus of the front end of the nuclear fuel cycle is to deliver fabricated fuel to the reactor.
    [Show full text]
  • Status of MOX Fuel Fabrication, Utilization, XA0201760 Design and Performance in Lwrs
    Status of MOX fuel fabrication, utilization, XA0201760 design and performance in LWRs V. Onoufriev International Atomic Energy Agency, Vienna Abstract. The overview looks at the present status of Pu recycle as MOX fuel in thermal power reactors as presented mainly at the IAEA Symposium on MOX Fuel Cycle Technologies for Medium and Long Term Deployment held in Vienna in May 1999, and some other international meetings. Present status in MOX fuel fabrication, performance and utilization technologies are discussed. Recycling of plutonium as MOX fuel in LWRs has become a mature industry. The technology is well understood and the facilities, institutions and procedures are in place (capacity extensions being planned) to meet the anticipated arisings of separated plutonium from the production of nuclear power. MOX fuel performance and its modelling are discussed in comparison with those of UO2 fuel. 1. INTRODUCTION Today the plutonium recycling with LWRs has evolved to industrial level in some countries relying on the recycling policy. This happened mainly because of the worldwide delays in development of fast reactor's programmes. More than 30 years of reactor experience using MOX fuel as well as the fabrication of 2 000 MOX assemblies with the use of 85 t of Pu separated from spent fuel from power reactors and good performance records indicate now that the recycling of plutonium as MOX fuel in LWRs has become a mature industry. The technology is well understood and the facilities, institutions and procedures are in place to meet the anticipated arisings of separated plutonium from the production of nuclear power. The number of countries engaged on plutonium recycling could be increasing in the near future, aiming the reduction of stockpiles of separated plutonium from earlier and existing reprocessing contracts.
    [Show full text]
  • Immobilized BTBP/Btphen Ligands
    Extraction of minor actinides, lanthanides and other fission products by silica- immobilized BTBP/BTPhen ligands Article Accepted Version Afsar, A., Distler, P., Harwood, L. M., John, J. and Westwood, J. (2017) Extraction of minor actinides, lanthanides and other fission products by silica-immobilized BTBP/BTPhen ligands. Chemical Communications, 53 (28). pp. 4010-4013. ISSN 1359-7345 doi: https://doi.org/10.1039/c7cc01286a Available at http://centaur.reading.ac.uk/69930/ It is advisable to refer to the publisher’s version if you intend to cite from the work. See Guidance on citing . To link to this article DOI: http://dx.doi.org/10.1039/c7cc01286a Publisher: The Royal Society of Chemistry All outputs in CentAUR are protected by Intellectual Property Rights law, including copyright law. Copyright and IPR is retained by the creators or other copyright holders. Terms and conditions for use of this material are defined in the End User Agreement . www.reading.ac.uk/centaur CentAUR Central Archive at the University of Reading Reading’s research outputs online Please do not adjust margins ChemComm COMMUNICATION Extraction of minor actinides, lanthanides and other fission products by silica-immobilized BTBP/BTPhen ligands† a b a b a Received 00th January 20XX, Ashfaq Afsar, Petr Distler, Laurence M. Harwood, * Jan John and James Westwood Accepted 00th January 20XX DOI: 10.1039/x0xx00000x www.rsc.org/ Novel BTBP [bis-(1,2,4-triazin-3-yl)-2,2’-bipyridine] / BTPhen products such as Ni(II), Pd(II), Ag(I) and Cd(II), complicating the 17 [bis-(1,2,4-triazin-3-yl)-1,10-phenanthroline] functionalized silica separation of trivalent actinides for transmutation.
    [Show full text]
  • The Fuel Cycle Implications of Nuclear Process Heat
    energies Article The Fuel Cycle Implications of Nuclear Process Heat Aiden Peakman 1,2,∗ and Robert Gregg 1 1 National Nuclear Laboratory, Chadwick House, Warrington WA3 6AE, UK; [email protected] 2 School of Engineering, University of Liverpool, Liverpool L69 3GH, UK * Correspondence: [email protected] Received: 13 August 2020; Accepted: 11 November 2020; Published: 20 November 2020 Abstract: International and UK fuel cycle scenario analyses performed to date have focused on nuclear plants producing electricity without considering in detail the other potential drivers for nuclear power, such as industrial process heat. Part of the reason behind the restricted applications of nuclear power is because the assumptions behind the future scenario are not fully captured, for example how big are demands from different sectors? Here we present a means to fully capture the potential opportunities for nuclear power using Sankey diagrams and then, using this information, consider for the first time in the UK the fuel cycle implications of decarbonising industrial heat demand in the year 2050 with nuclear power using the ORION fuel cycle code to study attributes related to spent fuel, uranium demand and decay heat from the spent fuel. We show that even in high industrial energy demand scenarios, the sensitivity of spent fuel masses and decay heat to the types of reactor deployed is relatively small compared to the greater fuel cycle demands from large-scale deployment of nuclear plants for electricity production. However, the sensitivity of spent fuel volumes depends heavily on the extent to which High Temperature Reactor and Light Water Reactor systems operating on a once-through cycle are deployed.
    [Show full text]
  • Influence of Dose Rate on the Radiolytic Stability of a BTBP Solvent For
    Radiochim. Acta 97, 319–324 (2009) / DOI 10.1524/ract.2009.1615 © by Oldenbourg Wissenschaftsverlag, München Influence of dose rate on the radiolytic stability of a BTBP solvent for actinide(III)/lanthanide(III) separation By A. Fermvik1,∗, C. Ekberg1,2, S. Englund3,M.R.S.J.Foreman1,2, G. Modolo4, T. Retegan1 and G. Skarnemark1 1 Nuclear Chemistry, Department of Chemical and Biological Engineering, Chalmers University, 41296 Gothenburg, Sweden 2 Industrial Materials Recycling, Department of Chemical and Biological Engineering, Chalmers University, 41296 Gothenburg, Sweden 3 OKG AB, 57283 Oskarhamn, Sweden 4 Forschungszentrum Jülich GmbH, 52425 Jülich, Germany (Received May 1, 2008; accepted in revised form December 19, 2008) Irradiation / Dose rate / Solvent extraction / Partitioning / a number of criterions such as being soluble in desired dilu- Actinides / Lanthanides / Distribution ratio / ents, chemically stable, preferably completely incinerable Separation factor (CHON-principle) [5], resistant towards hydrolysis and re- sistant towards radiolysis, a criterion that will be discussed in this paper. Summary. The recently developed ligand MF2-BTBP dis- A large number of heterocyclic N-donor SANEX ex- solved in cyclohexanone is a promising solvent for the group tractants have been developed in the European research separation of trivalent actinides(III) from the lanthanides(III). projects [6]. Unfortunately, most of them were only able to Its high stability against nitric acid has been demonstrated re- extract trivalent actinides from solutions of very low acid- cently. Since the solvent is also exposed to a continuously high ity and suffer from hydrolytic and radiolytic instability. One radiation level in the counter current process, the radiolytic stability of the solvent was examined in this study.
    [Show full text]