Mu*STAR ADSR Fuel Conversion Facility Evaluation and Cost Analysis

The use of depleted uranium dioxide (DUO2) particulates as fill material for repository waste packages (WPs) containing light-water reactor (LWR) spent nuclear fuel (SNF) was investigated. A repository WP would be loaded with SNF, and small DUO2 particulates (0.5 to 1.0 mm) would be added to fill the void space inside the WP - including the coolant channels inside SNF assemblies. The DUO2 fill slows release of radionuclides from the SNF by (a) creating a local chemically reducing environment that slows degradation of the SNF UO2 and (b) reducing groundwater flow through the WP. The depleted uranium (DU) fill minimizes the potential for long-term criticality in the repository by isotopic dilution of 233U and 235U. The potential for criticality is primarily determined by 235U (a) originally in the SNF and (b) from decay of 239Pu. The use of DU consumes excess DU from the production of enriched uranium. The mechanisms for improvements in repository performance with DUO2 fill are defined, but additional work is required to fully quantify the benefits and costs of such an approach.

The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium.

In Belgium 57% of the electricity is presently generated by 7 nuclear units of the PWR type located in Doel and Tihange. Their total output amounts to 5632 MWe. Part of the spent fuel unloaded from the first three units has been sent till 2000 for reprocessing in the Cogema facility at La Hague. As the reprocessing of the spent fuel produced by the last four units is not covered by the contracts concluded with Cogema, Synatom, the Belgian utilities’ subsidiary in charge of the front- and back-end of the nuclear fuel cycle for all PWR reactors in Belgium, decided to study the possible solutions for a temporary storage of this spent fuel. End of 1993, the Belgian government decided that reprocessing (closed cycle) and direct disposal (open cycle) of spent fuel had to be considered as equal options in the back-end policy for nuclear fuel in Belgium. The resolution further allowed continued execution of a running reprocessing contract (from 1978) and use of the corresponding Pu for MOX in Belgian NPP’s, but requested a reprocessing contract concluded in 1990 (for reprocessing services after 2000) not to be executed during a five-year period. During this period priority was to be given to studies on the once-through cycle as an option for spent fuel management. Figure 1 is a chart showing the two alternatives for the spent fuel cycle in Belgium. In this context, Synatom entrusted Belgatom to develop a dedicated flask (called “bottle”) for direct disposal of spent fuel, to perform a design study of an appropriate encapsulation process and to prepare a preliminary feasibility study of a complete spent fuel conditioning plant. Meanwhile preparation works were made for the construction of an interim storage facility on both NPP sites of Doel and Tihange in order to meet increasing storage capacity needs. For selecting the type of interim storage facility, Belgatom performed a technical-economical analysis. Considerations of design and safety criteria as well as flexibility, reversibility, technical constraints, global economical aspects and construction time led to adopt dry storage with dual purpose casks (in operation since end 1995) for the Doel site and wet storage in a modular pool for the Tihange site (in operation since 1997). In parallel, ONRAF/NIRAS, the Belgian Agency for the management of radioactive waste and enriched fissile materials and the Belgian nuclear research centre, SCK•CEN, conduct underground investigations in view of geological disposal. The paper describes the methodology that Belgatom has developed to provide the utilities with appropriate solutions (reracking, dry storage in casks, wet storage in ponds, etc.) and how Belgatom demonstrated also the feasibility of spent fuel conditioning with a view to direct disposal in clay layers. The spent fuel storage facilities in operation in Belgium, designed and built by Belgatom, are then briefly presented.

This report was generated jointly by Argonne National Laboratory (ANL) and Idaho National Laboratory to provide a high-level summary of the current knowledge on the behavior of fission product iodine during reprocessing of used nuclear fuel, as well as provide recommended path forward for research and development activities to fill particular knowledge gaps. The focus of this work is on pyrochemical processing as is applied to light water reactor (LWR) oxide-based used nuclear fuels (UNF), however, some discussion of electrorefiner behavior from metal fuel processing equipment is also included.

Study of XeF2 fluorination potential against Rh2O3, RuO2, ZrO2, and U3O8 for use in reactive gas recycle of used nuclear fuel

The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

The technology demonstration element of the Global Nuclear Energy Partnership (GNEP) program is aimed at demonstrating the closure of the fuel cycle by destroying the transuranic (TRU) elements separated from spent nuclear fuel (SNF). Multiple recycle through fast reactors is used for burning the TRU initially separated from light-water reactor (LWR) spent nuclear fuel. For the initial technology demonstration, the preferred option to demonstrate the closed fuel cycle destruction of TRU materials is a sodium-cooled fast reactor (FR) used as burner reactor. The sodium-cooled fast reactor represents the most mature sodium reactor technology available today. This report provides a review of the current state of development of fuel systems relevant to the sodium-cooled fast reactor. This report also provides a review of research and development of TRU-metal alloy and TRU-oxide composition fuels. Experiments providing data supporting the understanding of minor actinide (MA)-bearing fuel systems are summarized and referenced.

This work adapts both prismatic and pebble-bed micro high-temperature gas-cooled reactor (HTGR) point designs for Pu/transuranic (TRU) driver fuel under a continuous-recycle fuel cycle. This article indicates the potential for HTGR microreactors to be deployed as burner reactors in continuous-recycle fuel cycles. The adapted prismatic and pebble-bed micro-HTGR point designs are optimized to maximize fuel discharge burnup to reduce nuclear waste and environmental impact. Optimization was performed for inert matrix fuel (IMF) concepts, each using different composite moderators in the prismatic and pebble-bed designs. The prismatic and pebble-bed designs use TRU oxide tristructural-isotropic (TRISO) fuel entrained in a magnesium oxide (MgO) ceramic compact. The moderator materials considered include beryllium, beryllium oxide, yttrium hydride, and zirconium hydride entrained in the same MgO host matrix. For each prismatic IMF concept, optimization studies were performed to maximize discharge burnup by varying the TRISO packing fraction and assembly lattice pitch. For each pebble-bed IMF concept, optimization studies also varied the TRISO packing fraction while further considering variables that included the pebble fueled radius, the ratio of fueled to unfueled pebbles, and the active reactor core radius. Energy-normalized metrics, such as the amount of spent nuclear fuel (SNF) and waste, activity, and environmental impacts, are reported for the continuous-recycle prismatic and pebble-bed IMF concepts. These metric results are compared to those of the graphite reference design, once-through prismatic and pebble-bed IMF concepts, a light water reactor (LWR), and a small modular reactor (SMR). All the continuous-recycle IMF concepts are shown to outperform the graphite reference case for all the evaluated metrics. Additionally, all the continuous-recycle IMF concepts outperformed their once-through microreactor counterparts, the LWR, and the SMR for most of the evaluated metrics. For the mass of SNF and high-level waste disposed, the continuous-recycle designs saw a 92.7% reduction with a concomitant reduction in the volume of low-level waste disposed between a 29.7% to 31.1% reduction. In addition to the levelized metrics, the levelized cost of fuel (LCOF) was calculated for the continuous-recycle prismatic and pebble-bed microreactor designs. The LCOF analysis was performed using cost drivers, uncertainties, and potential for cost reduction developed by employing deterministic modeling, sensitivity analyses, and Monte Carlo simulations. Significant reductions were seen in the LCOF, particularly in the pebble-bed concepts. Compared to the graphite reference, the prismatic designs saw between a 72.6% and 76.5% reduction and the pebble-bed designs saw about a 77.8% reduction in the LCOF. Overall, the continuous-recycle IMF concepts significantly reduced nuclear waste, environmental impact, and LCOF compared to conventional once-through reactor fuel cycles. It is important to note that the cases in this paper are representative of the maximum theoretically possible performance in these systems, and that significant research and development would be required to realize these potential benefits.

An Integrated Process Strategy has been developed to package, condition, transport, and store in an interim storage facility the spent nuclear fuel (SNF) currently residing in the K-Basins at Hanford. Information required to support the development of the condition process and to support the safety analyses must be obtained from characterization testing activities conducted on fuel samples from the Basins. Some of the information obtained in the testing was reported in PNL-10778, K-Basin Spent Nuclear Fuel Characterization Data Report (Abrefah et al. 1995). That report focused on the physical, dimensional, metallographic examinations of the first K-West (KW) Basin SNF element to be examined in the Postirradiation Testing Laboratory (PTL) hot cells; it also described some of the initial SNF conditioning tests. This second of the series of data reports covers the subsequent series of SNF tests on the first fuel element. These tests included optical microscopy analyses, conditioning (drying and oxidation) tests, ignition tests, and hydrogen content tests.

The Power Reactor Innovative Small Module (PRISM) designed by GE Hitachi Nuclear Energy is a small, modular, sodium-cooled fast reactor. The PRISM core is located in a pool-type containment vessel and is fueled with metallic fuel. Each PRISM produces 311 MW of electricity. The PRISM is inherently safe due to its negative power reactivity feedback, large in-vessel coolant inventory, passive heat removal systems, below-grade siting, and atmospheric reactor vessel operating pressure. In NUREG-1368, “Preapplication Safety Evaluation Report for the Power Reactor Innovative Small Module (PRISM) Liquid-Metal Reactor,” the U.S. Nuclear Regulatory Commission stated that “On the basis of the review performed, the staff, with the ACRS [Advisory Committee on Reactor Safeguards] in agreement, concludes that no obvious impediments to licensing the PRISM design have been identified.” PRISM is able to fission electrometallurgically recycled used nuclear fuel (UNF) from light water reactors as well as weapons-grade materials. PRISM, with the associated Nuclear Fuel Recycling Center, represents a safe, diversion resistant, commercially viable technology for recycling UNF with a small modular reactor.

Click to increase image sizeClick to decrease image size Additional informationNotes on contributorsCharles W. BagnalCharles W. Bagnal Jr. (BS, nuclear engineering, Georgia Institute of Technology, 1980) is a consulting engineer at GRP Consulting, Inc., currently engaged in reactor physics methods development and nuclear engineering support services. He has been involved in nuclear reactor design analysis and software development for six years. His experience includes advanced light water reactor (LWR) and heavy water reactor designs, including extended burnup fuel cycles using gadolinia burnable poison.Gerard P. CavanaughGerard P. Cavanaugh (BS, physics, Massachusetts Institute of Technology, 1969; MS, 1970, and PhD, 1976, nuclear engineering, University of Illinois, Urbana-Champaign) worked from 1973 to 1975 in the Neutronics Physics Division at Oak Ridge National Laboratory. He is currently supervisor of radiation physics and criticality analysis at Combustion Engineering, Inc. (C-E). His primary technical interests are in the areas of radiation transport, stochastic sampling methodologies, and computer applications ranging from mainframes to personal computers.Robert P. HarrisRobert P. Harris (BS, 1974, and MS, 1975, nuclear engineering, and MS, 1984, metallurgy, Rensselaer Polytechnic Institute) has worked in the core design, safety analysis, and licensing of C-E pressurized water reactors since 1973. He is currently involved in the neutronic design of advanced burnable poisons and fuel assemblies. He consults in the areas of core fuel management, safety analysis, and startup physics testing.Regis A. MatzieRegis A. Matzie (BS, physics, U.S. Naval Academy, 1965; MS, 1971, and PhD, 1976, nuclear engineering, Stanford University) is currently the manager of analog plants in the nuclear engineering department at C-E. His current technical interests include advanced LWR designs, methods of improving fuel utilization and fuel cycle costs in current design LWRs, and alternate fuel cycles for advanced converter reactors.Laszlo B. TarkoLaszlo B. Tarko (BS, physics, University of Connecticut, 1980) is currently the lead engineer in the radiation physics group at C-E. His current technical interests include radiation transport, ex-core radiation physics, dose rates, energy deposition, and associated computer techniques.Howard OckenHoward Ocken (D. Eng., metallurgy, Yale University, 1966) is a project manager in the Nuclear Power Division of the Electric Power Research Institute. He has managed projects that addressed materials performance in nuclear core components. His current interests include materials aspects associated with reducing radiation fields, especially replacing cobalt sources and decontamination and preconditioning technology.Per Hedemann JensenPer Hedemann Jensen (BSc, electronic engineering, Technical University, Aarhus, Denmark, 1970) is a senior health physicist at Ris0 National Laboratory, where he is head of the Section for Applied Health Physics. He has been involved in studies of radiological consequences from hypothetical core-melt accidents at nuclear power plants. His current technical interest is experimental work on internal dosimetry problems.Ian J. HastingsIan J. Hastings (PhD, metallurgical science, University of Queensland, 1968) is head of the Fuel Properties and Behaviour Group in the Fuel Materials Branch at the Chalk River Nuclear Laboratories (CRNL) of Atomic Energy of Canada Limited. His current interests are in oxide fuel behavior under irradiation, particularly release of short-lived fission products under normal and accident conditions, and in fusion ceramics.Elio MizzanElio Mizzan (BA Sc, chemical engineering, University of Toronto, 1949) is a supervisor at the Fuel Materials Branch Hot Cell Facility at CRNL. His interests have been in the areas of postirradiation examination of reactor fuels and materials, particularly postdefect handling of spent U02 fuel.Alan M. RossAlan M. Ross (BSc, general, University of London, 1953) is a supervisor at the Recycle Fuel Fabrication Laboratories of the Fuel Materials Branch of CRNL. His interests include the fabrication, nondestructive assay, properties, irradiation behavior, and neutron radiography of current and advanced thermal reactor fuels.John R. KelmJohn R. Kelm is a research technician in the Fuel Materials Branch at CRNL. His research responsibility is the operation of a special project hot cell. His current interest is U02 oxidation and stress corrosion cracking.Real J. ChenierReal J. Chenier is a research technician in the Fuel Materials Branch at CRNL. His current research responsibility is operation of the postirradiation metallographic facilities.D. H. RoseD. H. Rose is a research technologist in the Fuel Properties and Behaviour Group of the Fuel Materials Branch at CRNL. His current responsibility is fabrication and operation of instrumented in-reactor experiments.J. NovakJ. Novak (BASc, engineering science, University of Toronto, 1974) is employed in Central Nuclear Services, Nuclear Generation Division of Ontario Hydro. Since 1976 he has been engaged in the areas of nuclear fuel production and development. His current technical interests include behavior of irradiated fuel in air, U02 powder characteristics, and fuel performance improvement.Claudio RonchiClaudio Ronchi (Dr. rer. nat. phys., State University, Milan, Italy, 1965) is a research officer at the Joint Research Centre (JRC) of the Commission of European Communities. After having worked in the research and development of advanced fuels for fast breeders, he is now engaged in reactor safety studies.Jacques van de LaarJacques van de Laar (Ingenieur Fysische Techniek, H.T.S., Heerlen, The Netherlands, 1974) works in thec mathematical modeling department of JRC-Karlsruhe. He is currently involved in the development and execution of computer programs for reactor fuel performance and safety analysis.Hubert BlankHubert Blank (PhD, metal physics, Technical University of Stuttgart, 1957), head of the Physics Division of the European Institute for Transuranium Elements, Karlsruhe, has directed the “Swelling of Advanced Fuels” project at this institute since 1973.Tetsuo FukasawaTetsuo Fukasawa (BS, 1976; MS, 1978; and Dr. Eng., 1981, nuclear engineering, Tohoku University) is a researcher in the Energy Research Laboratory (ERL), Hitachi, Ltd. He has specialized in the radiochemistry of actinide elements and is currently working in the field of low-level radioactive waste treatment and spent fuel reprocessing.Koichi ChinoKoichi Chino (BS, 1972, and MS, 1974, mechanical engineering, Tokyo Institute of Technology) is a researcher at ERL. His primary areas of interest are transport phenomena and radioactive waste management.Osamu KuriyamaOsamu Kuriyama (BS, 1973, and MS, 1975, Tohoku University) is a researcher at ERL. His current interests include the radioactive waste management and chemistry of nuclear waste.Fumio KawamuraFumio Kawamura (BS, chemical engineering, Gunma University,1970; MS, 1972, and Dr. Eng., 1976, Tohoku University) is a researcher at ERL where he is involved in radioactive waste management and reactor water chemistry.Hideo YusaHideo Yusa (BS, physics, Tohoku University, 1959; Dr. Eng., Osaka University, 1969) is a chief researcher at ERL. He is responsible for the research and development of radioactive waste management systems.Bernard L. CohenBernard L. Cohen (BS, Case Institute of Technology, 1944; MS, University of Pittsburgh, 1948; DSc, Carnegie Institute of Technology, 1950) did basic research on nuclear structure using accelerators at Oak Ridge National Laboratory from 1950 to 1958 and at the University of Pittsburgh until the mid-1970s. He then turned his research efforts to the environmental impacts of energy generation, including studies of radioactive waste, health effects of radiation, risk and risk aversion, radon problems, and reactor safety. He is a professor of physics at the University of Pittsburgh and a former director (1965 to 1978) of its Scaife Nuclear Laboratories.K. TasakaK. Tasaka (PhD, nuclear engineering, University of Tokyo, 1976) first worked in breeder reactor safety and fission product characteristics research for ten years and in light water reactor (LWR) safety research for nine years at the Japan Atomic Energy Research Institute (JAERI). He is the project leader of the Rig of Safety Assessment (ROSA) program, and his current interests include analysis of thermal-hydraulic behavior during a loss-of-coolant accident (LOCA) and an anticipated transient without scram in LWRs.M. SuzukiM. Suzuki (MS, mechanical engineering, Kyoto University, 1971) has worked for the ROSA program since 1974. His current interests include safety evaluation of an LWR.Y. AnodaY. Anoda (PhD, mechanical engineering, University of Tokyo, 1979) is a research engineer for the ROSA program. His current interests include two-phase flow during a LOCA.Y. KoizumiY. Koizumi (PhD, mechanical engineering, University of Tokyo, 1977) is a research engineer for the ROSA program. His current interests include analysis of thermal-hydraulic behavior during a LOCA with emphasis on two-phase flow characteristics and heat transfer.T. YonomotoT. Yonomoto (MS, nuclear engineering, University of Osaka, 1982) is a research engineer for the ROSA program. His current interests include analysis of thermal-hydraulic behavior during a LOCA.H. KumamaruH. Kumamaru (PhD, nuclear engineering, University of Tokyo, 1980) is a research engineer for the ROSA program. His current interests include core heat transfer under LOCA conditions.H. NakamuraH. Nakamura (MS, crystalline material engineering, Nagoya University, 1981) is a research engineer for the ROSA program. His current interests include analysis of thermal-hydraulic behavior during a LOCA.M. ShibaM. Shiba (MS, mechanical engineering, Waseda University, 1959) is the general manager of Reactor Safety Laboratory 1 at JAERI. He has worked for 22 years at JAERI in the field of reactor engineering and safety and is currently responsible for simulated LOCA experiments.Genichi MatsumotoGenichi Matsumoto (BS, physics, Tokyo University, 1946) has been an associate professor of nuclear engineering at Nagoya University from 1963 to the present, where he is responsible for research and education in nuclear reactor engineering. He worked as a member of the Scientific Research Institute of Tokyo following his graduation from Tokyo University, and from 1950 to 1963, he served on the technical staff of the Research Reactor Operation Division of the Japan Atomic Energy Research Institute. His current research field includes nuclear safety, neutron radiography (especially neutron television technology), and heat pipe technology.Soichi DoiSoichi Doi (BS, 1976, and MS, 1978, nuclear engineering, Nagoya University) has been employed by the Mitsubishi Atomic Power Industry Co. since 1978, where he is responsible for the nuclear fuel engineering of light water reactors.Kohei OhkuboKohei Ohkubo (BS, mechanical engineering, Aichi Institute of Technology, 1975) is a staff member of Nagoya University.Yasushi IkedaYasushi Ikeda (BS, physics, 1963, and PhD, nuclear engineering, 1982, Nagoya University) has been an assistant staff member of Nagoya University from 1964 to the present. He has investigated thermodynamics of nuclear materials using mass spectrometric technology; neutron radiography technology is also included in his research interests.Günther HesselGünther Hessel [BS, physics, Technical University Dresden, German Democratic Republic (GDR), 1970] is a research scientist at the Central Institute of Nuclear Research (CIN), Rossendorf. He is involved in experimental investigations and developments for technical diagnosis, especially at nuclear power plants.Hans-Erich KöppenHans-Erich Köppen (BS, computer science, Technical University Dresden, 1973) is a research scientist at the CIN. His area of expertise is microprocessor hard- and software and the construction of autonomously working monitors.Peter LiewersPeter Liewers (BS, physics, University of Halle, GDR, 1957; PhD, physics, University of Leipzig, GDR, 1962) is the scientific leader of the group for reactor diagnostics at the CIN. His interests and activities have shifted from reactor physics to technical diagnostics.Peter SchumannPeter Schumann (BS, nuclear engineering, Technical University Dresden, 1963; PhD, physics, Bergakademie Freiberg, GDR, 1970) is a staff scientist at the CIN where he is concerned with digital signal processing and computer software. Formerly he worked in the field of in-core measuring techniques and neutron spectrometryFrank-Peter WeiβFrank-Peter Weiβ (BS, physics, Technical University Dresden, 1973) is a research scientist in the group for technical diagnostics at the CIN. His current interest is the application of statistical pattern recognition to technical diagnostics.

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. The back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

Actinide breeding and reactivity variation in a thermal spectrum ADSR – Part 2: Enhanced sustainability in the thermal spectrum

Kazakhstan’s BN-350 fast reactor was shut down in 1999 and is in the process of being decommissioned in preparation for Safestore. A key achievement during 2010 was the removal of the known inventory of spent nuclear fuel (SNF) from the reactor to off-site secure storage. As a complementary activity, surveys of areas of the fuel discharge route where it was considered possible for fuel pins or fuel residues to have collected over many years of operation were also arranged to confirm that no significant amounts of fuel of remained at the plant. This paper reports on the remote radiation and visual (video and still photograph) surveys undertaken on the following areas of the BN-350 spent fuel route: • The waste storage repository or vault underneath the Post-Irradiation Examination Hot Cell in which non-destructive and destructive examination of irradiated fuels was undertaken throughout the operating life of the reactor. • The fuel transfer and washing cells within which irradiated fuel sub-assemblies were processed and residual sodium coolant removed. • The fuel storage ponds. Man access to several areas (particularly the waste vault) was not possible due to very high radiation levels from stored βγ-active wastes or residual contamination and in these cases specially engineered remote camera and radiation detector deployment systems were developed and used. In other areas, such as the ponds, limited man access was possible under prepared and controlled conditions. The survey results, together with associated radiochemical and radiation dose rate analysis, demonstrated that, despite several recorded handling incidents during Hot Cell operations, the maximum estimated amounts of nuclear material which could remain at BN-350 were sufficiently low to give no major safeguards concerns. The data also provides key information to guide decommissioning and dismantling planning in the future.