Transmutation Of Minor Actinides Research Articles

Dissolution behavior of MgO based inert matrix fuel for the transmutation of minor actinides

This study explores the dissolution properties of magnesia-based inert matrix nuclear fuel (IMF) containing transuranium elements (TRU). Pure MgO pellets as well as MgO pellets containing CeO2, as surrogate for TRU oxides, and are considered as model systems for genuine magnesia based inert matrix fuel were fabricated. The aim of this study is to identify conditions at which the matrix material can be selectively dissolved during the head-end reprocessing step, allowing a separation of MgO from the actinides, whereas the actinides remain undissolved. The dissolution behavior was studied in macroscopic batch experiments as a function of nitric acid concentration, dissolution medium volume, temperature, stirring velocity, and pellet density (85, 90, 96, and 99%TD). To mimic pellets with various burn-ups the density of the here fabricated pellets was varied. MgO is soluble even under mild conditions (RT, 2.5 mol/L HNO3). The dissolution rates of MgO at different acid concentrations are rather similar, whereas the dissolution rate is strongly dependent on the temperature. Via a microscopic approach, a model was developed to describe the evolution of the pellet surface area during dissolution and determine a surface normalized dissolution rate. Moreover, dissolution rates of the inert matrix fuel containing CeO2 were determined as a function of the acid concentration and temperature. During the dissolution of MgO/CeO2 pellets the MgO dissolves completely, while CeO2 (>99%) remains undissolved. This study intends to provide a profound understanding of the chemical performance of magnesia based IMF containing fissile material. The feasibility of the dissolution of magnesia based IMF with nitric acid is discussed.

Investigation of Novel Composite Material Based on Extra-Heavy Concrete and Basalt Fiber for Gamma Radiation Protection Properties

The paper presents a new composite material for radiation protection based on extra-heavy concrete reinforced by basalt fiber. Basalt fiber is a new material for concrete reinforcement, which provides improved mechanical characteristics of concrete, reduces the level of microcracks and increases the durability of concrete. Within the scope of present work, the gamma-ray radiation protection properties of concrete reinforced with basalt fiber was modeled. Two types of extra-heavy concrete were used for this paper. The main gamma-ray attenuation coefficients such as mean atomic number, mean atomic mass, mean electron density, effective atomic number, effective electron density, Murty effective atomic number were analyzed with help of WinXCom software. It has been shown that the addition of basalt fiber to concrete does not impair its gamma-ray radiation shielding properties. With increasing the basalt fiber dosage in concrete, the radiation properties against gamma radiation are improved.
 This research was carried out with the financial support of the IAEA, within the terms and conditions of the Research Contract 20638 in the framework of the Coordinated Research Project (CRP) “Accelerator Driven Systems (ADS) Applications and use of Low-Enriched Uranium in ADS (T33002)” within the project “The Two-Zone Subcritical Systems with Fast and Thermal Neutron Spectra for Transmutation of Minor Actinides and Long-Lived Fission Products”.

Minor actinide transmutation characteristics in AP1000

We study the minor actinide transmutation characteristics by coating different thickness minor actinide on fuel pellets of the AP1000. The calculation results indicate that the more minor actinide is loaded to AP1000, the more boric acid concentration need to be reduced from coolant to let keff return criticality. The relationship between the reduced boric acid concentrations in AP1000 coolant with minor actinide loading amount is not linear. The reduced boric acid concentration in coolant approaches a saturation level after about 300.0 kg minor actinide loading amount. This implies that too much minor actinide loading amount in a single AP1000 fuel loading is meaningless.Our simulation results also denote that loading minor actinide to AP1000 core does affect the power distribution of AP1000 in some way, we need to reduce the boric acid concentration and rearrange the burnable poison rods in the core simultaneously to achieve the purpose of flatting the power distribution. So a moderate minor actinide loading in AP1000 will not adversely affect the core power distribution strikingly.Our calculation results also display that fuel temperature coefficient (FTC) is still negative after adding minor actinide to reactor. In comparison with the standard AP1000 reactor core, the extent of the negative feedback effect of FTC after loading minor actinides is reduced to some extent; the moderator temperature coefficient (MTC) improves the negative feedback effect of the reactor core to a great extent. Taking into account all analytical results about FTC and MTC, the total negative feedback effect on reactor power level changes is improved after loading minor actinides to AP1000 reactor for transmutation. The transmutation rate calculation results show that AP1000 reactor can incinerate 44.0 kg MA nuclides after one year fuel cycle. So AP1000 reactor can incinerate almost 2 times the annual MA yield from a typical PWR reactor.

Preliminary Analysis of High-Flux RSG-GAS to Transmute Am-241 of PWR’s Spent Fuel in Asian Region

A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

An optimization methodology for heterogeneous minor actinides transmutation

In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors

In the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%), which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit.

A comparison of curium, neptunium and americium transmutation feasibility

Minor actinides transmutation is the process of decreasing the long term radiotoxicity of the nuclear spent fuel by submitting it to a neutron flux so as to achieve fission of the heavy nuclides concerned. In the case of a closed fuel cycle, minor actinides are the main contributors to the spent fuel radiotoxicity after a few centuries. The isotopic vector of the minor actinides feed to be transmuted depends heavily on the fuel cycle considered: PWRs with UOX fuels will mainly lead to neptunium and americium production while MOX fueled reactors will produce mainly americium and curium. Americium is the main element currently considered for transmutation due to its relatively short half-life and significant production level. On the other hand, neptunium is seen as a secondary candidate for transmutation due to its very long half life and low activity while Curium transmutation is generally ruled out due to the high activity of curium isotopes. Two modes of transmutation in fast reactors are generally opposed, namely the homogeneous approach in which minor actinides are directly mixed with the fuel while in the heterogeneous approach, the minor actinides are loaded in dedicated targets. It is shown in this paper that the impacts on the fuel cycle of heterogeneous americium transmutation are similar to the one of homogeneous curium transmutation. It is further shown that given the quantities of curium in the fuel cycle, only a limited number of reactors would be required to effectively transmute the curium production of fast reactors with americium bearing blankets. Curium transmutation thus appears a feasible option in a completely closed fuel cycle without significantly higher fuel cycle impacts than with only americium transmutation. It is finally verified that neptunium transmutation can be achieved regardless of the approach considered.

Structural response of Nd-stabilized zirconia and its composite under extreme conditions of swift heavy ion irradiation

Inert matrix fuel concept for minor actinide transmutation proposes stabilized zirconia as the major component for inert matrix. The present study explores Nd-stabilized zirconia (Zr0.8Nd0.2O1.9; Nd as surrogate for Am) and its composites for radiation tolerance against fission fragments. The introduction of MgO in the composite with stabilised zirconia is performed from the point of view to enhance the thermal conductivity. The radiation damage is also compared with Nd-stabilized zirconia co-doped with Y3+ (Zr0.8Nd0.1Y0.1O1.9) in order to mimic doping of minor actinides in Y3+ containing stabilized zirconia (Nd as surrogate for Am). The compositions were synthesized by gel combustion followed by high temperature sintering and characterised by XRD, SEM and EDS. Irradiation was carried out by 120 MeV Au ions at various fluences and irradiation induced structural changes were probed by in-situ X-ray diffraction (XRD). XRD demonstrated the retention of crystallinity for all the three samples but the extent of the damage was found to be highly dependent on the nominal composition. It was observed that introduction of Y3+ along with Nd3+ to stabilize cubic zirconia imparted poorer radiation stability. On the other hand, formation of a CERCER composite of MgO with Nd-stabilised zirconia enhanced its behaviour against swift heavy ion irradiation. Investigating these compositions by XANES spectroscopy post irradiation did not show any change in local electronic structure of constituent ions.

Optimization of minor actinides bearing blankets with industrial fuel cycle constraints

Heterogeneous minor actinides transmutation is the process of loading neptunium, americium and curium in dedicated targets, generally located at the core periphery, in order to transmute these long-lived heavy nuclides into shorter-lived fission products. This has a positive effect on long-term radiotoxicity of the spent fuel and on the short to medium term decay heat of the final waste packages. However, given their location at the core periphery in a low flux zone, these targets are generally loaded with a high content in minor actinides in order to maintain acceptable transmutation performances. A significant part of the heavy nuclides produced by capture during irradiation being energetic alpha-emitters or strong neutron emitters, this leads to a very high activity in the irradiated targets which has negative impacts on the fuel cycle. In this paper, an optimization methodology of heterogeneous minor actinides transmutation is applied to the fuel cycle case in order to assess the constraints limiting transmutation. It appears that transportation of the fresh assemblies is the first limiting criterion due to the high decay heat of 241Am. Sodium washing at the end of irradiation is the second limiting criterion. It is shown that the geometrical design on the target assembly is not a concern considering the current limitations on the fuel cycle. It is also shown that using current industrial limitations and plutonium feed, it is not possible to achieve equilibrium between production of americium in the core and consumption in the blankets. It appears then that research should focus on the removal of fuel reprocessing limitations rather than standard assembly optimization.

Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

Impact of impurity in transmutation cycle on neutronics design of revised accelerator-driven system

Partitioning and transmutation technology will be a promising technology to reduce the burden of the geological disposal of the high-level radioactive waste. The Japan Atomic Energy Agency has investigated a lead-bismuth eutectic cooled accelerator-driven system (ADS) to transmute minor actinides (MAs) in the transmutation cycle. This study aims to revise the ADS design and to investigate the impact of impurities in the transmutation cycle on the ADS neutronics design.The impact of impurities in the transmutation cycle is investigated for the revised design. For the uranium from the partitioning, the accompaniment of 20wt% U against the Pu weight is acceptable although the MA transmutation amount will be decreased slightly. For the rare earth (RE) from the partitioning, the accompaniment of 5wt% RE against the MA weight is allowable. In the reprocessing, the decontamination factor, DF=10 for RE is enough from the viewpoint of the neutronics design. The impact of the fuel composition accuracy is also investigated. The uncertainty of the ZrN ratio against the MA fuel should be less than 0.2% to minimize a surplus proton beam current due to the uncertainty.Through these investigations, the required conditions for the impurities in the partitioning, the MA fuel fabrication and the reprocessing processes were clarified from the viewpoint of the ADS neutronics design.

The results of the transmutation of fission fragments in the spectrum of neutrons of thermal and fast reactors

Radioactivity of spent nuclear fuel (SNF) discharged from nuclear reactor core is determined during the first 100 years by fission fragments (FF), after that the main contribution in the SNF activity is made by actinides. Existing scenarios of SNF handling are based on the transmutation of minor actinides (MA) into fission fragments accomplished in fast reactors. Scenarios of transmutation of fission fragments in thermal and fast neutron spectra and time-dependent radiation characteristics are examined in the present study. Nuclide composition of fission fragments is taken from the results of simulation of burnup of 439GT fuel assembly (TVSA type) for VVER-1000 nuclear reactor during 3 years performed using MCU-5 software complex. The obtained data were used for determining starting nuclide composition for different cooling-down times prior to the beginning of transmutation (irradiation in neutron fluxes) to be input in the ORIGEN2 code.The following three options of irradiation of fission fragments are presented: transmutation without cooling down, cooling down fission fragments during 4 years prior to irradiation, cooling down fission fragments during 30 prior to irradiation. Duration of irradiation was selected to be equal to 3 and 15 years. Efficiency of transmutation was determined using time-dependent “transmutation factor” equal to the ratio of radioactivity of nuclides in the process of transmutation and after its completion to their radioactivity without transmutation.The calculated values of transmutation factors proved to be noticeable only during irradiation in reactor core: these values reached 5–10 and were dependent only on the duration of fission fragment cooling down prior to the beginning of transmutation. After removal of fission fragments from neutron flux transmutation factor decreased to unity within several years. After one hundred more years after irradiation in neutron spectrum of thermal reactor transmutation factor reduces to 0.8–0.5 depending on the duration of the transmutation process. Slight growth of transmutation factor to the values of 1.2–1.8 was observed after irradiation in fast reactor spectrum within the time interval of 200–1000 years and after 1000 years following this its reduction to the value of 0.9–0.7 was noted.The main conclusion is that purposeful incineration of fission fragments is senseless because only insignificant gain in radioactivity (a little less than by the factor two) is achieved after 1000 years.The indifference of fission fragments with regard to transmutation can be partially explained by the fraction of stable nuclides which increases with extension of the period of fission fragments cooling down. Upon completion of the cycle of fuel use it contains approximately 15% of stable nuclides among fission fragments, and after 30 years of cooling down the fraction of stable isotopes reaches 85%.

Minor actinides transmutation on pressurized water reactor burnable poison rods

Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core.We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is coating a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods.Through considering the loading patterns, loading amount and self-shielding effect and so on, when the outer diameter of MA is 0.520cm, the transmutation rate of MA is 38.5%, it is the optimal configuration, firstly transmutation rate and loading amount of MA with the pattern of MA in the water gap of burnable poison rods are larger than ones with the pattern of MA substitute for part burnable poison. Secondly the highest transmutation rate means the least loading amount of MA and the lowest transmutation rate means the most loading amount of MA. In order to keep reasonable transmutation rate and loading amount, when the outer diameter of MA is 0.520cm, it would not reduce dramatically the amount of MA and the MA loading of this loading pattern is equivalent to the discharge amount of five pressurized water reactor, at the same time, the transmutation rate can keep reasonable level.

Insight into the Am-O Phase Equilibria: A Thermodynamic Study Coupling High-Temperature XRD and CALPHAD Modeling.

In the frame of minor actinide transmutation, americium can be diluted in UO2 and (U, Pu)O2 fuels burned in fast neutron reactors. The first mandatory step to foresee the influence of Am on the in-reactor behavior of transmutation targets or fuel is to have fundamental knowledge of the Am-O binary system and, in particular, of the AmO2-x phase. In this study, we coupled HT-XRD (high-temperature X-ray diffraction) experiments with CALPHAD thermodynamic modeling to provide new insights into the structural properties and phase equilibria in the AmO2-x-AmO1.61+x-Am2O3 domain. Because of this approach, we were able for the first time to assess the relationships between temperature, lattice parameter, and hypostoichiometry for fcc AmO2-x. We showed the presence of a hyperstoichiometric existence domain for the bcc AmO1.61+x phase and the absence of a miscibility gap in the fcc AmO2-x phase, contrary to previous representations of the phase diagram. Finally, with the new experimental data, a new CALPHAD thermodynamic model of the Am-O system was developed, and an improved version of the phase diagram is presented.

Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Summary Partitioning and transmutation of the minor actinides (MAs) from nuclear power plants are of importance for the nuclear-energy sustainable development. Fast reactors are applied to transmutation for the hard neutron spectrum and high neutron flux. However, the safety-related neutronic parameters will become worse when large amounts of MAs are loaded. In this paper, transients of a 600 MWe sodium-cooled fast reactor for MA transmutation are analyzed by using neutron transport simulation. The control rod withdrawal transients are calculated. Two cases are compared to investigate the influence of loading MAs into the fast reactor core. One is the common core loaded with mixed oxide (MOX) fuel, and the other one is the transmutation core loaded with MOX fuel and MAs. The results indicate that in order to apply similar operation criterion with the common core, the transmutation core with 6% weight fraction of MAs should be operated with more than 30% power reduction. In addition, the results of the transport-based transient analysis and the point kinetics transient analysis are compared. There are noticeable differences between them, which indicate that the usual way based on the point kinetics calculation is not suitable well for simulating the control rod introduced transients. Copyright © 2017 John Wiley & Sons, Ltd.

Electrochemical reduction of CerMet fuels for transmutation using surrogate CeO2-Mo pellets

One of the concepts chosen for the transmutation of minor actinides in Accelerator Driven Systems or fast reactors proposes the use of fuels and targets containing minor actinides oxides embedded in an inert matrix either composed of molybdenum metal (CerMet fuel) or of ceramic magnesium oxide (CerCer fuel). Since the sufficient transmutation cannot be achieved in a single step, it requires multi-recycling of the fuel including recovery of the not transmuted minor actinides. In the present work, a pyrochemical process for treatment of Mo metal inert matrix based CerMet fuels is studied, particularly the electroreduction in molten chloride salt as a head-end step required prior the main separation process. At the initial stage, different inactive pellets simulating the fuel containing CeO2 as minor actinide surrogates were examined. The main studied parameters of the process efficiency were the porosity and composition of the pellets and the process parameters as current density and passed charge. The results indicated the feasibility of the process, gave insight into its limiting parameters and defined the parameters for the future experiment on minor actinide containing material.

Effects of Am and Cm distributions on neutronic parameters of MINERVE reactor

Minor actinide recycling and transmutation are important stages in the advanced nuclear systems design such as Generation IV reactors that offers a high level of safety, waste reduction, and minimizes its environmental and radiological impacts. In the present work, two preliminary model configurations are used to examine the effects of adding two long-lived isotopes of minor actinides (243Am and 247Cm) in MINERVE reactor reference core (R1-UO2) that represents a LWR loaded with UO2 fuel rods. The effects are studied in two models of uniform and nonuniform minor actinides distributions, and include the numerical evaluation of neutronic parameters such as neutron spectrum, flux distribution and fission rate, using MCNPX transport code. The analysis of and comparison between the uniform and nonuniform models have been discussed to evaluate which of either is expected to be favored for minor actinide transmutation and energy production.

ZrO2-NdO1.5 system: Investigations of phase relation and thermophysical properties

Exploring the structural and thermophysical behaviour of Nd-Substituted ZrO2 is vital in context of matrices for minor actinide transmutation. Hence, phase relations have been studied in this system using high temperature sintered samples prepared by gel combustion route. Phase characterization of Zr1−xNdxO2−x/2 (0.0≤x≤1.0) series was carried out using X-ray diffraction (XRD), Raman spectroscopy, scanning electron microscopy (SEM) and energy-dispersive spectroscopy (EDS). With increasing NdO1.5 content, the system evolved in the sequence of (a) monoclinic zirconia (b) mixture of monoclinic+cubic stabilized zirconia, (c) cubic stabilized zirconia (d) mixture of stabilized zirconia+pyrochlore (e) pyrochlore (f) mixture of pyrochlore+hexagonal Nd2O3 and (g) pure hexagonal Nd2O3. Lattice and bulk thermal expansion of few selected compositions were characterized by high temperature XRD (298–1273K) and dilatometry (323–1673K). Average coefficient of thermal expansion decreased with increasing NdO1.5 fraction in Zr1−xNdxO2−x/2 series (x=0.2, 0.3, 0.4, 0.5) solid solutions, which has been attributed to increasing order in cation sub-lattice resulting in overall increase in lattice stability. Thermal conductivity has been reported for a representative composition Zr0.8Nd0.2O1.90.

Optimization of Minor Actinide–Bearing Radial Blankets for Heterogeneous Transmutation in Fast Reactors

A methodology dedicated to the optimization of the transmutation of minor actinides (MAs) in dedicated blankets is discussed here. This methodology relies on recently developed optimization tools. In the so-called heterogeneous transmutation approach, MAs are loaded into specific assemblies located at the periphery of a fast reactor core. Thus, the resulting perturbation of the core behavior is limited and the management of MAs is entirely decoupled from standard fuel management. This also allows greater flexibility in the blanket design, in terms of material, volume fraction, and neutron spectrum to be used. On the other hand, the low neutron flux level experienced at the periphery of the core slows down the transmutation process. If this effect can be compensated for by an increase of the MA fraction loaded in the blankets, this also strongly increases their decay heat and neutron source level, which complicates spent fuel reprocessing and handling. An optimization is carried out with regard to the neutron spectrum and americium concentration in the blankets, with the dual objective of maximizing the transmuted MA mass while minimizing the total MA inventory in the fuel cycle by limiting the cooling time of such blankets. Artificial neural networks are coupled with a genetic algorithm to reduce the total calculation time. It is shown here that regardless of the MA mass to be loaded, a slightly moderated neutron spectrum is the most promising option for heterogeneous transmutation. This result is confirmed by full-core calculations. An analysis of the irradiation time is also performed, and it is shown that maximization of the irradiation time should be sought in the specific case studied here. It is concluded that from a purely physical point of view, no breakthrough can be obtained for heterogeneous transmutation.

Reactor core transient analysis of an innovative high-level nuclear waste transmuter with metal fuel

For the sustainable development of nuclear energy, reducing high-level nuclear wastes in spent fuels discharged from commercial nuclear power plants is very important. Partitioning and transmutation of minor actinides from these spent fuels are critical for the effective disposal of high-level wastes. However, transmutation in current operational reactors is not efficient and brings some safety problems. Therefore, a dedicated accelerator-driven subcritical transmuter named highly efficient industrial transmuter (HEIT) was newly proposed to solve the problems. This system utilizes the uranium-free metallic dispersion fuel and is of high power density. This paper focuses on the transient analysis of HEIT to prove that it is feasible for the future nuclear industry. Temperatures, cladding stresses, and cumulative creep damage fractions are analyzed. Furthermore, the burnup dependence is also investigated. Three typical transients including the unprotected loss of flow, the beam overpower (BOP), and the unprotected transient overpower are evaluated. Numerical results conclude that in the reactivity insertion transient and the BOP transient, this new system exhibits excellently in safety. However, the shut-down system is required in flow loss case which lasts more than half an hour. The safety margin changes with core depletion. For unprotected loss of flow, the end of lifetime case is the most important one to care. While for BOP and unprotected transient overpower, both the beginning and end of lifetime should be accommodated. The conclusions will be useful in the future design of HEIT and similar systems. Copyright © 2017 John Wiley & Sons, Ltd.