Sodium Void Reactivity Research Articles

Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

Abstract An analysis of the influence of addition of minor actinides (MA) to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL) state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW) decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B4C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes.In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly.The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

Decomposition analysis of the sodium void reactivity of the Korean sodium-cooled fast reactor

Abstract To cope with increasing spent fuel disposals and limited domestic spent fuel storages in Korea, the Korea Atomic Energy Research Institute has developed an advanced sodium-cooled fast reactor for TRU transmutation with an electricity output of 600 MWe (called the KALIMER-600 TRU burner). The design philosophy of the KALIMER-600 TRU burner concept is highly focused on inherent safety mechanisms, i.e., passive responses to abnormal and emergency conditions, and thereby minimizes the need for engineered safety systems. Accordingly, the main concern is on the sodium coolant void reactivity, a very important safety parameter of the KALIMER-600 TRU burner. This study was therefore performed to analyze the sodium void reactivity of the KALIMER-600 TRU burner to the finest resolution possible, e.g., contributions from any isotope in each core region. Such detailed analysis could be valuable and applicable to further optimization of the core passive safety characteristics against severe coolant voiding accident conditions.

Design Studies of a Low Sodium Void Reactivity Core Able to Accommodate Degraded TRU Fuel

As part of the Fast Reactor Cycle Technology Development (FaCT) Project, JSFR (Japan Sodium-Cooled Fast Reactor) core design efforts have been made to cope with the transuranic (TRU) fuel compositions expected during the light water reactor (LWR)–to–fast breeder reactor transition period, during which various kinds of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and their recycling method. The sodium void reactivity, which is one of the major core safety parameters, is considerably influenced by TRU fuel compositions. The criteria assigned to the JSFR core include a void reactivity effect limited to ∼6 $; therefore, designing a core with reduced sodium void reactivity will offer a greater margin for the core to host TRU fuel.To this end, a new core concept called BUMPY is proposed. This homogeneous core exhibits a low sodium void reactivity, due to partial-length fuels with an upper sodium plenum interspersed within the core, among other standard fuel assemblies. This core configuration enhances the upward and lateral neutron leakage from the core fuel region toward the sodium plenum when voiding to reduce void reactivity. The BUMPY core is applied to the 750-MW(electric) JSFR core design. The core can meet the design target by adjusting the loading fraction of the partial-length fuels and the height of the step in fuel lengths. The calculated void reactivity of the selected BUMPY core is 2.5 $ (25% loading fraction, 30-cm step height), which is considerably reduced from the 5.3 $ value of the reference core. This allows the BUMPY core to accommodate 5% to 9% more minor actinides in the core compared to the reference core.

Comparison and sensitivity analysis of the core characteristics of a sodium-cooled fast breeder reactor with 750 MWe output evaluated by JENDL-4.0 and ADJ2000R

Core characteristics of a sodium-cooled fast breeder reactor (FBR) with 750 MWe output using highly decontaminated uranium and plutonium and highly minor-actinide-containing compositions were evaluated using the fast reactor cross-section set generated by the new Japanese nuclear data library JENDL-4.0. The core characteristics were compared with those obtained using the unified cross-section set ADJ2000R in order to investigate the differences between both the results. The effects on the core characteristics caused by the differences in the nuclear data of important reactions and nuclides in the cross-section sets were analyzed by a burnup sensitivity analysis. It was confirmed that adopting JENDL-4.0 to the FBR core design improves the breeding ratio, the burnup reactivity, and the reactivity control balance, because of the differences in the capture cross-sections of U-238 and Pu-239 of both the libraries. The difference in the sodium void reactivity evaluated with both the libraries was less than 1% because the increase caused by the differences in the elastic scattering cross-sections of sodium, the inelastic scattering cross-section, and the μ-average value of U-238 was practically cancelled out by the decrease caused by the differences in the capture cross-sections of Pu-239, the inelastic scattering cross-section of iron, and the capture cross-sections of Am-241.

Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

Trade-Off Study on the Power Capacity of a Prototype Sodium-Cooled Fast Reactor in Korea

The major roles of a prototype sodium-cooled fast reactor (SFR) planned to be developed at the Korea Atomic Energy Research Institute are (a) to provide an irradiation test capability for fuel and structural materials and (b) to obtain operational experience on the systems and components. The power level of the prototype SFR should be large enough to provide an appropriate irradiation test environment. Trade-off studies were therefore performed from a neutronics viewpoint to determine the power level. Specifically, core designs were performed for power levels of 125, 250, 400, and 500 MW(thermal). The selected core performance and economic efficiency indices became insensitive to the power at [approximately]400 to 500 MW(thermal) and sharply deteriorated at [approximately]125 to 250 MW(thermal) with decreasing core sizes. For the fuel management scheme, the transuranic (TRU) core performance compared with that of the uranium core, and the sodium void reactivity, were also evaluated with increasing power levels. It was found that increasing the number of batches shows a higher-burnup performance and economic efficiency. However, increasing the cycle length resulted in a lower economic efficiency. The irradiation performance of TRU and enriched TRU cores was improved by [approximately]20% and 50%, respectively. A maximum sodium void reactivity of 5.2 $ was confirmed as less than the design limit of 7.5 $. As a conclusion of our entire study, the power capacity of the prototype SFR should not be <250 MW(thermal), and would be appropriate at [approximately]500 MW(thermal) considering the performance and economic efficiency.

An analysis on the breeding capability and safety related parameters of advanced fast reactor fuels using recent cross-section set

This study focuses on the evaluation of breeding capability as well as safety related neutronic parameters of advanced fast reactor fuels which comprises of fissile–fertile combination of metal, oxide, carbide and nitride, using the recent neutron cross-section set ENDF/B-VI.7. Sodium cooled fast breeder reactor similar to prototype Fast Breeder Reactor (PFBR) is used to evaluate the performance of various fuel types involving fissile isotopes of 233U and Pu and fertile isotopes of Th and 238U. The analysis is restricted to a comparison of neutronic parameters of a fresh core and does not take into account the effects of burnup and fission products. The breeding potential of the fuels are also compared with European cross-section set JEFF-3.1. The breeding ratio of advanced fuels evaluated with ENDF/B-VI.7 and JEFF-3.1 was found to be in good agreement. From this study, it is found that Th–233U combination for almost all fuel types with the present geometry and composition gives a lower breeding ratio value. Safety neutronic parameters such as effective delayed neutron fraction, Doppler defect and sodium void reactivity were also computed. In terms of breeding potential and safety neutronic parameters, the performance of Th–Pu system especially the metal fuel type can be a better option for future large fast reactors. The large negative Doppler feedback along with a negative sodium void reactivity for metal and hybrid combinations of Th–233U system makes it an attractive fuel cycle option even though there is a penalty over its breeding capability.

Inherent safety aspects of metal fuelled FBR

Static and dynamic studies of metal fuelled fast breeder reactors (MFBR) are carried out to verify the passive shutdown capability and its inherent safety parameters. Static calculations are carried out to determine the vested reactivity feedback parameters from the fuel and coolant temperature rise separately. Power reactivity decrement of metal fuel reactor is found to be small as compared to oxide fuel reactor of same size.ULOF analysis of metal (U–Pu–6% Zr) 1000MWe pool type MFBR is studied with a flow halving time of 8s. The study is also made with considering uncertainties on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behaviour of 1000MWe core is benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal System (SGDHRS) capacity before the initiation of boiling. From the study, it is concluded that if the sodium void reactivity is limited (4.6$) then the inherent safety of 1000MWe design is assured, even with 20% uncertainty on the sensitive parameters and also it is found out higher primary pump flow halving time (15s instead of 8s) can avoid cliff edge effects in 1000MWe MFBR transient behaviour.

Fast breeder reactor core concept for heterogeneous minor actinide loading

As an alternative method to the homogeneous minor actinide (MA) recycling in fast breeder reactors, a heterogeneous MA loading core concept using a highly concentrated americium (Am)-containing fuel (Am target) is proposed. By the use of an extraction process for Am and curium (Cm) in the reprocessing of the spent fuel, Am (and a small amount of Cm) can be recovered and then concentrated to produce the target. The Am content in the heavy metal is assumed to range from 10 to 20 wt% in accordance with the target development scope. A mixed oxide fuel that contains uranium, plutonium, and neptunium is chosen as the base material of the target, so that the targets can generate a level of power equivalent to that of the driver fuels. It was found that a ring-shaped arrangement of Am targets between the inner and outer core regions exhibits a favorable MA transmutation performance without any significant deterioration in the core neutronic characteristics, including increases of the burnup reactivity and sodium void reactivity worth, and decreases of the breeding ratio and absolute value of the Doppler coefficient, etc., in comparison with those of a reference homogeneous MA loading case. It should be noted that the Am targets in this loading arrangement can contribute to the suppression of the core power distribution change along with burnup. A series of core designs, including core neutronics, thermal hydraulics, and fuel integrity evaluations, was also carried out for a representative Am target loading case. The results indicate that it is possible to design an Am target subassembly that can cope with the issues presented by highly concentrated Am, i.e., the deterioration of thermophysical properties and the accumulation of helium gas inside the target fuel pins. Therefore, the design feasibility of the heterogeneous target loading core has been enhanced.

A neutronics study for improving the safety and performance parameters of a 3600 MWth Sodium-cooled Fast Reactor

The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many performance advantages, but has one dominating neutronics drawback – a positive sodium void reactivity. The starting point for the present study is an SFR core design considered in the Collaborative Project on the European Sodium-cooled Fast Reactor (CP-ESFR). The aim is to analyze, for this reference core, four safety and performance parameters from the viewpoint of four different optimization options, and to propose possible optimized core designs. In doing so, the study focuses not only on the beginning-of-life state of the core, but also on the beginning of equilibrium closed fuel cycle.The four studied optimization options are: (a) introducing an upper sodium plenum and boron layer, (b) varying the core height-to-diameter (H/D) ratio, (c) introducing moderator pins into the fuel assembly, and (d) modifying the initial plutonium content. The sensitivity of the void reactivity, Doppler constant, nominal reactivity and breeding gain has been evaluated. In particular, the void reactivity, which is the most crucial safety parameter for the SFR, has been decomposed into its reaction-wise, isotope-wise and energy-group-wise components using a methodology based on the neutron balance equation.Extended voiding in the upper sodium plenum region – in conjunction with the effect of a boron layer introduced above the plenum – is found to be particularly effective in the void effect reduction while, at the same time, having almost no impact on the other considered parameters. A lower H/D ratio can also reduce void reactivity, but normally corresponds to worse neutron economy and, as such, leads to a less positive breeding gain and a smaller nominal reactivity margin. The Doppler constant, which is very sensitive to the neutron spectrum, can be significantly improved by the spectrum softening in the moderator pins case. Initial plutonium content effectively influences the nominal reactivity and breeding gain at the beginning-of-life, without significant changes of the safety related parameters.Finally, two different “synthesis core” concepts are proposed for further study, the common boundary condition set being to have a positive nominal reactivity margin at the end of equilibrium closed fuel cycle. Their considered safety and performance parameters at different fuel cycle states are presented in the end.

Correlations among FBR core characteristics for various fuel compositions

In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations.

Benchmark Calculations of Sodium-Void Experiments with Uranium Fuels at the Fast Critical Assembly FCA

The benchmark specification for the sodium-void reactivity experiments is given. The detailed benchmark calcu- lations for the new integral data were performed by using a continuous-energy Monte Carlo code (MVP) with use of the evaluated nuclear data libraries JENDL-3.2, -3.3, -4.0, ENDF/B-VII.0 and JEFF-3.1. The MVP calculations were performed with 2,000,000,000 particles. The statistical uncertainties of the MVP calculations were within several percents of the measured sodium-void reactivities. The ratios of calculated to experimental (C/E) values of sodi- um-void reactivities with respect to JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1 are less than those with respect to JENDL-3.2 and -4.0. The analysis results are similar to those of sodium-void reactivities previously obtained at the BFS facility which is another fast critical assembly in Russia. The benchmark calculations demonstrate the improve- ment of the reliability of the integral data such as the new integral data obtained at the FCA and the previously obtained data in the BFS and the usefulness of the new integral data for the validation of the re-evaluated cross sec- tion of 235 U. In addition, the benchmark tests with various correction factors by a deterministic calculation system are summarized for the help to validate the re-evaluated capture cross section of 235 U.

TRU burning by dual tier system of LWR-SFR

This paper presents a concept of the dual tier system consisting of the existing light water reactor (LWR) plants and sodium-cooled fast reactor (SFR) for transuranics (TRU) burning for the purpose of downsizing the required SFR. In this system, Pu is combusted by the LWR at first and then the remaining Np, Am, and Pu are destructed by the SFR. The iteration number of Pu combustion by the LWR is chosen to be twice owing to the sodium void reactivity limitation of $6. As a result of combustion calculation, the twice Pu burning of LWR lessens the TRU amount by 27% and changes the composition significantly. Moderator pins of zirconium hydride are deployed to the SFR fuel subassembly so as to enhance TRU burning and reduce the sodium void reactivity. The nuclear calculation found that the core characteristics become similar to the conventional SFR due to the moderator: the sodium void reactivity remains still $4 and the Doppler coefficient becomes −6 × 10 −3 Tdk/dT. This study concludes that this dual tier strategy can downsize the required SFR to approximately 40% of the single tier system of SFR with TRU conversion ratio of 0.6.

Studies on Influence of Sodium Void Reactivity Effect on the Concept of the Core and Safety of Advanced Fast Reactor

The paper is devoted to studies on the influence of the sodium void reactivity effect (SVRE) on the safety and technical and economical characteristics of the BN-1200-type reactor. Different core options are considered for application to this reactor. These core options differ in design, dimensions, and, hence, SVRE value. It is shown by the analysis that the most flattened core with sodium plenum at the top assures reactor self-protection under beyond-design-basis accident conditions. Sodium plenum abandonment and core height increase causing an SVRE value increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Self-protection means the possibility to avoid rapid core meltdown under conditions of the above-listed beyond-design accidents. The possibility of controlling beyond-design accidents (for instance, by restoring the power supply of the main pumps in a rather short time) is taken into account. Issues of choosing the optimal core design under these conditions are discussed.

Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback – a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case – the voiding of both inner and outer fuel zones – has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly by the influence of the sodium removal on the neutron spectrum, and this has been confirmed by the perturbation method. The neutron balance based decomposition has also shown that the prime consequence of the sodium voiding is the reduction of the absorption rate (especially of 238U capture), which provides the strongest positive component of the void reactivity. The changes in neutron production and leakage provide smaller and negative contributions. Sensitivity analysis results for individual isotopic reactions have been found to be, in a relative sense, consistent with the neutron balance based isotope-wise decomposition. Finally, the neutron balance method has been applied to analyze the void reactivity differences between the three different SFR fuel cycle states considered, viz. BOL, BOC and BEC.

Development of analysis method for sodium void reactivity of step type FBR cores by using group-wise Monte Carlo code

A new method is proposed to separate the sodium void reactivity of step type FBR cores to components including non-leakage terms and a leakage term by using a newly developed perturbation code MCPERT where fluxes and adjoint fluxes are derived from a group-wise Monte Carlo code. The step type FBR core is a core where the height of the inner core is smaller than that of the outer core and a large sodium plenum region is located above the core so as to decrease the sodium void reactivity. The conventional diffusion perturbation method cannot treat such a large void region due to the diffusion approximation, while the Monte Carlo code can treat it exactly. In this study, a group-wise Monte Carlo code GMVP with a 70-group constant set JFS-3-J3.3 is employed to evaluate the neutron fluxes and adjoint fluxes which are used as inputs to the MCPERT code to evaluate the non-leakage terms. The leakage term is derived from the difference of the total sodium void reactivity evaluated by the direct calculation of GMVP and the summation of the non-leakage terms. It is found that the proposed method can provide the result approximately consistent to the ratio of the reactivity components derived from the conventional method.

Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500MWe and 1000MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500MWe and 1000MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000MWe design is assured, even with 20% uncertainty on the sensitive parameters.

An Axially Heterogeneous Sodium-Cooled Fast Reactor Designed to Transmute Minor Actinides

An axially heterogeneous sodium-cooled fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting 241Am into 238Pu. This 238Pu is then co-recycled with the spent driver fuel to make new driver fuel. Because 238Pu is significantly more fissionable than 241Am in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because the 241Am neutron capture worth is significantly greater in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap that recovers some of the axial leakage lost by the active core.A low transuranic conversion ratio is achieved by a degree of core flattening that increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Contrary to a homogeneous design, concentrating minor actinides (MAs) in an axial blanket mitigates the problem of above-threshold multiplication during sodium voiding. Because minor actinides are irradiated only once in the axial target region, elemental partitioning of the minor actinides from plutonium is not required. This fact enables the use of metal targets with pyroprocessing. After reprocessing, the target’s newly bred 238Pu and remaining unburned MAs become the feedstock for the next batch of driver fuel.