Actinide Inventory Research Articles

Physical feasibility of minor actinides transmutation in a two-component nuclear energy system in Russia

A transition to a two-component nuclear power structure with a reactor fleet consisting of thermal and fast reactors as envisioned in the Russian nuclear power development strategy to 2050 and outlook to 2100 will require optimal spent nuclear fuel and radioactive waste management solutions. A core issue in this regard is managing the long-lived minor actinide (MA) inventory that affects overall nuclear power ecological safety. The study examines several options for homogenous MA (Am and Np) transmutation using modern calculation codes with MA transmutation rate and material balances taken into account. Results demonstrate that if fast reactor installed capacity reaches 92 GWe by 2100 there would not be any need for dedicated MA-burners as the MA issue would be gradually resolved within the two-component nuclear energy system by the end of the century.

Sensitivity-Analysis-Driven Surrogate Model for Molten Salt Reactors Control

The numerical analysis for the controllability assessment of a new design nuclear reactor is typically carried out by means of complex multiphysics codes, solving high fidelity partial differential equations governing the system neutronics as well as the fluid dynamics. Multiphysics codes deliver very accurate solutions at the expense of high computational times, which could be of several hours depending on the specific case study. In this work, to efficiently reduce runtimes, a sensitivity analysis (SA) is carried out to identify the most important input parameters affecting the solution of a multiphysics model developed for the controllability assessment of molten salt reactors (MSRs). The numerical modeling of these innovative systems is fundamental to allow for a safer and more sustainable power production (e.g., due to the lower radiotoxicity of the actinide inventory in MSRs and to the possibility of operation at atmospheric pressure). In this paper, four global sensitivity measures are calculated first, including the Pearson correlation coefficient, δ, Kolmogorov–Smirnov and Kuiper indices, whose results are aggregated by an ensemble strategy and confirmed by the CUmulative SUm of NOrmalized Reordered Output (CUSUNORO) plot. The results of the SA point out that the fuel density is the most important parameter yielding the largest variations in the system reactivity, fundamental for guaranteeing the MSR controllability. In light of this result, a simplified, surrogate model is then developed, which uses density as the only input parameter to determine reactivity, guaranteeing runtime reductions from several hours to a few seconds and, at the same time, a comparable level of accuracy of the multiphysics model. This result demonstrates the capability of global sensitivity analysis approaches to effectively identify the most relevant parameters in MSR systems, supporting the development of simplified, control-oriented models for these innovative reactors.

Neutronic Analysis of SiC/SiC Sandwich Cladding Design in APR-1400 under Normal Operation Conditions

Our aim is to study the neutronic behaviour of potential accident-tolerant fuel (ATF) claddings in a pressurised water reactor under normal operations. This work compares ATF silicon carbide composite (SiC/SiC) cladding to conventional ZIRLOTM cladding in APR-1400. Additionally, a “sandwich” cladding design developed by the CEA is used for SiC/SiC. The design structure includes a liner in between two layers of the composite to ensure leak tightness. The two proposed liners are Niobium (Nb) and Tantalum (Ta). Serpent 2, a Monte Carlo reactor physics lattice code, is employed to model both cladding materials in APR-1400 at three different levels: pin cell, fuel assembly, and core. The criticality, neutron spectrum, actinide inventory, and power distribution as a function of burnup are investigated. The simulations show that SiC/SiC with the Nb liner displays a far superior performance than the Ta liner across all examined characteristics. Ta leads to a harder neutron spectrum and increased Pu-239 content throughout the cycle, while Nb presents negligible effects. In fact, SiC/SiC with the Nb liner performs very similarly to ZIRLOTM at all model levels. The results indicate that, in terms of neutronics, the adoption of the SiC/SiC composite would entail little to no changes to current APR-1400 operations.

Neutronic Performance of a Compact 10 MWe Nuclear Reactor with Low Enrichment (ThxU1\u2212x)N Fuel

Recent interest in compact nuclear reactors for applications in space or in remote locations drives innovation in nuclear fuel design, especially non-oxide ceramic nuclear fuels. This work details neutronic modeling designed to support the development of a new nuclear fuel concept based on a mixture of thorium and uranium nitride. A Monte Carlo N-Particle Version 6.2 (MCNP-6) model of a compact 10 MWe reactor design which incorporates (ThxU1−x)N fuel is presented. In this context, a “compact” reactor is a completely assembled reactor which may be emptied of coolant and transported by specialized commercial vehicle, deployed by a C130J aircraft, or launched into space. Core geometry, reflector barrels, and the heat exchange zones are designed to support reduction of overall reactor volume of core components while maintaining criticality with a fixed total fuel mass of 4500 kg. Dense mixed nitrides of thorium nitride (ThN) additions in uranium nitride (UN) in 5 wt.% increments between 0.05 le x le 0.5 have been considered for calculation of k_{infty } and k_{{{text{effective}}}}. ThN additions in UN results in a slight increase in the magnitude of the temperature coefficient of reactivity, which is negative by design. The isotopic distribution of the principal actinide inventory as a function of burnup, time, and initial fuel composition is presented and discussed within the context of the proliferation risk of this core design.

Trip turbine transient and stability analysis of two different fuel assembly designs aimed for the recycling of minor actinides and plutonium in BWRs

Recycling in thermal reactors is an alternative that could help to reduce minor actinide inventories produced by depleted fuel assemblies of nuclear reactors; the current paper assesses two proposed BWR fuel assemblies to recycle MA. These fuel designs aim to no introduce any perturbation in the electricity production from the Boiling Water Reactor, and they must satisfy the reactor core safety constraints. Such constraints must be considered under steady and transient conditions without the need to make any modification to the nuclear reactor safety systems. Previously, these two fuel designs were assessed under steady state conditions. In this study, the two proposed fuel assembly designs are tested for stability analysis and trip turbine transitory conditions. These fuel assembly designs are based on a 10 × 10 standard fuel assembly, and they have four MA-bearing rods that are replacing four fuel rod assemblies in three of the six axial fuel cells that compose the fuel assembly. The first design is a UO2 fuel assembly; the second one is a MOX fuel assembly where also, 8 fuel pin rods have been replaced by water rods in three of the six axial fuel cells that compose the fuel assembly. These two proposed fuel assemblies have been designed to be equivalent by producing the same amount of energy of the BWR reactor. The stability analysis is performed considering a global reactor oscillation; results show that in the MOX fuel assembly, the control rod pattern must be modified to provide an acceptable shutdown margin and to have adequate stability conditions. Both fuel designs provide acceptable conditions under the trip turbine transient condition and provide two fuel assembly designs that could be used for MA recycling in a BWR without any perturbation in energy production.

Priority issues and key findings from evaluation of disposal records for a legacy radioactive waste site

This paper reports on a detailed investigation into the disposal procedures and records from the operational period (1960–1968) of the Little Forest Legacy Site (LFLS) in eastern Australia. The aims of the paper are firstly, to highlight the priority issues which are relevant to the radiological assessment of the LFLS, and secondly, to present key lessons that may help to guide future investigations of the records at similar sites. Particular effort was put into assessing the various types of relevant documents and the relationships between them. A specific objective of this work was to evaluate an inventory of the wastes which was reported shortly after the time of disposals. A major finding of the study is that the original actinide inventory for LFLS relied solely on estimates from a limited number of specific records known as ‘Scrap Disposal Reports’ (SDRs). For example, the estimated amount of plutonium disposed at the LFLS was based on only seven SDR records. Given that there are approximately 50000 buried items, it is possible that other Pu-contaminated items could make a significant additional contribution to the amount of Pu present at the site. For some waste components (e.g. beryllium) the documentation shows that rough estimates of disposal quantities were made, based on the number of disposed Be-contaminated items in each trench. The use of such approximations casts some doubt on the accuracy of the previous inventory of wastes at the site. In addition, the early summaries of radionuclide disposals, which categorized radionuclides into groups according to their radiological hazard, contained significant underestimates of the radionuclide inventory in the most hazardous category (referred to at the time as ‘Group I’ radionuclides). This was mainly due to the omission of the Pu (which had been recorded on the SDRs) from the Group I inventory, but was also in part because the Group I radionuclide content of disposed sludge drums (from a wastewater treatment plant) was not taken into account for most of the disposal period. Establishing the disposal history and radionuclide inventory at legacy sites is an important pre-requisite to evaluating their radiological impact and developing management options. The detailed investigation of the LFLS records shows the importance of understanding the operational practices of the period and the derivation of the original inventories. These insights should help guide future efforts to better understand disposal histories and inventories at LFLS and elsewhere.

Coupled CLASS and DONJON5 3D full-core calculations and comparison with the neural network approach for fuel cycles involving MOX fueled PWRs

The scenario code CLASS relies on infinite assembly simulation to predict fuel actinide inventories at exit burnup. In the current work, we replace these assembly calculations by full-core simulations and evaluate the impact on actinide inventories predicted by CLASS. To achieve this goal, we generate neural network training databanks for CLASS using the lattice code DRAGON5. For UOX fuels, the databanks are sampled stochastically for exit burnup, moderator boron concentration and uranium 235 enrichment while for MOX fuels an eight-dimensional grid is sampled that also accounts for plutonium and americium-241 initial contents. DRAGON5 is used to generate the databases for DONJON5 3D full-core diffusion calculations in CLASS. Results obtained using neural networks CLASS and DONJON5/CLASS calculations are then compared to assess the different assumptions used in classical scenario simulations and determine the major source of errors. A simple UOX scenario involving long-term fuel storage and a more complex scenario involving reprocessed UOX spent fuel and MOX fabrication are then studied. They show that inventories of uranium 235 and minor actinides are sensitive to full-core simulations. Moreover, the neural networks CLASS simulations can be improved using an adapted kthreshold that depends on the initial fuel composition.

Analysis of uranium oxide fuel transmutation in VVER-1000 reactor using VISTA and WIMS-D4 codes

In this study, the front end and back end of Uranium Oxide (UOX) fuel cycle were first calculated using nuclear fuel cycle simulation system (VISTA) code. The front end calculated values are 19595.3 ton/year for ore-U, 196 ton/year for natural-U, 196 ton/year for UF6, 111.4 metric ton separative work unit (MTSWU) for enrichment requirements, 172.211 ton/year depleted-U and 23.8 ton/year for Fresh Fuel (FF) requirements. The back end calculated values were 23.8 ton/year for Spent Fuel (SF), 0.024 ton/year for Actinide Inventory (AI) and 0.981 ton/year for Fission Product (FP) isotopes. After that the transmutation of UOX fuel results calculated by VISTA benchmarked with WIMS-D4 code. The calculations carried out in 30,000, 35,000, 40,000 and 45,000 MWd/tHM discharge burnup stages. The comparison of results showed that variance was less than 15%.

Development and Optimization of Voltammetry for Real Time Analysis of Multi-Component Electrorefiner Salt

There is interest by several countries in pursuing pyroprocessing of spent nuclear fuel –including both nuclear weapons states and non-nuclear weapons states. This technology features electrorefiners that can be operated to recover fissile actinides from spent nuclear fuel. This separation capability can be viewed as a proliferation risk factor, which necessitates the implementation of nuclear safeguards. For non-nuclear weapons states, safeguarding would be the responsibility of the International Atomic Energy Agency. Unlike nuclear facilities in which items can be discretely tracked, inspecting and safeguarding pyroprocessing facilities requires complicated monitoring systems and material control and accountability approaches to support the safeguards system. Therefore, there needs to be developed high precision monitoring technology for electrorefining and other key unit operations in pyroprocessing that can be implemented in real or near real time. Currently in research and development facilities such as the Fuel Conditioning Facility at Idaho National Laboratory, they have been employing mass tracking simulation supplemented with destructive analysis for the last two decades. This approach meets neither the need for timeliness or accuracy needed to detect significant quantities of material diversion in commercial facilities. Since much of the inventory of actinides in electrorefiners are held up in the molten salt, electrochemical sensors would be an ideal solution to accurately monitor actinide concentrations in real time. Electrochemical analysis methods such as cyclic voltammetry (CV), chronoamperometry (CA), and chronopotentiometry (CP), and normal pulse voltammetry (NPV) can be used to develop correlations between electric signals measured and concentrations of various ions in the molten salt. In salts with multiple components and high overall concentration of active solute, one major difficulty in applying these electroanalytical methods is to prevent area growth of the working electrode (WE)—which is largely unpredictable and affects the observable current density. Practically, all of the aforementioned voltammetry methods can encounter this problem. However, during an NPV scan, the metal deposited onto the WE is periodically oxidized back into the salt, which should effectively minimize working electrode area growth. NPV was, thus, the main focus of our study. Pulse time, relaxation time, and relaxation potential were all varied in an attempt to isolate the diffusion-limited partial current values for the different active ions in the salt. CV and CA were also applied to supplement and validate the NPV results. CV was employed to determine the potentials at which different ions reduce as well as the relaxation potential used to clean off deposited metals from the WE. CA was utilized to validate if the pulse time selected was still within the time period that the change in current was described by the Cottrell equation. Therefore, a method that was a combination of NPV, CV, and CA was applied to develop and optimize real time analysis of electrorefiner salt. Two salt mixtures were investigated in this study, LiCl-KCl-UCl3-MgCl2 and LiCl-KCl-UCl3-GdCl3, where Mg was the surrogate for Pu and Gd represented the rare earth elements in the electrorefiner salt based on the similarity in standard reduction potentials. While previously, correlations between limiting current and concentration in molten salt had been reported only up to about 1.7 wt% UCl3, this study included limiting current and concentration correlations for binary systems with up to 9 wt% UCl3. The highest MgCl2 concentration tested was 1.5 wt%, which represented 5.5% PuCl3 because the limiting current is proportional to molar concentration rather than weight percentage. When comparing limiting current density with MgCl2 concentration at fixed 1 wt% UCl3, the results did not show a good linear fit, as the slope increased with each concentration increment. However, R2 of 0.9996 was obtained when the data was fitted with a second order polynomial. The nonlinear results show that there was possibly a slight WE surface area increase. Also it is possible that the diffusion coefficient increased significantly with high concentration, resulting in higher current density. Further investigation with shorter pulse time is required to validate these hypothesizes. Nevertheless, the results show that electrochemical sensors would be an ideal solution to accurately monitor electorefiner salt concentrations in real time.

A comparison between oxide and metallic fueled ASTRID-like reactors

In order to limit the greenhouse emissions and ensure energy security, the global climate change perspective implies a modification in the future energy sector supply. Nuclear energy is an alternative that would decrease the amount of greenhouse gas emissions, and specifically, the fourth generation reactors that present a favorable outlook for clean energy supply. In particular, the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) is a fourth generation sodium-cooled pool type fast nuclear reactor of 1500MWth that seems to be a promising design into the commercial fast nuclear reactors. One of the different core designs for ASTRID is a low void effect core (CFV: Cœur à Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. The current paper presents a heterogeneous ASTRID model developed in MCNP6 with JEFF-3.2 cross sections library, in which three neutronic parameters (Keff at End of Cycle, Doppler constants and coolant void worth) are validated with the European Benchmark on the ASTRID-like low-void-effect core characterization, presented in the ICAPP conference in Nice, France, on May 2015. Validation of the ASTRID model was successfully accomplished, having slightly differences with the benchmark participants in the neutronic parameters. Once the validation was done, a metallic fueled ASTRID-like design is proposed and compared with the oxide fueled reactor. Criticality analysis was done in order to understand the reactor behavior during one cycle of operation as well as Doppler constants and coolant void worth comparison. Actinide inventory was analyzed for both models and 239Pu was tracked with the purpose of observing the breeding achieved in the fertile blankets. Reactivity at the End of Cycle had only a 3pcm delta Keff difference in comparison with the oxide design, Doppler constants behaved as expected and coolant void worth for the whole core has a negative reactivity effect, which is the essential characteristic of the CFV core. Metallic model performed 8.57% more breeding in isotope 239Pu than the conventional oxide fueled design in the fertile zone, and in the global actinide inventory achieved 3% more 239Pu. Metallic design enhances the plutonium management because, besides that the reactor can be sustained in a supercritical state with less initial Pu enrichment, breeding has a better performance.

Actinide inventory in Herndon’s georeactor operating throughout geologic time

Herndon proposed a nuclear fission reactor at the center of the Earth to explain changes in the geomagnetic field and the 3He/4He ratios observed from deep mantle sources. This study investigated the neutronic properties of the planetary-scale reactor by performing rigorous depletion simulations over geologic time by using a modified TRITON sequence in SCALE6. We also conducted analytical calculations of the rates of change of various actinides in the reactor core to identify the primary mechanisms involved in the nuclear system as a function of the operating time. The sound agreement between analytical and TRITON calculations on the predicted variations of the amounts of important actinides revealed that (1) the hypothetical nuclear georeactor is a fast-spectrum converter reactor burning only 235U; (2) the efficiency of fuel conversion approaches 0.9, and can be sustained for billions of years based on the cycle of 238U/239Pu/235U, rather than of 238U/239Pu or 232Th/233U; and (3) under appropriate conditions, the georeactor can operate at a constant power of 3TW for up to 6.5billionyears.

Investigation of the MSFR core physics and fuel cycle characteristics

The adoption of Th fuel in fast reactors is being reconsidered due to the potential favorable impact on actinide waste management and resource availability. A closed Th cycle leads to an actinide inventory with lower radiotoxicity and heat load for the first several thousands of years. Due to the typically low TRansUranic (TRU) Conversion Ratio (CR), Th can also be advantageous to expedite the consumption of legacy TRU. One of the main obstacles to the implementation of Th is the highly radioactive recycled fuel which requires remote handling under heavy shielding, inevitably penalizing economics and challenging conventional pin-based fuel manufacturing. From this perspective, the development of liquid-fuelled reactors, with Molten Salt Reactors regarded as the most promising, appears particularly attractive as fuel handling would be greatly simplified. The present paper investigates the fuel cycle performances of the reference GEN-IV Molten Salt Fast Reactor (MSFR) in terms of isotope evolution, radiotoxicity generation and safety-related parameters. Similarly to most MSR concepts proposed in the past, the MSFR is based on the fluoride molten salt technology, but it features the novelty of a fast neutron spectrum. Calculations are performed using state-of-the-art equilibrium-cycle methodologies, i.e., the ERANOS-based EQL3D procedure developed at the Paul Scherrer Institut and extended to the simulation of the MSFR. Selected results have been benchmarked with the Monte Carlo code PSG2/SERPENT. These results have also been used for the assessment of a diffusion module based on the COMSOL multi-physics toolkit, which is the subject of current studies aimed at efficiently simulating the peculiar MSFR transient behavior.

Performance of Large Breed-and-Burn Core

A sodium-cooled fast reactor breed-and-burn (B&B) core and fuel cycle concept are proposed to achieve uranium utilization in the vicinity of 50% without separation of most of the fission products from the actinides. This core is to be fueled with depleted uranium (DU) with the exception of the initial core loading that uses fissile fuel to achieve initial criticality. When the cladding reaches its radiation damage limit, the melt-refining process is used to recondition the fuel, and then the fuel is reloaded into the core. This fuel reconditioning continues until the fuel reaches the neutronically maximum attainable burnup. When a fuel assembly is discharged at its maximum attainable burnup, it is replaced with a fresh DU assembly.The maximum burnup attainable in a large 3000-MW(thermal) B&B core is found to be 57% fissions per initial metal atoms (FIMA). The discharged fuel characteristics such as the inventory of actinides, radiotoxicity, and decay heat are one order of magnitude smaller, per unit of energy generated, than those of a light water reactor operating with the once-through fuel cycle.It is also found that the minimum burnup required for sustaining the B&B mode of operation is 19.4% FIMA. The fuel discharged at this burnup has sufficient excess reactivity for establishing initial criticality in a new large B&B core. The theoretical minimum doubling time for new core spawning is estimated to be [approximately]10 effective full-power years; there is no need for any external fissile material supply beyond that required for the initial “mother” reactor.Successful development and deployment of the B&B core along with fuel reconditioning could possibly provide up to 3000 yr worth of the current global nuclear electricity generation by using the DU stockpiles already accumulated worldwide. However, a number of important feasibility issues are yet to be resolved.

Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium–plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 ( n, γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX.

Reprocessing Silicon Carbide Inert Matrix Fuel by Molten Salt Corrosion

AbstractSilicon carbide is one of the prime matrix material candidates for inert matrix fuels (IMF) which are being designed to reduce plutonium and long half-life actinide inventories through transmutation. Since complete transmutation is impractical in a single in-core run, reprocessing the inert matrix fuels becomes necessary. The current reprocessing techniques of many inert matrix materials involve dissolution of spent fuels in acidic aqueous solutions. However, SiC cannot be dissolved by that process. Thus, new reprocessing techniques are required.This paper discusses a possible way for separating transuranic (actinide) species from a bulk silicon carbide (SiC) matrix utilizing molten carbonates. Bulk reaction-bonded SiC and SiC powder (1 μm) were corroded at high temperatures (above 850 °C) in molten carbonates (K2CO3 and Na2CO3) in an air atmosphere to form water soluble silicates. Separation of Ceria (used as a surrogate for the plutonium fissile fuel) was achieved by dissolving the silicates in boiling water and leaving behind the solid ceria (CeO2).

Development of a method of evaluating an inventory of fission products for a pebble bed reactor

The design of the high temperature gas-cooled reactor (HTGR) has evolved and the relevant safety requirements have been defined; accordingly, the source term to be used as the basis for licensing must also be developed. However, analysis of the source term in the HTGR has not been adequately investigated and there has not been definite improvement in this respect. Because radioactivity in normal operation must be well understood, the purpose of this study is to establish a method for activity evaluation by the code combination MCNP-ORIGEN-MONTEBURNS-MOTEX. The sophisticated method, which constructs the HTR-10 core by using the unit lattice of a hexagonal prism, is developed for core modeling. The MCNP modeling is used to simulate the generation of fission products with an increase of burnup, and ORIGEN is utilized for depletion calculation of each fission product. Continuous fuel management is divided into five discrete periods for the feeding and discharging of fuel pebbles. MONTEBURNS is used for discrete fuel management. In short, this work by aid from MOTEX traces 41 isotope nuclides, the results of which seem highly probable. In addition, the inventory of actinides at the end of each cycle is also investigated. It would be informative when the waste management of spent fuel of HTGRs would be taken into account. This article lays the foundation for future work on the analysis of the source term in HTGRs and will hopefully serve as a platform from which the safety assessment of radioactive material release during accidents can be undertaken in future.

SIMS characterisation of actinide isotopes in irradiated nuclear fuel

Numerical codes are used to calculate the inventory of plutonium and minor actinides produced during irradiation in industrial power reactors. Secondary ion mass spectrometry (SIMS) is a useful method for validating theses inventory calculations especially with regards to isotopic characterisation. Isotopic ratios measured by SIMS on the radius of an UO 2 irradiated fuel pellet are presented and compared with calculated values in this paper. The SIMS results tally well with the calculations, as both show an increase in the actinide concentration on the pellet periphery due to the rim effect.

Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO2 Fuel Assemblies

The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8×8-2 UO2 assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of 235U, well reproduce those of 239Pu and 240Pu, yet underestimate those of 236U, 237Nd, 238Pu, 241Pu, and 242Pu. The inventories of minor actinides are underestimated by the present analysis except for 241Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8×8-4 UO2 assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of ±10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8×8-4 UO2 assemblies were well reproduced by the burnup calculations of MVP-BURN.

Isotopic Analysis of the In-Zinerator Actinide Management System

Efficient burn up of minor actinides is one of the most promising alternatives for minimizing waste in advanced nuclear fuel cycles. This work examines the concept of employing Z-pinch driven fusion source in a sub-critical transmutation reactor designed to burn up actinides and generate constant power. Its fuel cycle is designed to allow on-line fission product removal and fuel replenishment. The variation of the actinide inventory is an essential quantity used to calculate the energy multiplications and neutron spectrum, as well as to design an appropriate reactivity control mechanism.In this paper we develop a method to calculate timedependent isotopic distributions, fuel feeding rate and fission product removal rate necessary to obtain a constant power level. The calculation is performed by using both MCise, a Monte Carlo isotopic inventory code, and MCNP5. An important feature of MCise for this system is the ability to simulate the on-line removal of fission products from the actinide mixture.In addition to reporting the actinide inventory and burn rates, the impact of the actinide inventory on the fission/fusion energy multiplication will be examined.

Effects of Accelerator-Driven Transmutation System on Radiotoxicity of High-Level Radioactive Wastes

We have developed a computation tool, WAste COMposition (WACOM) for performing a scoping study of the effects of the accelerator-driven transmutation of waste (ATW) system with a lead-bismuth-eutectic-cooled transmuter on actinide inventory and radiotoxicity reduction. WACOM consists of a simplified burnup model for a chain of 18 actinide isotopes and a fuel cycle model to evaluate high-level waste (HLW) generation from the reference ATW plant. Interpolation formulas for effective one-group cross sections as a function of the actinide mass fraction have been developed. Three kinds of HLW generation were considered: (a) HLW from uranium separation for light water reactor (LWR) spent fuel, (b) HLW from the partitioning process in multicycle ATW operation, and (c) the last core of the transmuter at the decommissioning of the ATW system. The latter two HLW sources resulting from multicycle ATW operation have been found to be greater than the first source. Potential benefits of ATW deployment have been found to be (a) reduction of the total actinide toxicity by a factor of 48 at the time of waste generation and (b) conversion of the actinide mixture into a more proliferation-resistant configuration, by effective transmutation of 239Pu, 241Am, and 237Np included in the LWR spent fuel. The total actinide radiotoxicity further decreases to 1/260 for the time period of 100 000 yr, which would improve the performance of the Yucca Mountain Repository.