Transmutation Rate Research Articles

Method for accounting for macroscopic heterogeneities in reactor material balance generation in fuel cycle simulations

This paper describes the addition of spatially dependent power sharing to a methodology used for calculating the input and output isotopics and burnup of nuclear reactors within a nuclear fuel cycle simulator. Neutron balance and depletion calculations are carried out using pre-calculated fluence-based libraries. These libraries track the transmutation and neutron economy evolution of unit masses of nuclides available in input fuel. The work presented in the paper generalizes the method to simulate reactors that contain more than one type of fuel as well as strong spatial and flux heterogeneities, for instance breeders with a driver–blanket configuration. To achieve this, spatial flux calculations are used to determine the fluence-dependent relative average fluxes inside macroscopic spatial regions. These fluxes are then used to determine the average power of macroscopic spatial regions as well as to more accurately calculate region-specific transmutation rates. The paper presents several cases where the fluence based approach alone would not have been sufficient to determine results.

Preliminary studies of a new accelerator-driven minor actinide burner in industrial scale

Pursuing high transmutation rate of minor actinide (MA), a preliminary conceptual design of a lead-bismuth (LBE) cooled accelerator-driven system (ADS) is proposed in this study. Parametric studies are performed to optimize the neutronics and thermal–hydraulics performances. The proton energy and axial position of the proton beam impact is investigated to obtain high neutron source efficiency and spallation neutron yield. The influences of MA/Pu mixing ratio and the ratio of pin pitch to diameter (P/D) are also optimized to control the burnup reactivity swing and the minimum coolant velocity for adequate cooling. To reduce the power peak, three kinds of power flattening techniques are adopted and compared. The results show that the inert matrix ratio zone loading method seems more versatile. Based on the analyses, an optimized three zone loading pattern is proposed for the 800MWth subcritical core. The total transmutation rate of MA is 328.8kg per effective full power year. Preliminary safety analyses based on the balance of power method (BOP) are performed and the results show that in the reactivity insertion and beam overpower transients, the core shows inherent safety, but the scram is necessary by cutting off the beam current to protect the core from possible damages caused by the loss of flow.

High-Safety Fast Reactor Core Concepts to Improve Transmutation Efficiency of Long-lived Radioactive Waste

Abstract High-safety FR core concepts to improve transmutation efficiency of MAs were proposed. A heterogeneous MA loaded core was designed based on the 1000MWt-ABR breakeven core, which assumes the same amount of MAs with the 5%-MA homogeneous loaded core. The heterogeneous MA loaded core with 25%-ZrH 1.6 loaded moderator targets had a better transmutation performance than the 5%-MA homogeneous loaded core. As one of the possible design options for the MA target, the annular pellet rod design was proposed. It was shown that using the annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0wt.%, 4.0wt.%, 6.0wt.%, 8.0wt.%, 10.0wt.% and 12.0wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library.In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void worth increased with the MA weight fraction, while the Doppler feedback and the effective delayed neutron fraction decreased with the increase of the MA weight fraction.

Neutron-induced transmutation reactions in 237Np, 238Pu, and 239Pu at the massive natural uranium spallation target

Transmutation reactions in the 237Np, 238Pu, and 239Pu samples were investigated in the neutron field generated inside a massive (m=512kg) natural uranium spallation target. The uranium target assembly QUINTA was irradiated with the deuteron beams of kinetic energy 2, 4, and 8GeV provided by the Nuclotron accelerator at the Joint Institute for Nuclear Research (JINR) in Dubna. The neutron-induced transmutation of the actinide samples was measured off-line by implementing methods of gamma-ray spectrometry with HPGe detectors. Results of measurement are expressed in the form of both the individual reaction rates and average fission transmutation rates. For the purpose of validation of radiation transport programs, the experimental results were compared with simulations of neutron production and distribution performed by the MCNPX 2.7 and MARS15 codes employing the INCL4-ABLA physics models and LAQGSM event generator, respectively. In general, a good agreement between the experimental and calculated reaction rates was found in the whole interval of provided beam energies.

Seed and blanket ADS using thorium–reprocessed fuel: Parametric survey on TRU transmutation performance and safety characteristics

Conceptual designs of accelerator driven systems (ADS) that utilize thorium fuel as blanket and reprocessed fuel as seed for fission reaction in order to transmute the transuranic elements in spent nuclear fuel and produce energy from thorium utilization was proposed. In order to analyze the TRU transmutation performance of the system while the safety margin during the operation is satisfied, a parametric survey is done in this study. The impact of minor actinide (MA) amount and the reactor size on transmutation efficiency and safety characteristics of the ADS is investigated using MCNPX code using the ENDF/B-VII. As the results, with increasing MA content, TRU transmutation rate is increased while void reactivity of the system becomes less negative. Besides, increasing the core size by introducing more thorium and reprocessed fuel assemblies into the system reduces the TRU transmutation rate and reactivity swing due to burnup of the system. In order to match the efficient transmutation purpose and safety features during operation, higher MA content is suggested to introduce into the system.

康普顿γ光源SLEGS对长寿命裂变产物126Sn的嬗变潜力

By combining the simulation program 4D-MCLCSS of laser Compton scattering and the package of Geant4,the gamma rays generation of Shanghai laser-electron gamma source( SLEGS)and its impact on the photo-transmutation process of the long-lived fission products126 Sn was simulated. The dependency of transmutation ratio and the geometry parameters of transmutation target was researched,and the transmutation product distribution based on SLEGS was initially diagnosed. It is shown that the best target thickness and the best radius to obtain the transmutation target of126 Sn are respectively 16 cm and 0. 4 cm,and the corresponding transmutation rate is 1. 89 × 106/ s. The research results show that the light photo-transmutation rate based on SLEGS is twice higher than the laser-driven bremsstrahlung,and its upgrade potential is huge.

Validation of minor actinides fission neutron cross-sections

Verification of neutron fission cross-sections of minor actinides from some recently available evaluated nuclear data libraries was carried out by comparison of the reaction rates calculated by the MCNP6.1 computer code to the experimental values. The experimental samples, containing thin layers of 235U, 237Np, 238,239,240,241Pu, 242mAm, 243Cm, 245Cm, and 247Cm, deposited on metal support and foils of 235U (pseudo-alloy 27Al + 235U), 238U, natIn, 64Zn, 27Al, and multi-component sample alloy 27Al + 55Mn + natCu + natLu + 197Au, were irradiated in the channels of the tank containing fluorine salts 0.52NaF + 0.48ZrF4, labelled as the Micromodel Salt Blanket, inserted in the lattice centre of the MAKET heavy water critical assembly at the Institute for Theoretical and Experimental Physics, Moscow. This paper is a continuation of earlier initiated scientific-research activities carried out for validation of the evaluated fission cross-sections of actinides that were supposed to be used for the quality examination of the fuel design of the accelerator driven systems or fast reactors, and consequently, determination of transmutation rates of actinides, and therefore, determination of operation parameters of these reactor facilities. These scientific-research activities were carried out within a frame of scientific projects supported by the International Science and Technology Center and the International Atomic Energy Agency co-ordinated research activities, from 1999 to 2010. Obtained results confirm that further research is needed in evaluations in order to establish better neutron cross-section data for the minor actinides and selected nuclides which could be used in the accelerator driven systems or fast reactors.

Sensitivity analysis of minor actinides transmutation to physical and technological parameters

Minor actinides transmutation is one of the three main axis defined by the 2006 French law for management of nuclear waste, along with long-term storage and use of a deep geological repository. Transmutation options for critical systems can be divided in two different approaches: (a) homogeneous transmutation, in which minor actinides are mixed with the fuel. This exhibits the drawback of “polluting” the entire fuel cycle with minor actinides and also has an important impact on core reactivity coefficients such as Doppler Effect or sodium void worth for fast reactors when the minor actinides fraction increases above 3 to 5% depending on the core; (b) heterogeneous transmutation, in which minor actinides are inserted into transmutation targets which can be located in the center or in the periphery of the core. This presents the advantage of decoupling the management of the minor actinides from the conventional fuel and not impacting the core reactivity coefficients. In both cases, the design and analyses of potential transmutation systems have been carried out in the frame of Gen IV fast reactor using a “perturbation” approach in which nominal power reactor parameters are modified to accommodate the loading of minor actinides. However, when designing such a transmutation strategy, parameters from all steps of the fuel cycle must be taken into account, such as spent fuel heat load, gamma or neutron sources or fabrication feasibility. Considering a multi-recycling strategy of minor actinides, an analysis of relevant estimators necessary to fully analyze a transmutation strategy has been performed in this work and a sensitivity analysis of these estimators to a broad choice of reactors and fuel cycle parameters has been carried out. No threshold or percolation effects were observed. Saturation of transmutation rate with regards to several parameters has been observed, namely the minor actinides volume fraction and the irradiation time. Estimators of interest that have been derived from this approach include the maximum neutron source and decay heat load acceptable at reprocessing and fabrication steps, which influence among other things the total minor actinides inventory, the overall complexity of the cycle and the size of the geological repository. Based on this analysis, a new methodology to assess transmutation strategies is proposed.

Electron linac based mixed-field transmutation of 129I

The goal of this work was to determine the feasibility of using electron linacs for transmutation of selective long-lived fission products (LLFPs), particularly 129I. We have designed a converter system that produces both photons and neutrons, thus creating a mixed radiation field. The strength of the field depends on electron beam energy and current, as well as the converter material and geometry. The converter was optimized to maximize transmutation rate of 129I via two channels: 129I(γ,n)128I and 129I(n,γ)130I. The optimum electron beam energy was found to be around 100MeV. Computer simulations were benchmarked by experimental data on 133Cs (129I surrogate) photon-induced and neutron-induced transmutation rates. While further optimization of the system is necessary, we have demonstrated that accelerator-driven photonuclear reactions can be practical for 129I transmutation.

Uncertainty analysis of infinite homogeneous lead and sodium cooled fast reactors at beginning of life

The objective of the present work is to estimate breeding ratio, radiation damage rate and minor actinide transmutation rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed taking into account uncertainty in nuclear data and composition of the reactors. We use the recently released ENDF/B-VII.1 nuclear data library and restrict the work to the beginning of reactor life.We work under multigroup approximation. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. Modeling error and numerical error are estimated.The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the responses. The generalized adjoint fluxes are used to calculate first order sensitivities of the responses to model parameters. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the responses.We show that the uncertainty in model parameters is the dominant source of uncertainty, followed by modeling error, input data precision and numerical error. The uncertainty due to composition of the reactor is low. We identify main sources of uncertainty and note that the low-fidelity evaluation of 16O is problematic due to lack of correlation between total and elastic reactions.

Transmutation characteristics of minor actinides in a low aspect ratio tokamak fusion reactor

The transmutation characteristics of minor actinides in the transmutation reactor of a low aspect ratio (LAR) tokamak are investigated. One-dimensional neutron transport and burn-up calculations coupled with a tokamak systems analysis were performed to determine optimal system parameters. The dependence of the transmutation characteristics, including the neutron multiplication factor, produced power, and the transmutation rate, on the aspect ratio A in the range of 1.5–2.0 was examined. By adding Pu239 to the transmutation blanket as a neutron multiplication material, it was shown that a single transmutation reactor producing a fusion power of 150MWth can destroy minor actinides contained in the spent fuels for more than 38units of 1GWe pressurized water reactors (PWRs) while producing a power in the range of 1.8–6.8GWth.

On the Use of a Molten Salt Fast Reactor to Apply an Idealized Transmutation Scenario for the Nuclear Phase Out

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

Neutron production rates by inverse-beta decay in fully ionized plasmas

Recently we showed that the nuclear transmutation rates are largely overestimated in the Widom-Larsen theory of the so called `Low Energy Nuclear Reactions'. Here we show that unbound plasma electrons are even less likely to initiate nuclear transmutations.

Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: 235U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes.

Conceptual study of fusion-driven system for nuclear waste transmutation

A conceptual study of a fusion-driven system for nuclear waste transmutation using a low aspect ratio (LAR) tokamak as a neutron source is performed. A configuration of the LAR tokamak neutron source optimised with respect to both transmutation rate and the tritium breeding ratio for aspect ratio A in the range of 1.5–2.0 is found. The transmutation characteristics of both transuranic actinides and minor actinides are investigated and compared. When the transuranic actinides are loaded in the blanket, the neutron multiplication factor decreases from its initial value, keff=0.95, but with the minor actinides loaded in the blanket, the neutron multiplication factor shows a peak value during burn-up. The peak value can be controlled by adjusting the blanket dimensions. To transmute the nuclear waste effectively, an equilibrium fuel cycle is developed for both transuranic actinide and minor actinide transmutation.

MYRRHA: A multipurpose nuclear research facility

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback.As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

Study of Thorium Fuel Cycles for Light Water Reactor VBER-150

The main objective of this paper is to examine the use of thorium-based fuel cycle for the transportable reactors or transportable nuclear power plants (TNPP) VBER-150 concept, in particular the neutronic behavior. The thorium-based fuel cycles included Th232+Pu239, Th232+U233, and Th232+U and the standard design fuel UOX. Parameters related to the neutronic behavior such as burnup, nuclear fuel breeding, MA stockpile, and Pu isotopes production (among others) were used to compare the fuel cycles. The Pu transmutation rate and accumulation of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The Th232+U233 fuel cycle proved to be the best cycle for minimizing the production of Pu and MA. The neutronic calculations have been performed with the well-known MCNPX computational code, which was verified for this type of fuel performing calculation of the IAEA benchmark announced by IAEA-TECDOC-1349.

Transmutation of minor actinides in the pressurized water reactors

We study the effects of adding MA nuclides to the PWR core and the MA nuclide transmutation rate in the PWRs. Our calculation results show that even 1% homogeneous addition of MA nuclides to the nuclear fuel can reduce keff of the PWRs drastically. The uniform distribution of MA nuclides in the uranium dioxide fuel can also affect the lifetime of a fuel loading. Calculation results also indicate that the uniform distribution of MA nuclides in the uranium dioxide fuel results in a reactivity mismatch and control difficulties in the PWRs. The spatial self-shielding effects of the heterogeneous distributions of MA nuclides in the PWR core can avoid the initial reactivity to drop significantly, and the reactivity mismatch and control difficulties in the PWRs can be also overcome.During 300-day-exposure of MA nuclides in the PWRs 14.8% Pu-238 and 7.7% Pu-239 of the initial MA nuclides are created, this explains at least 22.5% MA nuclides transmute to plutonium isotopes during 300-day-exposure in PWRs by various nuclear processes. We may incinerate plutonium-239 and plutonium-238 isotopes in the subsequent MOX fuel loading of the PWRs. Alternatively, we may also use Pu-238 created in the transmutation of MA nuclides in the PWRs to fabricate the nuclear batteries. Pu-238 mainly transmuted from neptunium-237 in PWRs, neptunium-237 constitutes 56.2% of the total minor actinides in the depleted nuclear fuel of PWRs. The majority of commercial reactors in operation in the world are PWRs, if we get the neptunium-237 transmuted in the PWRs during their power generating, the inventory of high level long-lived radioactive minor actinides in the world will be greatly reduced.Our study show MA nuclides actually can act as the burnable poisons in the PWRs. MA nuclide transmutation materials may be used to partially substitute for the burnable poisons in the PWRs, or reduce the concentration of the boric acid in the coolant of the PWRs and increase the negative temperature coefficient of the PWRs. This is a distinct advantage to transmute MA nuclides in the PWRs. The transmutation characteristics of MA nuclides in PWRs and high flux thermal reactors are also compared in this research.

Behavior of Transuranic Mixed-Oxide Fuel in a CANDU-900 Reactor

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation, and heat loads of spent material. The burning of transuranic (TRU) fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while significantly reducing the fast reactor infrastructure needed. This paper examines the features of actinide mixed-oxide (MOX) fuel, TRUMOX, in a CANDU® nuclear reactor. The actinide concentrations used were based on extraction from 30-year-cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, supercell calculations were analyzed in DRAGON, and full-core analysis was executed in the RFSP two-group diffusion code. A time-average full-core model was produced and analyzed for reactor coefficients, reactivity device worth, and online fueling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 29.91 MWd/kg heavy element and an actinide transmutation rate of 35% for a single pass. A fully TRUMOX-fueled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing, and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle.