Transmutation Rate Research Articles

Feasibility of Photonuclear Transmutation of Long-Lived Fission Products Extracted from Used Nuclear Fuel

To achieve the goal of net-zero carbon emission in energy production, nuclear power capacity and waste generation are expected to expand significantly in the next few decades. In the condition of continuous fuel recycling, long-lived fission products (LLFPs) are dominant contributors to the disposal impacts of nuclear waste. In this study, six LLFPs, including 99Tc, 129I, 135Cs, 126Sn, 93Zr, and 79Se, were identified as the primary contributors to more than 99% of long-term radiotoxicity of disposed nuclear waste across a wide range of fuel cycle scenarios. To reduce the amounts of LLFPs sent to geological repositories, the nuclear transmutation of LLFPs is being pursued. Specifically, this work systematically assessed the feasibility of transmuting LLFPs via photonuclear reactions. Photon transmutation is physically viable for the identified primary LLFPs except for 99Tc. For the five transmutable LLFPs, the achievable photon transmutation performance without isotopic separation was evaluated based on scoping calculations and consideration of nuclear data uncertainties. Using an extremely intense laser Compton photon source of 1019 γ /s, the effective transmutation half-life can be reduced to a few years. However, the absolute transmutation rates of LLFPs remain below 1 kg/yr. The energy required to power the photon source for transmuting all LLFPs produced in a nuclear reactor exceeds the net energy output of the reactor. Several potential strategies for improving photon transmutation performance were analyzed. None can substantially enhance the performance to make it practical for industrial applications.

Searching for an optimal fuel for a supercritical water reactor (SCWR) to dispose of minor actinides resulting from the nuclear waste

This work investigates the optimal treatment of Minor Actinides (MAs) produced in the spent fuel of nuclear reactors operating around the world. This process is very important in addressing the challenges associated with nuclear waste and reducing environmental impact. A three-dimensional model of a supercritical water reactor (SCWR) has been designed using MCNPX to make a comprehensive neutronic analysis. MAs with a concentration of 1 % have been added to UO2, (Th, U)O2, (Th, 233U)O2 and (Th, rgPu)O2. These fuels have been uploaded in the SCWR assembly and burned in a separate fuel cycle. The infinity multiplication factor (kinf) of the suggested has been investigated with and without adding the minor actinides to analyze the effect of MAs on the reactor reactivity. The fuel constituents, plutonium concentration, MAs concentration and transmutation rate have been tracked with fuel burnup. The reactivity temperature coefficients have been calculated for the suggested cases to ensure the validity of the suggested fuels. The power peaking factor (PPF) and radial power distribution have been calculated for the suggested fuels. The neutronic analysis confirms the suitability of the suggested fuel in burning a significant amount of MAs.

Preliminary study of transuranic transmutation in a small modular chloride salt fast reactor

The management of radioactive waste poses a significant challenge to the sustainable development of nuclear energy. Efficient transmutation of nuclear wastes is crucial to minimize their accumulation. A small modular chloride salt fast reactor (sm-MCFR) capable of transmuting transuranic elements (TRU) is proposed in this paper, combining the advantages of the small modular reactor (SMR) and the molten salt reactor (MSR). The sm-MCFR is characterized by a high fuel loading and a compact core structure that can be quickly deployed around large commercial reactors to achieve TRU transmutation. To evaluate the TRU burnup capability of the sm-MCFR, several fuel salts and reprocessing modes were analyzed using the internally developed TRITON MODEC Coupled Burnup Code (TMCBurnup) tool. NaCl-MgCl3 with 98 % enrichment in 37Cl is chosen as carrier salt for the sm-MCFR, which can achieve 76.7 % TRU transmutation rate in average and 355 kg·GW−1·a−1 TRU transmutation quality at a continuous reprocessing rate of 10 L/d for 50 operation years. The optimized sm-MCFR reduced the radioactive toxicity of TRU by 84 %, thereby simplifying waste reprocessing. In addition, the sm-MCFR has a negative temperature feedback coefficient of −7.195 pcm/K, favoring safe reactor operation.

TRU utilization and MA transmutation in thorium-based fluorinated molten salt fast reactor

In this paper, a thorium-based fluorinated molten salt fast neutron reactor is taken as the research object, and the utilization of thorium resources, reuse of transuranic waste (TRU), and transmutation of minor actinide (MA) in fast reactor are the research purpose. The fast reactor adopts a multi-layer core structure, in which the fuel is divided into an inner fission fuel zone and an outer breeding fuel layer, respectively loaded with fluorinated molten salts LiF+ThF4+XF4 (where X is 233U, TRU, or 233U+MA) and LiF+ThF4. In addition to serving as nuclear fuel, these two molten salts also act as coolants in the primary circuits.Through the cross sections of nuclear reactions, transmutation chain of actinide nuclides, and the particle population density formula of actinide nuclides in nuclear reaction equilibrium, the possible nuclear reactions of actinide nuclides in thorium-based fast reactors are analyzed. The calculated results show that the thorium-based fast reactor has the following characteristics: fast neutron energy spectrum, negative temperature reactivity coefficient, effective utilization of thorium resources, the ability to breed fissle fuel such as 233U, production of high-purity radioisotope 238Pu material, and the ability to incinerate Pu and MAs in the TRU spent fuel.In the 233U fuel scheme, the 233U fuel is breeded, breeding ratio of fissile fuels is 1.21, the reaction products contain very little Np, and almost no Pu, Am, and Cm long-lived radionuclides. In the TRU fuel scheme, the annual capacity to incinerate MAs and Pu is 1072.6 kg, while producing a large amount of 233U and 233Pa fuel (together 1162.5 kg/year).In the 233U+2.0 wt% TRU-MAs fuel scheme, 127.35 kg of MA waste can be incinerated annually, and their transmutation rate is 15.95%, which is equivalent to the annual production of MA nuclear waste from 4.9 conventional 1000 MWe light water reactors. Here, TRU-MAs is the mixtured MA fuels after separation of Pu from the TRU fuel. At the same time, radioactive fuel 238Pu, which can be used for space nuclear power, is produced with an annual yield of 87.89 kg and an isotopic abundance of 87.57%. If the added TRU-MAs fuel is replaced with 2.0 wt% 237Np fluorinated salt, the annual yield and isotope abundance of 238Pu are further increased to 130.24 kg and 95.4%.

Development of a new method for evaluating the nuclear transmutation rate of radioactive isotopes

Development of a new method for evaluating the nuclear transmutation rate of radioactive isotopes

Transmutation of Plutonium and Minor Actinide in PWR Thorium-Transuranic Fuel Assembly

Long-lived radioactive waste is often considered a concerning issue on utilizing nuclear power. This waste is in the form of plutonium and minor actinides (MA), formed due to successive neutron capture of 238U. Both are not particularly hazardous radiologically, but their long half-life caused an issue in the public acceptance of radioactive waste disposal. Thereby, this issue must be resolved, either politically or technically. One of the technical solutions to address the issue of long-lived radioactive waste is the incineration of transuranic (TRU) elements in a pressurized water reactor (PWR) fuel assembly. Mixing TRU with thorium in a PWR fuel assembly can theoretically reduce TRU stockpile more effectively than uranium. This paper discusses plutonium and MA transmutation in a Thorium-Transuranic (Th-TRU) PWR fuel assembly using MCNP6 code and ENDF/B-VII library. A Westinghouse 17×17 PWR fuel assembly was chosen to determine the feasibility of TRU incineration. Three assemblies of 50:50, 55:45, and 60:40 mixture of UO2 and Th-TRU fuel rods are compared to one reference UO2 assembly from the criticality and burn-up point of view. From the calculations, the plutonium incineration rate was observed to be 9.47 %, 10.91 %, and 12.2 %, while MA incineration rates were found to be 11.5 %, 11.93 %, and 12.3 % for Th50-TRU50, Th55-TRU45, and Th60-TRU40, respectively. This observation indicates that a higher thorium fraction in the fuel assembly can increase the TRU transmutation rate. Therefore, a high thorium fraction is recommended to increase the transmutation rate of the TRU.

Transmutation of Plutonium and Minor Actinide in PWR Thorium-Transuranic Fuel Assembly

Long-lived radioactive waste is often considered a concerning issue on utilizing nuclear power. This waste is in the form of plutonium and minor actinides (MA), formed due to successive neutron capture of 238U. Both are not particularly hazardous radiologically, but their long half-life caused an issue in the public acceptance of radioactive waste disposal. Thereby, this issue must be resolved, either politically or technically. One of the technical solutions to address the issue of long-lived radioactive waste is the incineration of transuranic (TRU) elements in a pressurized water reactor (PWR) fuel assembly. Mixing TRU with thorium in a PWR fuel assembly can theoretically reduce TRU stockpile more effectively than uranium. This paper discusses plutonium and MA transmutation in a Thorium-Transuranic (Th-TRU) PWR fuel assembly using MCNP6 code and ENDF/B-VII library. A Westinghouse 17×17 PWR fuel assembly was chosen to determine the feasibility of TRU incineration. Three assemblies of 50:50, 55:45, and 60:40 mixture of UO2 and Th-TRU fuel rods are compared to one reference UO2 assembly from the criticality and burn-up point of view. From the calculations, the plutonium incineration rate was observed to be 9.47 %, 10.91 %, and 12.2 %, while MA incineration rates were found to be 11.5 %, 11.93 %, and 12.3 % for Th50-TRU50, Th55-TRU45, and Th60-TRU40, respectively. This observation indicates that a higher thorium fraction in the fuel assembly can increase the TRU transmutation rate. Therefore, a high thorium fraction is recommended to increase the transmutation rate of the TRU.

Neutronic evaluation for different external neutron sources in a Small-Subcritical fast reactor

This study proposes an optimized small-sized subcritical fast reactor with an external neutron source for actinide transmutation. The reactor integrates features of Accelerator Driven Systems (ADS) and Fusion-Fission Systems (FFS) based on Tokamak technology. Both external neutron sources were evaluated, one from fusion reactions and the other from spallation reactions in the designed small subcritical fast reactor. The analysis assessed neutron flux and actinide transmutation rates through fuel burnup simulations at 1200 K. The results underscore the significance of external neutron sources in subcritical systems, with the fusion-fission source presenting high actinide transmutation. The SSFR utilizes thorium oxide as fuel and Lead-Bismuth Eutectic (LBE) as coolant. Neutronic parameter comparisons between FFS and ADS neutron sources were conducted, showing advantages for actinide transmutation in the system with the FFS neutron source, which enhances transmutation, minimizing radiotoxicity. The neutronic parameters were obtained using the MCNP, while the MONTEBURNS code, which links the MCNP and the ORIGEN depletion code, was utilized to perform the fuel burnup simulations. The cross-section libraries were generated by the NJOY-2012 using the nuclear data library JEFF-3.3.

An attempt to find the optimal loading pattern of minor actinides in the PWR fuel rods

Three approaches of minor actinide (MA) loading in the fuel rods of a pressurized water reactor were examined. They comprised inserting MAs and fuel mixture in the center of duplex fuel rods, coating of MAs and fuel mixture in the outer layer of fuel rods and lastly, mixing MAs homogeneously with the entire fuel rod elements. AP1000 fuel assembly was chosen as the reference design. Two fuel types namely conventional UO2 and UTh MOX fuel were selected as the base fuel models for the study. Although shorter cycle length than the UO2 fuel design was achieved by the UTh fuel design, UTh fuel exhibited more favorable FTC and yielded a lower plutonium stockpile at the end of the fuel cycle. When MAs were added in the outer layer of fuel rods, it resulted in higher rate of transmutation, however it had an adverse impact on the fuel reactivity, discharge burnup and pin power distribution. On the contrary, when MAs were loaded in the central region, it led to incomplete incineration of MAs. Therefore, based on the satisfactory transmutation rate, MTC, most negative FTC and uniform pin power distribution, the homogenous MA loading strategy can be concluded as the optimal loading pattern for the AP1000 nuclear reactor.

An alternative approach for excess reactivity control of TRISO particles

This study explores an innovative approach to manage the initial excess reactivity in High-Temperature Gas-Cooled Reactors (HTGRs) by modifying the fuel and non-fuel compositions of TRISO particles. Choosing a single fuel block of the 350 MWth Prismatic HTR as the reference, minor actinides (MAs) in three loading amounts (2 wt.%, 5 wt.%, and 10 wt.%) were introduced into the TRISO fuel kernels across three alternating rings of the assembly. Four more models were developed by replacing the TRISO particles in these rings with QUADRISO, featuring distinct layers of four burnable absorbers (B4C, Er2O3, Eu2O3, and Gd2O3), each having a thickness of 2 µm. These approaches were compared against using conventional B4C poison rods, placed at adjacent and opposite ends of the hexagonal assembly. Depletion analysis over a period of 600 EFPDs was carried out using open-source Monte Carlo code OpenMC. The Eu2O3 layered particles and oppositely placed B4C rods exhibited the most effective reduction in initial reactivity, amounting to 0.28 Δk/k and 0.17 Δk/k respectively. The 10 wt.% MA loading approach proved to be versatile, as it efficiently lowered the initial reactivity by 0.15 Δk/k while also achieving minor actinide transmutation, yielding transmutation rates of 23.1%/yr for 237Np, 51.5%/yr for 241Am, and 30%/yr for 243Am. Fuel cycle parameters were calculated for multi-batch refueling schemes using the linear reactivity model (LRM). While MAs reduced burnup and cycle length, the use of B4C, Er2O3 and Eu2O3 layers improved fuel cycle performance, with B4C demonstrating the best results. Due to its high absorption cross section, Gd2O3 rendered the core sub-critical initially, resulting in negative fuel cycle parameters. The fuel temperature coefficient (FTC), relative pin-power distribution, and energy-dependent neutron flux of the models were also evaluated.

Irradiation performance and first examinations of Americium bearing blanket fuel from the MARINE irradiation experiment

Americium (Am) is a strong contributor to the long-term radiotoxicity of high-level waste from nuclear fuels. Transmutation of long-lived nuclides like 241Am by irradiation in nuclear reactors is therefore an option for the reduction of radiotoxicity and heat production of waste volumes to be stored in a repository. The MARINE irradiation experiment is the latest in a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR) in Petten (The Netherlands). The development and irradiation of MARINE was carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). Dismantling was completed and post-irradiation examinations (PIE) were started within the Dutch national research programme PIONEER. Destructive PIE is foreseen within the Euratom H2020 funded project PATRICIA.The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus pellet fuel, in particular the role of microstructure and temperature on fission gas and helium release dynamics on fuel swelling.The MARINE experiment was irradiated for 359 Full Power Days in the HFR in 2016 and 2017. This paper discusses results from irradiation, i.e. power and temperature history and transmutation rates as well as preliminary results from post irradiation examinations, assessing a.o. clad strains and helium and fission gas release and first ceramographic observations, putting a preliminary upper bound on fuel swelling.

Enhancing transmutation rates of minor actinides using Zr and Y hydride and deuteride coatings in an LFR

The possibility of transmutation of minor actinides (MAs) in lead-cooled fast reactors provides an excellent opportunity to reduce the overall repository of radioactive waste. The present work seeks to augment the transmutation rates of six minor actinides (237Np, 241Am, 243Am, 243Cm, 244Cm, and 245Cm) by increasing neutron capture through the use of moderator coatings around the fuel elements. The reference design for the fuel assembly was based on the design concept of ALFRED, the European Lead Cooled Demonstrator Reactor, and consists of 127 fuel elements loaded with 5 wt.% of minor actinides (MAs) homogeneously mixed with MOX (mixed oxide) fuel. Four different moderators (ZrH1.6, ZrD1.6, YH2, and YD2) with five different coating thicknesses (0.01 cm, 0.02 cm, 0.03 cm, 0.04 cm, and 0.05 cm) were selected for this study and variation in kinf, transmutation rates and fuel cycle parameters (employing the linear reactivity model or LRM) were determined. Results showed that both the addition of a moderator and the increase in moderator thickness improved transmutation rates but had a negative impact on fuel cycle parameters. ZrH1.6 resulted in the highest increase in transmutation rate, but also incurred the largest penalties in fuel burnup, and cycle length. For ZrH1.6 coatings of 0.01 cm, the obtained transmutation rates were 13.57%/y for 237Np, 12.6%/y for 241Am, and 9.26%/y for 243Am. The concentration of Cm isotopes increased over time due to neutron capture of Am, U, and Pu. Deuteride moderators minimized the decrease in fuel cycle parameters but compromised transmutation rates. The fuel temperature coefficient (FTC), energy dependent neutron flux, and relative pin power distribution of the reference design and the models were also determined in order to analyze the effect of moderator addition on these factors. Finally, a suggestion of the best model was put forth in consideration of both transmutation efficiency and fuel cycle parameters.

Transmutation study of 237Np in CiADS

The China initiative Accelerator Driven System (CiADS) is a megawatt-scale principleverification facility. Since 237Np is the main long-lived MA nuclide in the discharged fuel of Pressurized Water Reactors (PWRs), the transmutation efficiency of 237Npin CiADS is studied. We use the code FLUKA to simulate the spallation reaction between the proton and the target material. The neutron transport and nuclides burnup coupling codeIMPC-Burnup 2.0 is used to simulate the transmutation of 237Np. Three different positions are selected to place the transmutation assemblies. The simulation results showed that after 3000 days of transmutation, the disappearance rates of 237Np at positions 1, 2 and 3 are 10.82%, 8.54% and 7.52%, respectively. And the effective transmutation rates of 237Np caused by fission reaction at positions 1, 2 and 3 are 1.66%, 0.71% and 0.26%, respectively. This work can provide a basic reference for the design of CiADS transmutation proof-of-principle experiments.

Intelligent solution predictive networks for non-linear tumor-immune delayed model

In this article, we analyze the dynamics of the non-linear tumor-immune delayed (TID) model illustrating the interaction among tumor cells and the immune system (cytotoxic T lymphocytes, T helper cells), where the delays portray the times required for molecule formation, cell growth, segregation, and transportation, among other factors by exploiting the knacks of soft computing paradigm utilizing neural networks with back propagation Levenberg Marquardt approach (NNLMA). The governing differential delayed system of non-linear TID, which comprised the densities of the tumor population, cytotoxic T lymphocytes and T helper cells, is represented by non-linear delay ordinary differential equations with three classes. The baseline data is formulated by exploiting the explicit Runge-Kutta method (RKM) by diverting the transmutation rate of Tc to Th of the Tc population, transmutation rate of Tc to Th of the Th population, eradication of tumor cells through Tc cells, eradication of tumor cells through Th cells, Tc cells’ natural mortality rate, Th cells’ natural mortality rate as well as time delay. The approximated solution of the non-linear TID model is determined by randomly subdividing the formulated data samples for training, testing, as well as validation sets in the network formulation and learning procedures. The strength, reliability, and efficacy of the designed NNLMA for solving non-linear TID model are endorsed by small/negligible absolute errors, error histogram studies, mean squared errors based convergence and close to optimal modeling index for regression measurements.

The diagnostic accuracy of currently available laboratory tests compared to PCR test for COVID-19: a systematic review and meta-analysis

Coronavirus disease (COVID-19) is caused by SARS-COV-2 and represents the causative agent of a potentially fatal and pandemic viral disease that is of great global public health concern and high transmutation rate and mechanisms. A comprehensive systematic search was performed on PubMed, Scopus, Web of Science, and Google Scholar, to find articles published until March 30, 2020. All relevant articles that reported clinical characteristics and laboratory result information of hospitalized COVID-19 patients were included in the analysis. Based on our analysis, however, articles not used RT-PCR and CT as were excluded all of the randomized controlled trial (RCT) were homogeneous (I2=94.58% and p≤0.001) statistical tests showed no publication bias/small study effect. Accordingly, the result of meta-analysis illustrated that RT-PCR increased cases of laboratory confirmation by more than 99% (95% CI: -1.3.0; 4.74) than other laboratory tests. As a result, this finding, is recommended to pilot and scale-up well and accurate laboratory techniques is used in developing countries, and is essential to prevent and control the pandemic diseases with spread in the world. However, it is recommended to further investigate the adverse effects of Real-time reverse transcriptase polymerase chain reaction (RT-PCR) before fully implementing the confirmation.

Computational simulation of minor actinide burning in a BN-600 reactor with fuel without uranium and plutonium

The paper presents the results of studies on the burning of minor actinides (MA) extracted from SNF of thermal reactors in a BN-600 reactor, which uses the complete set of MAs instead of traditional nuclear fuel types: uranium and/or plutonium. The advantages of such approach to MA burning are that long-lived waste is recycled and energy is produced that can be used, e.g., to generate electricity. Besides, where, e.g., a reactor with uranium or MOX fuel is used for transmutation, apart from burning “foreign” minor actinides, it will additionally generate “its own” MAs. The studies have shown that such reactor can be efficient only if based on fast neutrons, which is due to the specific properties of the minor actinide neutron capture and fission cross-sections as compared with traditional fuel nuclides. The calculation results have shown rather a high rate of MA transmutation and burning in a reactor fueled with minor actinides.

An assessment of scenarios for transuranic burning and 233U production in a small molten chloride fast reactor

Molten Chloride Fast Reactor (MCFR) attracts more and more attention owing to its significantly hard neutron spectrum and high solubility for actinide trichlorides, making it an appealing option for burning transuranic (TRU) elements discharged from pressurized water reactors (PWRs). Meanwhile, in recent years, the development of thorium fuel cycle has aroused considerable interest worldwide with the purpose of saving uranium reserves and further reducing the production of long-lived minor actinides. A small MCFR incorporating TRU burning and 233U production is proposed in this work in order to evaluate its incineration and production capabilities. By employing an in-house developed tool based on SCALE 6.1, the fuel cycle features of the small MCFR with three fuel loading scenarios, including pure TRU (scenario 1), TRU plus depleted uranium (scenario 2) and TRU plus thorium (scenario 3), are analyzed. The evolution behaviors of main nuclides, reactivity worth, neutron spectra and temperature reactivity coefficient (TRC) are discussed in detail. It is found that the MCFR can burn TRU effectively and the transmutation rates for the three scenarios are 930 kg/GWe/year, 375 kg/GWe/year and 768 kg/GWe/year, respectively, achieving a maximum capability for TRU burning of about 3.7 times the annual TRU production in a 1-GWe PWR with 50 MWd/kgU discharge burnup. Meanwhile, scenario 3 has the highest 233U production capability of about 110.3 kg per year, although it does not offer processing advantages over scenario 1 and scenario 2. The analysis of TRC indicates that the fuel salt TRC and the total core TRC are both strongly negative for all three scenarios owing to the thermal expansion of liquid fuel salt. Finally, a nuclear energy system consisting of both fast and thermal molten salt reactors is proposed and evaluated to establish a closed nuclear fuel cycle with very low radioactive wastes.

Analysis of Compositional Nuclear Transmutation Effects in Vanadium Alloys under Irradiation

The effects of nuclear transmutation in vanadium-based alloys under irradiation are considered in different reactor units including fast breeder reactors, such as (1) BOR-60 (experimental fast reactor), (2) EBR-II (experimental breeder reactor), and ITER (international thermonuclear experimental reactor). A comparison of nuclear transmutation rates for the analyzed [V–x% Cr(Ni)–y% Ti] alloys by data mining facilities is performed. It is demonstrated that the dominant types of interactions between neutrons and tested materials correlate with the level of thermalization in the neutron spectrum of a reactor, which has an effect on the value of nuclear transmutation rates to lead to distinctions in the resulting chemical composition of alloys after irradiation.

Transmutation of MAs and LLFPs with a lead-cooled fast reactor

The management of nuclear wastes has long been a problem that hinders the sustainable and clean utilization of nuclear energy since the advent of nuclear power. These nuclear wastes include minor actinides (MAs: 237Np, 241Am, 243Am, 244Cm and 245Cm) and long-lived fission products (LLFPs: 79Se, 93Zr, 99Tc, 107Pd, 129I and 135Cs), and yet are hard to be handled. In this work, we propose a scheme that can transmute almost all the MAs and LLFPs with a lead-cooled fast reactor (LFR). In this scheme, the MAs and the LLFPs are loaded to the fuel assembly and the blanket assembly for transmutation, respectively. In order to study the effect of MAs loading on the operation of the core, the neutron flux distribution, spectra, and the keff are further compared with and without MAs loading. Then the LLFPs composition is optimized and the support ratio is obtained to be 1.22 for 237Np, 1.63 for 241Am, 1.27 for 243Am, 1.32 for 79Se, 1.53 for 99Tc, 1.02 for 107Pd, and 1.12 for 129I, respectively, indicating that a self-sustained transmutation can be achieved. Accordingly, the transmutation rate of these nuclides was 13.07%/y for 237Np, 15.18%/y for 241Am, 13.34%/y for 243Am, 0.58%/y for 79Se, 0.92%/y for 99Tc, 1.17%/y for 107Pd, 0.56%/y for 129I. Our results show that a lead-cooled fast reactor can be potentially used to manage nuclear wastes with high levels of long-lived radioactivity.

Results of the 244Cm, 246Cm and 248Cm neutron-induced capture cross sections measurements at EAR1 and EAR2 of the n_TOF facility

Accurate neutron capture cross section data for minor actinides (MAs) are required to estimate the production and transmutation rates of MAs in light water reactors, critical fast reactors like Gen-IV systems, and other innovative reactor systems such as accelerator driven systems (ADS). In particular, 244Cm, 246Cm and 248Cm play a role in the transport, storage and transmutation of the nuclear waste of the current nuclear reactors, due to the contribution of these isotopes to the radiotoxicity, neutron emission, and decay heat in the spent nuclear fuel. Also, capture reactions in these Cm isotopes open the path for the formation of heavier elements. In this work, the results of the capture cross section measurement on 244Cm, 246Cm and 248Cm performed at the CERN n_TOF facility are presented. It is important to notice that the Cm samples used in the experiment at n_TOF have been used previously in an experiment at J-PARC, this experiment and the previous one done in the 70s with a nuclear explosion were the only previous capture experiments for these isotopes. At n_TOF, the capture cross section measurements of 244Cm, 246Cm and 248Cm were performed at the 20 m vertical flight path (EAR2) with three C6D6 total energy detectors. In addition, the cross section of 244Cm was measured at the 185 m flight path (EAR1) with a Total Absorption Calorimeter (TAC). The combination of measurements in EAR1 and EAR2 has contributed to controlling and reducing the systematic uncertainties in the results. The compatibility of the different measurements performed and the techniques to obtain the results are presented in this paper as well as the procedure to obtain the resonance parameters.