Transmutation Of Fission Products Research Articles

Yield of long-lived fission product transmutation using proton-, deuteron-, and alpha particle-induced spallation

Yield of long-lived fission product transmutation using proton-, deuteron-, and alpha particle-induced spallation

Monte Carlo–Informed Decay Heat Model for Cermet LEU-NTP Systems

Two methodologies for performing decay heat analysis with Monte Carlo simulations were developed and implemented on a representative nuclear thermal propulsion (NTP) system. This paper presents the underlying theory, discusses the methodology, and states the key results. This work investigated the importance of utilizing a time-dependent Q-value for fission in NTP systems due to their short burn time. Two approaches for deriving the Q-value were taken: one based on deconvolving the fission rate from the reactor power to yield the rate of fission energy deposition, and the other based on the convergence of the fission product decay power during a long burn. The fission product decay power method is hypothesized to be the more accurate representation of an NTP system as it captures more of the underlying physics occurring during burnup, such as fission product transmutation. The calculated Q-values were employed to derive decay power profiles that were compared to the current state-of-the-art NTP decay power model. According to these new models, it is shown that the cooling requirements for decay heat removal calculated with the state-of-the-art model differ from the developed methods by as much as 23.3%. There exists a need to experimentally validate, and by extension improve, the proposed methods to better understand the nature of decay heat production in NTP systems.

Chemical evolution effects on phase and microstructure of [CsxBay][Ti3+2y+xTi4+8-2y-x]O16 ceramic waste forms for radioactive cesium immobilization

The effects of fission product transmutation on the structural stability of ceramic waste forms are a critical concern due to the changes of valence and ionic size, which can impair the stability of the host matrices. In this study, [CsxBay][Ti3+2y+xTi4+8-2y-x]O16 (0.3 ≤ x, y ≤ 0.75) ceramics were designed and successfully synthesized to immobilized 137Cs and its daughter product 137Ba. The evolution of phase and microstructure of the hollandite-ceramic waste forms were investigated by the substitution of Ba2+ into the Cs+ site and the incorporation of Ti3+ for charge compensation. The results show that the prepared hollandite ceramics exhibit excellent capacity to accommodate Cs+ and Ba2+ in its lattice while retaining a stable crystalline structure with a compact microstructure. Additionally, all elements of Cs, Ba, Ti and O homogeneously distributed in the sintered hollandite ceramics. These results further indicate that hollandite ceramics are a promising candidate for selective and durable immobilization of radioactive cesium.

Evaluation of the specific radioactivity of 40 elements created by nuclear transmutation of fission products

A burnup calculation was performed to analyze Après ORIENT process which aims at creating highly-valuable elements by nuclear transmutation of fission products (FPs) separated from LWR spent nuclear fuels. In this paper, numerical evaluation of the specific radioactivity of 40 created elements, from reloaded each FP element, with atomic number from 31 to 70 at the end of the time of 5-year-cooling after the irradiation for 1125 days in each LWR and FBR was carried out. These 40 created elements were classified in 6 categories according to levels of the specific radioactivity and the length of additional cooling period, which was needed for the specific radioactivity to decrease below the exemption level defined by International Atomic Energy Agency. As a result, created 31Ga, 32Ge, 33As, 35Br, 68Er, and 70Yb did not contain any radioisotopes at the end of the 5-year-cooling. It should be noted that created 37Rb, 57La, and 60Nd had much lower specific radioactivities than natural composition of them. Moreover, specific radioactivities of created 40Zr, 42Mo, 44Ru, 46Pd, 49In, and 54Xe were sufficiently lower than their exemption levels at the end of the 5-year-cooling. On the other hand, created 39Y, 45Rh, 50Sn, 52Te, 58Ce, 59Pr, 65Tb, and 66Dy needed additional cooling period less than 10 years until their specific radioactivities decreased below their exemption levels. Then, each additional cooling period required for created 48Cd, 51Sb, 64Gd, and 69Tm was estimated at 10–100 years. Additionally, specific radioactivities of other 13 created elements would not decrease below their exemption levels even if they had been stored for 100 years. There could be significance to create important elements as resources classified in first 4 of the 6 categories defined in this paper, by nuclear transmutation of fission products. In consideration of the efficiency of creation, the radioactivity, and the importance as resources of each product, 44Ru, 46Pd, 52Te, 60Nd, and 66Dy were specially selected as the most important created elements to be more researched in the future Après ORIENT program.

Feasibility study of nuclear transmutation by negative muon capture reaction using the PHITS code

Feasibility of nuclear transmutation of fission products in high-level radioactive waste by negative muon capture reaction is investigated using the Particle and Heave Ion Transport code System (PHITS). It is found that about 80 % of stopped negative muons contribute to transmute target nuclide into stable or short-lived nuclide in the case of 135 Cs, which is one of the most important nuclide in the transmutation. The simulation result also indicates that the position of transmutation is controllable by changing the energy of incident negative muon. Based on our simulation, it takes approximately 8.5 × 10 8 years to transmute 500 g of 135 Cs by negative muon beam with the highest intensity currently available.

LINAC-BASED PHOTONUCLEAR APPLICATIONS AT THE IDAHO ACCELERATOR CENTER

In this paper, current Idaho Accelerator Center (IAC) activities based on the exploitation of high energy bremsstrahlung photons generated by linear electron accelerators will be reviewed. These beams are used to induce photonuclear interactions for a wide variety of applications in materials science, activation analysis, medical research, and nuclear technology. Most of the exploited phenomena are governed by the familiar giant dipole resonance cross section in nuclei. By proper target and converter design, optimization of photon and photoneutron production can be achieved, allowing radiation fields produced with both photons and neutrons to be used for medical isotope production and for fission product transmutation. The latter provides a specific application example that supports long-term fission product waste management. Using high-energy, highpower electron accelerators, we can demonstrate transmutation of radio-toxic, long-lived fission products (LLFP) such as 99Tc and 129I into short lived species. The latest experimental and simulation results will be presented.

Resourceability on nuclear fuel cycle by transmutation approach

A resourceability on nuclear fuel cycle by transmutation of fission products in the spent fuel of nuclear reactors is discussed in this paper to investigate the feasibility of “creation and utilization” of Apres ORIENT from Adv.-ORIENT cycle, in which chemical “separation and utilization” of nuclear rare metals (platinum group metals, Mo, Tc, rare earth, etc.) has been proposed since FY2006. Apres ORIENT research program was newly initiated in FY2011 for nuclear transmutation of fission products into stable or short-lived highly-valuable elements. In the resourceability of rare earth metals from fission products, non-radioactive Nd and Dy can be created from Pr and Tb, respectively, by transmutation. Especially, the Dy creation has a relatively high feasibility of about 10–20 %/y in creation rate. A proper moderation of neutrons in blanket of fast reactors may be required to provide a high creation rate of La from Ba. In light platinum group metals, non-radioactive Ru can be created from Tc by transmutation, of which creation rate is about 4–5 %/y in blanket of fast reactors. Pd created from Rh is almost non-radioactively depending on the isotope fraction of 107Pd. Rh creation from Ru is not feasible under the neutron irradiation of typical nuclear reactors.

Fast Accelerator Driven Subcritical System for Energy Production Using Burned Fuel

Accelerators Driven Systems (ADS) are an innovative type of nuclear system, which is useful for long-lived fission product transmutation and fuel regeneration. The ADS consist of a coupling of a sub-critical nuclear core reactor and a proton beam produced by a particle accelerator. These particles are injected into a target for neutrons production by spallation reactions. The neutrons are then used to maintain the fission chain in the subcritical core.The aim of this study is to investigate the nuclear fuel evolution of a lead cooled accelerator driven system used for energy production and high-level waste transmutation. The fuel studied is a mixture based upon 232Th-233U and Pu-MA extracted from PWR spent. The target is a lead spallation target and the core is filled with a hexagonal lattice. In order to reduce the negative reactivity caused by the presence of protactinium, moderator is not used.In this work is used the Monte Carlo code MCNPX 2.6.0, that presents the depletion/burnup capability, combining a ADS source and kcode-mode. The keff evolution, the neutron energy spectra in the core and the nuclear fuel depletion during the burnup are evaluated. Keywords: ADS, thorium, MCNPX.

Study of an ADS Loaded with Thorium and Reprocessed Fuel

Accelerator-driven systems (ADSs) are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods wasT232hO2forU233production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use ofT232hO2without the initial requirement ofU233enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations). The multiplication factor (keff) evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use ofT232hO2and reprocessed fuel allowedU233production without the initial requirement ofU233enrichment.

Neutron production evaluation from a ADS target utilizing the MCNPX 2.6.0 code

Accelerators Driven Systems (ADS) are an innovative type of nuclear system, which is useful for long-lived fission product transmutation and fuel regeneration. The ADS consist of a coupling of a sub-critical nuclear core reactor and a proton beam produced by particle accelerator. These particles are injected into a target for the neutrons production by spallation reactions. This target is of utmost importance for an ADS, representing the coupling of the accelerator and the sub-critical core. The determination of optimal materials for these targets is fundamental for the design of an ADS. The main characteristic of an ideal target is the high production of neutrons per incident proton. In this work are shown results for the neutron production of various types of targets using the MCNPX 2.6.0 code. Furthermore, it is performed a comparative study of transport models to describe the spallation reactions available in this code.

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the 99TC and 129I and 135Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of 6Li in lithium (Li) and the theoretical density (TD) of Li2O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ≥ 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.

A comprehensive study on neutronics of a lead–bismuth eutectic cooled accelerator-driven sub-critical system for long-lived fission product transmutation

The aim of this study is to investigate the high-level waste (HLW) transmutation and fissile breeding potentials of a lead–bismuth eutectic (LBE) cooled accelerator-driven system (ADS) for the various configurations (the target radius, R T = 10–50 cm and the radial thickness of the sub-critical core, δ SC = 50–80 cm) and for the various fuel compositions (the fuel volume fraction, VF F = 10%, 12%, 15% and 20% and the fissile fraction, FF = 10–24%) under sub-critical condition. The long-lived fission products (LLFPs: 99Tc, 129I and 135Cs nuclides) and the uranium mono carbide (UC) ceramic fuel are considered as the HLW and the fissile fuel, respectively. The neutronic calculations have been performed per the incident proton (1000 MeV) with the high-energy Monte Carlo code MCNPX in coupled neutron and proton mode using the LA150 library. The numerical results bring out that the case of R T = 30 cm, δ SC = 80 cm, VF F = 10% and FF = 23% is the optimum configuration and fuel composition, from the energy gain point of view, and has a high neutronic performance for an effective LLFP transmutation and fissile breeding.

The Cost of Transmutation of Fission Products in Nuclear Reactors

The operating regime of a VVER reactor in which the most important long-lived fission products 99Tc and 129I are transmuted is investigated. Estimates are presented for the decrease in the fuel burnup and decrease in the run time as a result of transmutation. Two methods for inserting the nuclides to be transmuted are examined – by adding to the nuclear fuel or the coolant. It is established that 99Tc and 129I transmutation with the rate of accumulation in a reactor decreases burnup by 5.1 GW·days/metric ton, i.e., by 12.7% of the standard burnup. This corresponds to electricity underproduction 110 GW·days per run or 37 GW·days per year of operation. This result is independent of the method used to insert the nuclides to be transmuted. These energy losses are the price to be paid for transmuting nuclides without removing them during reactor operation.

Accumulation and Transmutation of Spallation Products in the Target of Accelerator-Driven System

Analysis of the radiological burden of the spallation products (SP) was performed for various spallation targets (lead, tungsten, tin). The radiological burden was discussed in terms of toxicity based upon the concept of Annual Limit on Intake (ALI) shows that alpha-emitting rare earths (146Sm, 148Gd, 150Gd, 154Dy) are dominant. Their toxicity was estimated at equilibrium state with and without their transmutation. It is concluded that, in terms of toxicity, accumulation of SP in the target is quite comparable with transmutation of fission products (FP) in the blanket of Accelerator-Driven System (ADS).

Accumulation and Transmutation of Spallation Products in the Target of Accelerator-Driven System.

Analysis of the radiological burden of the spallation products (SP) was performed for various spallation targets (lead, tungsten, tin). The radiological burden was discussed in terms of toxicity based upon the concept of Annual Limit on Intake (ALI) shows that alpha-emitting rare earths (146Sm, 148Gd, 150Gd, 154Dy) are dominant. Their toxicity was estimated at equilibrium state with and without their transmutation. It is concluded that, in terms of toxicity, accumulation of SP in the target is quite comparable with transmutation of fission products (FP) in the blanket of Accelerator-Driven System (ADS).

Fusion-Driven Transmutation of Fission Product Cesium in its Elemental Form

Assuming fission reaction as a dominant energy source for a long-term perspective, the goal of transmutation of fission products is to cut their increasing accumulation and to keep their inventories at easily manageable level. Opposite to relatively short-lived 137Cs (T1/2=30yr) whose natural decay converge equilibrium mass to the level of order of 11 per GW of fission energy, an approach to similar equilibrium inventory for long- lived 135Cs (T1/2=2.3×106 yr) requires artificial transmutation that preassumes its isotopic separation in the most studies. The present paper addresses cesium transmutation without preliminary isotope separation that means an approach to equilibrium for all the isotopes including stable 133Cs. A high-flux blanket driven by Fusion Neutron Source with ITER-like parameters is proposed to transmute cesium in the elemental form. Transmutation efficiency is estimated in terms of equilibrium inventory and characteristic time to reach equilibrium both governed by the mean life-time of nuclides in transmuter. The analytical results show that the mean life-time of the target isotope 135 Cs is as short as 21 yr which is in more than order of magnitude shorter than achieved in advanced fission reactors. It reveals that one Fusion Neutron Source with ITER-like parameters could transmute elemental cesium from 23 PWRs, the fraction of power associated with transmutation being as small as 3%.

On the main objectives of transmutation cycles for long-lived fission products and minor actinides

The present paper makes an effort to look at the problem of radiowaste from the view point of recent trends in fuel cycle development. Assuming nuclear fission as an energy source for a long-term perspective, the obvious goal of transmutation of fission products is to stop their accumulation or, in other words, to approach equilibrium state and keep this at easily manageable level. Characteristic time to approach equilibrium and loses in multirecycling are addressed in this paper as the most important criteria in selecting appropriate transmutation technology. For the minor actinides transmutation is proposed in a way to upgrade the proliferation resistance of the fuel cycles. Analysis given in the present paper reveals the favourable potential of external neutron sources in solving the transmutation problems.

Neutron Excess Generation by Fusion Neutron Source for Self-Consistency of Nuclear Energy System

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities.

Neutron Excess Generation by Fusion Neutron Source for Self-Consistency of Nuclear Energy System.

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities.

Transmutation of fission products by high-field tokamak reactor with force-balanced coils

A DT sub-ignited tokamak reactor is investigated as a transmuter of fission products with an average neutron wall loading of 1 MW m −2 basing on a novel coil concept to generate strong magnetic fields. By the use of the force-balanced coils (FBC) which reduces the net centering force by a factor of ten from that exerted on toroidal magnetic coils of the same dimension, the steady-state tokamak reactor can be designed to achieve the fusion reaction of Q DT≃3.7. To transmute the fission product 129I effectively, a blanket is designed which consists of the neutron multiplier Be, the tritium breeder Li 2O and a neutron reflector C. Using the optimal blanket, the amount of 129I is effectively halved in 2 years with quasi-self-consistent condition, Q NET≃5.3, local TBR>1.7 and nuclear heating rate in superconducting magnets <1 mW cm −3.