Long-lived Fission Products Research Articles

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.

Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of 229Th

In the epi-cadmium neutron-induced fission of 229Th, cumulative yields of relatively long-lived fission products within the mass range of 77 to 151 were measured by using an off-line gamma-ray spectrometric technique. The mass yields were obtained from the cumulative fission product yields by using charge distribution correction. The peak-to-valley (P/V) ratio, full-width at tenth-maximum of light and heavy mass wings, average light mass <AL> and heavy mass <AH>, and average neutron number <ν> were obtained. The P/V ratio was obtained for the first time and was found to be about three times lower in the epi-cadmium neutron fission than in the thermal neutron fission of 229Th, which shows the role of excitation energy. The fine structure of the mass yield distribution in the 229Th(nf,f) reaction was explained from the viewpoint of nuclear structure effect and the Standard I and Standard II asymmetric modes of fission.

137Cs direct measurement in water via LSC techniques

This work explores the screening technique of 137Cs concentration dissolved in water via the Cherenkov counting method on a Liquid Scintillation Counter Quantulus 1220. In the event of a nuclear emergency or the potential radioactive leakage from the power plants, the prompt detection of two long-lived fission products, 90Sr/90Y and 137Cs, is of the greatest radiological concern. Here we present an innovative, rapid, and simple LSC method for 137Cs content determination that enables quick evaluation of the radioactive contamination status of water samples without any sample pre-treatment. The novelty in the presented Cherenkov counting technique is the addition of ionic liquid 1-butyl-3-methylimidazolium salicylate in a sample vial. Its presence in the amount of ∼1 g increases efficiency up to ∼30% via a wavelength-shifting mechanism, thus reducing the detection limit of the method more than 4 times. Without the ionic liquid's addition, the method's main parameters were detection efficiency of 6.98(11)% and a detection limit of 2.6 Bq L−1 for 300 min of counting in a 20 mL sample volume. Interference with 90Sr/90Y, a radionuclide that can contribute to the Cherenkov radiation spectrum, was also investigated. The method of distinguishing individual radionuclides (137Cs and 90Sr/90Y) in the cumulative Cherenkov spectrum is presented as well.

Uncertainty estimation in neutron TOF measurements with ANNRI

To improve data accuracy of neutron-capture and total cross sections of minor actinides and long-lived fission products, the time-of-flight experimental instrument named ”Accurate Neutron-Nucleus Reaction measurement Instrument” was constructed, and neutron time-of-flight experiments have been performed with an intense pulsed-neutron beam at MLF in J-PARC. Detailed analysis methods and deduced uncertainties for capture and total cross-section measurements are presented using an example of the experiments of 243Am. The deduced uncertainties are categorized into five types.

Verification of Bison fission product species conservation under TRISO reactor conditions

When assessing the reliability and predictive capabilities of a simulation tool, code verification is used to ensure that the implemented numerical algorithm is a faithful representation of its underlying mathematical model, including partial differential or integral equations, initial and boundary conditions, and auxiliary relationships. During this process, numerical results in a discrete solution are compared to the analytical solution of the mathematical model. In this paper, the code verification process is applied to one-dimensional spatiotemporal problems that exercise partial differential equation governing the conservation of fission product species (or mass diffusion). Numerical experiments were performed in the Bison fuel performance code to evaluate its predictive capability under various TRISO reactor conditions such as base irradiation and safety heating test conditions for either short- or long-lived fission product species, as well as a case concerning evaporation from the outer surface of a particle. The code predictions were compared with the expected exact results obtained from the analytical expressions, and the fact that they demonstrate the correct analytical behavior provides strong evidence of proper numerical algorithm implementation.

Assay of nuclear grade primary sodium in the fast breeder test reactor for radionuclides and corrosion elements

Periodic assay of primary sodium used as a coolant in the fast breeder test reactor (FBTR) at Kalpakkam is necessary and pre-requisite for the safe and continuous operation of the reactor. Contamination of nuclear grade purity of the primary sodium is monitored either online or offline for the radionuclides and chemical impurities to ascertain the integrity of the various components. Chemical elements which can entrain into sodium due to the corrosion of structural materials is troublesome as they might clog the coolant channels resulting in to a reduced coolant flow which further leads to fuel clad failure. The corrosion elements present in the primary sodium were determined by ICP-OES and their activation products were assayed by high resolution gamma spectrometry. In case of fuel clad rupture, the fission products such as 137Cs, 134Cs, 136Cs, 131I, 90Sr, 125Sb etc. along with actinides U and Pu are released into primary sodium. This paper reports actinides, fission products, corrosion elements as well as their activation products and concentration levels present in the primary sodium of FBTR. Activity variation of one of the long-lived fission products, viz. 137Cs over the period of time, before and after fuel clad rupture is profiled and discussed.

Sn(IV) Sorption onto Illite and Boom Clay: Effect of Carbonate and Dissolved Organic Matter

126Sn is a long-lived fission product and it is important to assess its sorption onto the host rocks surrounding a possible nuclear waste repository. Boom Clay (BC) is under investigation in Belgium as a potential host rock. To better understand Sn(IV) sorption onto the clay minerals constituting BC, sorption of Sn(IV) was here investigated on Illite du Puy (IdP), from pH 3 to 12. Sorption isotherms at pH ~8.4 were acquired in the presence and absence of carbonate, and in the presence and absence of BC dissolved organic matter (DOM). Sn(IV) strongly sorbed on IdP over the full range of the pHs and concentrations investigated. In the presence of carbonates, Sn(IV) sorption was slightly decreased, highlighting the Sn(IV)–carbonate complexation. DOM reduced the Sn(IV) sorption, confirming the strong complexation of Sn(IV) with DOM. The results were modelled with the 2-site protolysis non-electrostatic surface complexation model. The surface complexation constants and aqueous complexation constants with carbonate and DOM were optimized to describe the experimental data. The applicability of the component additivity approach (CAA) was also tested to describe the experimental Sn(IV) sorption isotherm acquired on BC in BC pore water. The CAA did not allow accurate prediction of Sn(IV) sorption on BC, highlighting the high sensitivity of the model to the Sn(IV)-DOM complexation.

МЕДИКО-САНИТАРНЫЕ ПОСЛЕДСТВИЯ РАДИАЦИОННЫХ АВАРИЙ ПРИ ОБРАЩЕНИИ С ОТРАБОТАННЫМ ЯДЕРНЫМ ТОПЛИВОМ

Purpose: To consider the features of potential health effects for personnel and for the population in the event of radiation accidents at facilities for post-reactor spent nuclear fuel management (including spent nuclear fuel pools and general plant storage facilities, transportation of nuclear materials and their subsequent processing at radiochemical plants). Results: Based on the analysis of publications, the potential radiation consequences for personnel and for the population in the event of various types of radiation accidents at the stages of spent nuclear fuel management were systematized. Conclusion: Advanced technologies underlying the closed nuclear fuel cycle provide a high level of radiation safety for personnel and for the population. At the same time, a large amount of accumulated radioactive and nuclear materials makes it necessary to maintain the health care system in preparedness at all stages of spent nuclear fuel management. Health effects for personnel may be associated with exposure because of criticality accident and internal intake of fission products of uranium and actinides (by inhalation and through wound surfaces). In case of accidents at radiochemical production, combined radiation-thermal and radiation-chemical injuries are also possible. The main potential hazard for the population in the event of radiation accident at the spent nuclear fuel storage facility is the contamination of the environment and exposure to long-lived uranium fission products and actinides. This requires clarification and development of appropriate criteria and derived intervention levels for making decisions on protective measures. Hypothetically, the health effects in case of radiation accident when managing “fresh” spent nuclear fuel can be comparable to a large-scale reactor accident and require urgent protective measures, including evacuation and iodine prophylaxis. An important factor that should also be taken into account when planning health care measures is the need for decontamination of victims.

Evaluation of transmutation rate of some LLFP in experimental fast reactor JOYO

A transmutation process of three long-lived fission products (79Se, 99Tc and 107Pd) in the experimental fast reactor JOYO is postulated. The possibility of increasing the transmutation rate utilizing the high neutron flux present in the JOYO reactor by loading neutron-moderating subassemblies in the reflector zone has been investigated. A cluster of reflector subassemblies was replaced with beryllium or zirconium hydride (ZrH1.65) moderated subassemblies. These moderated subassemblies surrounded one central test subassembly that would contain the three long-lived fission products (LLFP) simultaneous and without isotopic separation. ChainSolver 2.34 code is used to calculate the transmutation rates. In this study, the new characteristics of LLFP transmutation in a fast reactor using moderator materials were shown for future applications.

Microplastics as vectors of radioiodine in the marine environment: A study on sorption and interaction mechanism

Radioiodine is one of the long-lived fission products and also an important radionuclide released during nuclear accidents, which generates interest in its environmental fate. Its sorption has been studied in a wide range of materials, but no equivalent study exists for microplastics, an emerging environmental vector. Weathering and biofilm formation on microplastics can enhance radioiodine sorption. For the first time, we're reporting how radioiodine interacts with different types of polyethylene derived microplastics (pristine, irradiated, and biofilm developed microplastics). This study revealed that exposure to radiation and the marine environment significantly alters the physico-chemical properties of microplastics. In particular, in marine-exposed samples, a signature of biofilm development was detected. Speciation study indicates that iodine exists in the iodide form in the studied marine environment. The study revealed that, iodide ions attach to biofilm-developed microplastics via electrostatic, ion-dipole, pore filling, and van der Waals interactions. Pore filling, ion-dipole, and van der Waals interactions may cause iodide binding to irradiated microplastics, whereas pore-filling and van der Waals interactions cause iodide binding to pristine microplastics. The distribution coefficient (Kd) of iodine on microplastics is positively correlated with biofilm biomass, which signifies the role of biofilm in radioiodine uptake. The Kd indicates microplastics are potential iodide accumulators and could be a possible vector in the marine system.

Test of 107Pd transmutation with macroscopic quantities

ABSTRACT A test of 107Pd transmutation with macroscopic quantities was conducted. Palladium is an essential material for industrial use and is one of the nuclides targeted for recycling from spent nuclear fuel. Reasonable nuclear reaction paths for 107Pd and the cross-sections for proton- and deuteron-induced spallation in inverse kinematics have been investigated. However, a transmutation experiment using long-lived fission products (LLFPs) as the target would be required for an actual system. To effectively detect the reaction products of the 107Pd + d reaction, we prepared a material with a 107Pd-concentration of almost 100% by ion implantation. The implanted samples were irradiated for several days with deuterons produced by the AVF Ring Cyclotron at RIKEN RIBF. After cooling, gamma-ray measurements of the irradiated sample were conducted. These revealed 105Ag and 106mAg originating from the 107Pd + d reaction. Furthermore, 105Pd and 106Pd produced from transmutation of 107Pd were estimated using DCHAIN. The 107Pd in the irradiated samples were measured by ICP-MS. The isotopic ratios 105Pd/107Pd and 106Pd/107Pd obtained from the experimental results are consistent with those obtained by calculation using PHITS. Furthermore, the 107Pd transmutation using deuteron was verified.

Concept of a small fast breeder reactor system capable of transmuting MA and LLFP

The issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive wastes. Important radioactive wastes are minor actinides (MA : 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFP : 129I, 99Tc, 79Se, etc.). The purpose of this study was to construct a system that simultaneously achieves the breeding of fissile materials and the transmutation of MA and LLFP using small fast reactors. Transmutation was achieved by loading the innovative Duplex type MA-containing fuel in the core region and the LLFP-containing moderator in the first layer of the radial blanket. Breeding of the entire small fast reactor system was achieved by combining two small fast reactors: one for transmutation and the other for breeding. As a result, it was clarified that in this small fast reactor system, a breeding ratio of approximately 1.03 was obtained, and support ratios (SRs) of 2 or more were achieved for MAs and LLFPs.

Transmutation of long-lived fission products in an advanced nuclear energy system

Disposal of long-lived fission products (LLFPs) produced in reactors has been paid a lot attention for sustainable and clean nuclear energy. Although a few transmutation means have been proposed to address this issue, there are still scientific and/or engineering challenges to achieve efficient transmutation of LLFPs. In this study, we propose a novel concept of advanced nuclear energy system (ANES) for transmuting LLFPs efficiently without isotopic separation. The ANES comprises intense photoneutron source (PNS) and subcritical reactor, which consist of lead–bismuth (Pb-Bi) layer, beryllium (Be) layer, and fuel, LLFPs and shield assemblies. The PNS is produced by bombarding radioactive cesium and iodine target with a laser-Compton scattering (LCS) γ-ray beam. We investigate the effect of the ANES system layout on transmutation efficiency by Monte Carlo simulations. It is found that a proper combination of the Pb-Bi layer and the Be layer can increase the utilization efficiency of the PNS by a factor of ~ 10, which helps to decrease by almost the same factor the LCS γ-beam intensity required for driving the ANES. Supposing that the ANES operates over 20 years at a normal thermal power of 500 MWt, five LLFPs including 99Tc, 129I, 107Pd, 137Cs and 79Se could be transmuted by more than 30%. Their effective half-lives thus decrease drastically from ~ 106 to less than 102 years. It is suggested that this successful implementation of the ANES paves the avenue towards practical transmutation of LLFPs without isotopic separation.

Caesium and iodine release from spent mixed oxide fuels under repository relevant conditions: Initial leaching results

Autoclave leaching experiments are conducted on three well-characterised, irradiated, and cladded mixed oxide fuel-rod segments with burnups ranging from 29 GWd/tHM to 52 GWd/tHM to investigate the instant release fraction of fission gases and long-lived fission products and to assess the long-term fuel matrix corrosion. The segments are exposed to bicarbonate solutions as reference groundwater at neutral pH and a synthetic young cementitious water at pH 13.5 under reducing atmosphere (4 vol% H2 in Ar at 40 bar pressure), since 2018. The initial leaching results for the fission products caesium and iodine as representative elements of the instant release fraction were found to depend on the leachate composition as well as on the fuel burnup.Graphical abstract

Inert-matrix fuel for transmutation: Selected mid- and long-term effects on reprocessing, fuel fabrication and inventory sent to final disposal

Partitioning and transmutation (P&T) fuel cycles provide a technical approach to ease the problem of radioactive waste disposal. Some of the partitioned components of the waste stream are irradiated while others can be used for energy production or are sent to final storage. Minor actinides are planned to be irradiated in a fast spectrum nuclear facility to transmute them into stable or short-lived isotopes. As minor actinides have negative effects on reactor dynamics, subcritical, accelerator-driven systems are proposed to increase their fraction in the fuel. An example is the MYRRHA research reactor to be built in Mol, Belgium.This reactor is modeled for depletion calculations. The behavior of special fuel elements that mirror fuel composition as envisioned for large scale transmutation facilities, namely inert-matrix fuels with an increased minor actinide content, are investigated within this reactor environment. It turns out that gamma dose rates, activity and residual heat from the spent fuel elements present significant challenges for implementing a P&T program. Spent inert-matrix fuel element show significantly higher levels than spent fuel elements from fast reactors. This requires long cooling periods and poses unprecedented challenges to reprocessing technology. The problem is amplified by the fact that it is generally agreed upon that due to low transmutation efficiencies several transmutation steps would be necessary. Looking at the radiotoxicity index, the efforts suggested to reduce the minor actinide content in a final repository are justified. The long-term safety case of deep geological repositories, however, implies that certain long-lived fission products are more relevant. The build-up of some of these radionuclides is investigated for two hypothetical German P&T scenarios. Naturally, the amount of fission products increases with continued irradiation. But namely the fraction of Cs-135 increases over-proportionally when inert-matrix fuel rich on minor actinides is used.

A hybrid (γ, n) and (n, γ) transmutation study for long-lived fission products and thermal analysis of converter cooling system

A hybrid (γ, n) and (n, γ) transmutation study for long-lived fission products and thermal analysis of converter cooling system

О ТРАНСМУТАЦИИ МИНОРНЫХ АКТИНОИДОВ В ЖИДКОСОЛЕВОМ РЕАКТОРЕ-СЖИГАТЕЛЕ

Reduction in the amount of accumulated minor actinides (Np, Am, and Cm isotopes), as well as, the long-lived fission products, to the extent possible, is a top-priority task for the nuclear power engineering. Transmutation of this waste in a special-purpose molten-salt burner reactor has been currently proposed as a possible solution for this problem. The objective of the present paper is to investigate basic regularities of Np, Am, and Cm transmutation in the molten salt burner reactor and to determine optimal operational conditions of the reactor. In the optimal equilibrium mode just fluorides of these elements are added into the fuel composition but the fission products are extracted from it. This mode are reached by maintaining a specified actinide concentration. In order to provide the reactor criticality with a lower than optimum actinide concentration, it is required to add plutonium in the feed fuel instead of a certain amount of minor actinides, thus impairing the transmutation efficiency. To keep the critical state in the equilibrium mode, in situations where the actinide concentration is higher than the optimum one, it is essential to extract a portion of fuel with high content of 238Pu. It has been shown that the reactor configuration is the primary factor responsible for optimum actinide concentration value and that this value varies little with the feed fuel composition, the type of salt dissolving agent, and the fuel reprocessing mode. Optimal concentration for the molten-salt burner reactor with the core volume ranging from 2 to 30 m3 is within 17…10 mole %.

Hydrogen dioxide-boosted acid elution of radionuclides from high level vitrified wastes

• A wet chemical elution was used to extract radionuclides from nuclear waste glass. • H 2 O 2 largely enhanced the elution capability of HNO 3 under mild conditions. • The insoluble Zr was well extracted by H 2 SO 4 solution through coordination. • The PGMs were extracted using liquid Sn extraction under reductive heat-treatment. The extraction of radionuclides from high-level vitrified wastes (HLVWs) is of importance for eventual harmlessization of nuclear wastes by nuclear transmutation, but remains great challenges. Here, we showed an efficient elution method of various radionuclides from the simulated HLVWs with chemical durability using the mixed H 2 O 2 /HNO 3 solutions under mild conditions. The introduction of H 2 O 2 largely improved the elution capability of HNO 3 solution toward radionuclides contained in HLVWs. Most elements including glass main components (except the insoluble Si) and the simulated radionuclides (except for Zr and platinum group metals (PGMs: Pd, Ru, Rh)) were significantly leached from the HLVWs with the mixed solution of 1 ml H 2 O 2 /1 M HNO 3 at 90 °C for 5 h. The insoluble Zr and PGMs were further extracted by H 2 SO 4 solution and liquid tin extraction, respectively. Through optimization of the extraction conditions, the proposed wet chemical elution method and high temperature metallurgy to extract the elements from the HLVWs would open a new avenue to reduce the radioactivity and lifetime of the long-lived fission products (LLFPs) and minor actinides (MAs) in HLVWs through advanced nuclear transmutation in future. It also provides a feasible method for the reuse of some valuable radionuclides in HLVWs.

Design of Cesium target assembly with YH2 moderator and Gd thermal neutron shielding to produce 135Cs for LLFP transmutation study using the experimental fast reactor Joyo

In this design study, we investigated the feasibility of producing a sufficient amount of 135Cs long-lived fission product (LLFP) in the Joyo fast reactor by irradiation of natural cesium (133Cs). The irradiation is expected to contribute to providing 135Cs sample for transmutation cross-section measurements for enhancing the accuracy of the LLFP nuclear data. Two irradiation fields, one located in the outer reflector and another one in the inner reflector which was adjacent to the core were investigated, and appropriate target subassemblies were designed by using the MCNP and FISPACT-II codes. In the inner reflector case, the YH2 moderator material was adopted for the neutron spectrum adjustment within the target subassembly. The gadolinia thermal neutron shielding is proposed to suppress the power peak of the adjacent fuel assembly. The result confirmed that a sufficient amount of 135Cs can be produced at 1200 Effective Full Power Day (EFPD) in Joyo reactor (Mark IV core, 100 MWth).

Concept of a fast breeder reactor to transmute MAs and LLFPs

The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.