From Fission Products Research Articles

Measurement of spent nuclear fuel burn-up using a new H(n, [formula omitted]) method

Measurement of neutrons from spent nuclear fuel is performed in this study using the H(n, γ) method, which detects 2.223 MeV γ rays from neutron capture reaction of hydrogen using a highly pure germanium (HPGe) detector. The detection of the 2.223 MeV γ ray is affected by intense γ ray emission from fission products (FPs) because the emission rate of γ rays from the FP is seven orders of magnitude higher than the emission rate of neutrons. To shield the intense γ ray from the FP, the HPGe detector is placed off the axis of a collimator, whereas a polyethylene block is placed on the axis. In this geometry, the detector is shielded from the intense γ rays from the FP, but the detector can measure 2.223 MeV γ rays from the H(n, γ) reactions in the polyethylene block. The measured count rate of the 2.223 MeV γ rays is consistent with the expected rate within the statistical error, which is calculated based on the nuclide composition, which is primary 244Cm, estimated via depletion and decay calculations. Accordingly, the H(n, γ) method is considered feasible to quantify the number of neutron leakage from spent nuclear fuel assembly, which is applicable to certify burn-up of the assembly.

Quantitative prediction of rare earth concentrations in salt matrices using laser-induced breakdown spectroscopy for application to molten salt reactors and pyroprocessing

Pyroprocessing of spent nuclear fuels is an electrochemical separation method where spent metallic fuel is dissolved in a molten salt bath to allow uranium (U) and plutonium (Pu) to be isolated from fission products (FPs) and other impurities.

Separation and recovery of palladium from nitric acid solution by silica based benzo-15-crown-5 ether resin

The Aprés ORIENT research program, as a newly concept of advanced nuclear fuel cycle, was initiated in FY2011 aiming at creating stable, highly-valuable elements by nuclear transmutation from fission products (FPs). As creation of palladium (Pd) from rhodium (Rh) in FPs is one of the most important targets of the program, the research and development of highly effective separation methods between Rh and Pd are strongly required. In the present paper, an ability of silica based benzo-15-crown-5 ether (B15C5) resin for adsorption and elution of Ru(III), Rh(II), and Pd(II) was investigated in nitric acid (HNO3) solution. As a result of the batch experiment for adsorption, Pd(II) were strongly adsorbed by B15C5 resin in HNO3. On the other hand, Ru(III) and Rh(III) were not adsorbed by the resin. As it was difficult to elute adsorbed Pd(II) by the change of the concentration of HNO3, ethylenediamine (EDA) was employed as an eluent for Pd(II). As a result of the column chromatography for adsorption and elution, only Ru(III) and Rh(III) were eluted in 2 mol/L of HNO3, and then Pd(II) was clearly eluted in a mixed solution including 2 mol/L of HNO3 and 2 mol/L of EDA. Therefore, by using EDA as the eluent, the possibility of B15C5 resin for separation and recovery of Pd created by transmutation of Rh in FPs has been found.

A Simple Method to Create Gamma-ray-source Spectrum for Passive Gamma Technique

A simple but sufficiently accurate method for constructing a gamma-ray source for photon-transportation calculations is developed for canisters containing fuel debris from the Fukushima Daiichi nuclear power plant. The new method couples a baseline spectrum evaluated using the ORIGEN2 code with a line spectrum obtained from fission products (FPs) of interest. One of the advantages of the method is its ability to take bremsstrahlung X rays into consideration using the bremsstrahlung libraries of ORIGEN2. The new gamma-ray source is used to calculate the response of a high-purity germanium (HPGe) detector, and these results are compared with a benchmark test of experimental measurements of irradiated fuel pins. The shape of the simulated gamma-ray spectrum agrees well with that of the spectrum obtained by experiment. Moreover, bremsstrahlung X rays are confirmed to have little effect on spectrum shape formation because the baseline gamma-ray spectrum is mainly formed by Compton scattering of strong gamma rays from FPs such as Cs-137 and Eu-154. Gamma-ray source data are created for a debris composition recently evaluated at JAEA, and subsequent photon-transportation calculation is performed to evaluate the leakage gamma-ray spectra from a canister containing fuel debris for nuclear-material-accountancy purposes.

Stability of isoHex-BTP/SiO2-P adsorbent against acidic hydrolysis and γ-irradiation

Separation of trivalent minor actinides (MA(III): Am(III), Cm(III)) from fission products (FP) in high-level liquid waste (HLLW) is an important task in advanced nuclear-fuel reprocessing systems. For this purpose, an advanced aqueous partitioning process based on extraction chromatography method was studied. Because R-BTP extractants (R-BTP: 2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl)pyridine, R = alkyl group) exhibit high selectivity for MA(III) over trivalent rare-earth elements (RE(III)), a novel adsorbent isoHex-BTP/SiO2-P was prepared by impregnating isoHex-BTP extractant into the macroporous SiO2-P support with a mean diameter of 60 μm. The stability of isoHex-BTP/SiO2-P against nitric acid and γ-irradiation was investigated. It was found that isoHex-BTP/SiO2-P adsorbent shows good adsorption affinity to Dy(III). The hydrolytic and radiolytic stability of isoHex-BTP/SiO2-P adsorbent in 0.01 mol/L HNO3 was fairly promising. However, the adsorption amount Q of Dy(III) decreased dramatically in 3 mol/L HNO3 with the increase of the absorbed dose and became nearly zero at the absorbed dose over 46 kGy. These results suggest that with the synergetic effect of radiation and acidic hydrolysis, the adsorbent instantly loses its efficacy.

Irradiation effect on the structure change for Sr2Fe1.5Mo0.5O6−δ perovskite ceramic

Abstract The incorporation of radioactive elements derived from fission products (FPs) into a stabilized crystalline structure is an active area of research in the nuclear fuel cycle. The compound Sr 2 Fe 1.5 Mo 0.5 O 6− δ (SFM) incorporates several FPs (Sr and Mo) into the crystalline network simultaneously while maintaining a stabilized perovskite structure. This polycrystalline SFM sample was irradiated with 200 keV He ions to a fluence of 5 × 10 16 ions cm −2 (corresponding to a peak dose at 1.2 dpa) at room temperature to study radiation damage effects. Irradiated SFM sample decomposed into a layered Sr 4 FeMoO 8− δ based phase and a Fe metallic phase, resulting in nano-crystalized grains with particle sizes around 7 nm. It was rationalized that He ion irradiation created a reduced atmosphere on the surface of SFM sample by preferentially sputtering oxygen atoms, consequently decomposing SFM and reducing Fe ions into metallic Fe phases. These results suggest that the stability of crystalline structures in reducing atmospheres may be an additional factor for consideration in the search for FP hosts in crystalline materials.

Challenges to develop single-column MA(III) separation from HLLW using R-BTP type adsorbents

In order to directly separate trivalent minor actinides (MA: Am, Cm) from fission products (FP) containing rare earths (RE) in high level radioactive liquid waste (HLLW), the authors have challenged to develop a simplified MA separation process by extraction chromatography using a single column. Attention has been paid to a new type of nitrogen-donor ligands, R-BTP (2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl) pyridine, R: alkyl group) as an extractant because it shows high extraction selectivity for Am(III) over RE(III). It is known that the R-BTP ligands show different properties such as adsorbability and stability by having different alkyl groups. Therefore, some novel adsorbents were prepared by impregnating different types of R-BTP ligands (isohexyl-, isoheptyl- and cyheptyl-BTP) and a similar ligand to the R-BTP, ATP (2,6-bis(1-aryl-1H-tetrazol-5-yl)pyridines), into the porous silica/polymer support (SiO2-P particles). This work deals with comparison in adsorption and desorption properties of Am and some FP in HNO3 solution onto such R-BTP type adsorbents, as well as chemical and radiolytic stability of the adsorbents. Then the possibility of a single-column separation of MA from main FP was pursued by evaluating the results of column experiments using the most promising adsorbent (isohexyl-BTP/SiO2-P) under temperature control. In addition, elution behaviors of U and Pd were also estimated.

Development of a simplified separation process of trivalent minor actinides from fission products using novel R-BTP/SiO2-P adsorbents

From a viewpoint of direct separation of trivalent minor actinides (MA: Am, Cm etc.) from fission products (FP) including rare earths (RE) in high level radioactive liquid waste, the authors have developed a simplified separation process using a single column packed with novel extraction adsorbents. Attention was paid to a new type of nitrogen-donor ligand, R-BTP (2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl)pyridine, R: alkyl group) as an extractant because it has higher extraction selectivity for Am(III) than RE(III). Since the R-BTP ligands show different properties such as adsorbability and stability when they have different alkyl groups, several R-BTP extraction adsorbents were prepared by impregnating the R-BTP ligands with different alkyl groups (isohexyl-, isoheptyl- and cyheptyl-BTP) into a porous silica/polymer composite support (SiO2-P particles). This work investigated: (1) fundamental properties of the synthesized R-BTP/SiO2-P adsorbents, (2) adsorption and desorption properties of Am and FP in nitric acid solution and water using the adsorbents in a batch experiment, (3) radiolytic and chemical stabilities of the adsorbents, and (4) the possibility for developing a simplified separation process of MA using the most promising adsorbent (isohexyl-BTP/SiO2-P) under temperature control between 25 and 50°C.

New approaches to processing and fabrication of oxide nuclear fuels

It was shown that, in contrast to the Purex process using aggressive and environmentally hazardous 8M HNO3 solutions for dissolving spent oxide nuclear fuel (SNF), this fuel can be easily dissolved in aqueous subacid ([H+] ∼0.1 M) solutions of Fe(III) nitrate (chloride) with partial separation of uranium and plutonium from fission products (FP). The low acidity of the solutions obtained (pH ∼1) allows direct application of modern technologies of finishing processing of nuclear fuel by fluoride, carbonate, oxalate, or peroxide precipitation of uranium and plutonium. It was established that U(VI) is isolated from nearly neutral nitric acid solutions as a poorly soluble uranyl hydroxylaminate complex after adding hydroxylamine. It was shown that on thermal decomposition at 200–300°C under ambient atmosphere this compound converts into uranium dioxide. A similar approach was applied to obtain mixed oxide uranium-plutonium fuel (MOX fuel).

Recovery of Transuranium Elements from Real High-Level Liquid Waste by Pyropartitioning Process

A pyropartitioning process is under development to recover minor actinide elements from high-level liquid waste (HLLW) generated by Purex reprocessing. This pyropartitioning process consists of a denitration step that converts various elements in the HLLW into oxides, a chlorination step that converts the oxides into chlorides, and a reductive-extraction step that separates the actinide elements from fission products (FPs) in the chlorination product. The feasibility of each step was confirmed using simulating FPs and unirradiated transuranium elements (TRUs). In the present study, approximately 520 g of real HLLW was prepared to demonstrate the feasibility of the pyropartitioning process. Almost 100% of each TRU originally contained in the HLLW was recovered in the liquid cadmium phase in the reductive-extraction step, which showed that the expected chemical reactions were completed and the mass loss of TRUs was negligible in the denitration, chlorination, and reductive-extraction steps. The separation behaviors of actinide elements, including americium and curium, from FPs in the reductive-extraction step were quite similar to those observed in previous experiments using unirradiated materials. Hence, the pyropartitioning process was successfully verified.

Recovery of Transuranium Elements from Real High-Level Liquid Waste by Pyropartitioning Process

A pyropartitioning process is under development to recover minor actinide elements from high-level liquid waste (HLLW) generated by Purex reprocessing. This pyropartitioning process consists of a denitration step that converts various elements in the HLLW into oxides, a chlorination step that converts the oxides into chlorides, and a reductive-extraction step that separates the actinide elements from fission products (FPs) in the chlorination product. The feasibility of each step was confirmed using simulating FPs and unirradiated transuranium elements (TRUs). In the present study, approximately 520 g of real HLLW was prepared to demonstrate the feasibility of the pyropartitioning process. Almost 100% of each TRU originally contained in the HLLW was recovered in the liquid cadmium phase in the reductive-extraction step, which showed that the expected chemical reactions were completed and the mass loss of TRUs was negligible in the denitration, chlorination, and reductive-extraction steps. The separation behaviors of actinide elements, including americium and curium, from FPs in the reductive-extraction step were quite similar to those observed in previous experiments using unirradiated materials. Hence, the pyropartitioning process was successfully verified.

｀２３５´Ｕ，｀２３９´Ｐｕおよび｀２３３´Ｕの高速中性子核分裂による核分裂生成物からのγ線崩壊熱の測定

Gamma decay heat released from fission products (FPs) has been measured for fast-neutron fissions of 235U, 239Pu and 233U using the radiation spectrometry method. The sample irradiations were for 10 and 100s in the fast neutron source reactor YAYOI of the University of Tokyo. Spectral data for γ-ray were obtained at post-irradiation time intervals ranging 11-25, 950s using a NaI (Tl) scintillation detector. The data were processed to the form of γ-ray energy release rates per fission for each set of time-interval parameters. The standard representation of the decay heat following fission pulse (in cooling times ranging 19-24, 000s) was provided from the γ-ray energy release rates. The uncertainties (1σ) of the decay heat data were about 5% for 235U and 239Pu, and about 7% for 233U.The data obtained from the present experiments are compared with three summation calculations using JNDC, TASAKA and ENDF/B-IV decay data libraries, and with other experimental results. The summation calculation results using the JNDC FP decay data library are in better agreement with the present data than the other calculations.