Experimental measurements of short-lived fission products from uranium,neptunium, plutonium and americium

Fission yields are especially well characterized for long-lived fission products. Modeling techniques incorporate numerous assumptions and can be used to deduce information about the distribution of short-lived fission products. This work is an attempt to gather experimental (model-independent) data on short-lived fission products. Fissile isotopes of uranium, neptunium, plutonium and americium were irradiated under pulse conditions at the Washington State University 1 MW TRIGA reactor to achieve ~10⁸ fissions. The samples were placed on an HPGe (high purity germanium) detector to initiate counting in less than 3 min post irradiation. The data was analyzed to determine which radionuclides could be quantified and compared to the published fission yield data.     

Fluoride volatility method for reprocessing of LWR and FR fuels

Fluoride volatility method is based on direct fluorination of powdered spent fuel with fluorine gas in a flame fluorination reactor, where the volatile fluorides (represented mainly by UF₆, partially NpF₆) are separated from the non-volatile ones (e.g. PuF₄, AmF₃, CmF₃, fluorides of majority of fission products), the objective being to separate a maximum fraction of uranium component from plutonium, minor actinides and fission products. The current research and development work in the area of fluoride volatility method is focused on the experimental program carried out at the semi-technological line called FERDA, which is a follow-up of the previous FREGAT-2 technology. The experimental test program, launched in 2004 by the Nuclear Research Institute ?e? plc, has been focused mainly to the study of flame fluorination process, which is considered to be the crucial unit operation of the technology. The fluorination experiments were realized in the first instance with pure uranium oxide fuel and later on with simulated spent oxide fuel. Follow-on tests are planed with oxide fuels with inert matrixes. The experimental program is further supplemented by the system studies focused mainly to the process flow-sheet design and calculations and to the requisite modification of some apparatuses for the future verification of the process with irradiated fuel in hot conditions.     

The EURO-GANEX Process: Current Status of Flowsheet Development and Process Safety Studies

A new hydrometallurgical grouped actinide extraction process has been developed to separate the transuranic actinide ions from dissolved spent fuel solution (after an initial uranium extraction cycle). This “EURO-GANEX” process is aimed towards the homogeneous recycling of plutonium and minor actinides in a future closed fuel cycle. The separation process is based on the co-extraction of actinides and lanthanides from aqueous nitric acid into an organic phase followed by selective co-stripping of actinides. A suitable organic phase has been formulated and distribution ratios determined for lanthanides, actinides and some problematic fission products under extraction and stripping conditions. The process flowsheet has been proven on surrogate feed solutions as well as with spent fast reactor fuel; excellent recoveries of the actinides and good decontamination factors from the lanthanides and other fission products were obtained. A variation on the EURO-GANEX flowsheet (the “TRU-SANEX” process) has now been designed to produce separate Pu+Np and Am+Cm products for heterogeneous recycling. Progress on underpinning process chemistry and safety studies as well as flowsheet tests are summarized.     

New reprocessing system for spent nuclear reactor fuel using fluoride volatility method

Our proposed spent nuclear fuel reprocessing technology named FLUOREX, which is a hybrid system using fluoride volatility and solvent extraction, meets the requirements of the future thermal/fast breeder reactors (coexistence) cycle. We have been done semi-engineering and engineering scale experiments on the fluorination of uranium, purification of UF₆, pyrohydrolysis of fluorination residues, and dissolution of pyrohydrolysis samples in order to examine technical and engineering feasibilities for implementing FLUOREX. We found that uranium in spent fuels can be selectively volatilized by fluorination in the flame type reactor, and the amount of uranium volatilized is adjusted from 90% to 98% by changing the amount of F₂ supplied to the reactor. The volatilized uranium is purified using UO₂F₂ adsorber for plutonium and purification methods such as condensation and chemical traps for fission products provide a decontamination factor of over 10⁷. Most of the fluorination residues that consist of non-volatile fluorides of uranium, plutonium, and fission products are converted to oxides by pyrohydrolysis at 600-800 °C. Although some fluorides of fission products such as alkaline earth metals and lanthanides are not converted completely and fluorine is discharged into the solution, oxides of U and Pu obtained by pyrohydrolysis are dissolved into nitric acid solution because of the low solubility of lanthanide fluorides. These results support our opinion that FLUOREX has great possibilities for being a part of the future spent nuclear fuel cycle system.     

Fission yield measurements by inductively coupled plasma mass-spectrometry

Correct prediction of the fission products inventory in irradiated nuclear fuels is essential for accurate estimation of fuel burnup, establishing proper requirements for spent fuel transportation and storage, materials accountability and nuclear forensics. Such prediction is impossible without accurate knowledge of neutron induced fission yields. The uncertainty of the fission yields reported in the ENDF/B-VII.0 library is not uniform across all of the data and much of the improvement is desired for certain fissioning isotopes and fission products. We discuss our measurements of cumulative fission yields in nuclear fuels irradiated in thermal and fast reactor spectra using Inductively Coupled Plasma Mass Spectrometry.     

A fast separation for antimony produced in heavy-ion induced fission

Fission fragments from heavy ion induced fission were stopped in thin magnesium foils. A fast procedure based on evolution of stibine was developed to separate the antimony isotopes embedded in the foil. A separation system, and a glass pressure filtration system was constructed for this purpose. The chemical yield measured by three independent methods was 80–90%. The degree of decontamination from other fission products was >10². The whole separation took eight minutes.     

Hydrometallurgical Reprocessing of BREST-OD-300 Mixed Uranium-plutonium Nuclear Fuel

The duration of external fuel cycle of BREST-OD-300 reactor with mixed U-Pu nitride fuel (MNIT) including hydrometallurgical reprocessing should not exceed 3 years. An average burnup of the fuel should be 6% of heavy metal (HM) with the potential increase up to 10% HM. Therefore, the technology should provide the reprocessing of spent nuclear fuel (SNF) after less than 2 years cooling time and with fissile materials (FM) content of 10 – 15%. Pellets technology has been chosen for the MNIT fuel production. That means necessity to receive the recycled actinides oxides of high purification coefficient (∼ 10⁶). Currently on a laboratory scale, the following process stages have been tested on the real products: actinide oxides production and rare-earth and trans-plutonium elements separation. Moreover, on a pilot scale the process of high level radioactive waste (HLW) and intermediate level radioactive waste (ILW) concentration by evaporation has been tested, as well as the Am-Cm separation. In 2015, the design of the MNIT SNF reprocessing facility has been started, placed at the JSC Siberian Chemical Plant site as a part of the pilot demonstration power complex (PDPC) with BREST-OD-300 reactor. MNIT SNF reprocessing plant (RP) should be put in operation after 2020.     

燃料电池反应器在化学品与电能共生中应用

燃料电池作为能源转换装置能够高效地将化学能转化为电能,随着技术的发展人们将其作为反应器来完成高附加值的化学品的合成,同时产生一定的电能. 燃料电池反应器因具有反应条件温和、反应过程可控、产物选择性高、能源利用效率高等特点,而被广泛地应用于医药中间体的制备、气体分离、水处理等多个领域. 本文首先按照反应器中阴阳极区域发生反应的类型进行分类,介绍燃料电池反应器在化学品与电能联产中的研究现状和研究进展. 随后描述了燃料电池反应器中存在的问题,并依照催化剂、反应过程等方向对解决方案进行探讨. 最后,对几种新型燃料电池反应器的研究进行了简要的介绍并对其发展做出了展望.     

Transuranium elements and fission products in technological channels of unit No. 2 of Chernobyl Nuclear Power Plant

A gamma-spectrometric and radiochemical analysis was carried out of a material (Zr-2.5% Nb alloy) of technological channels (TC) of Chernobyl Nuclear Power Plant (NPP) power unit No. 2 RBMK-1000 reactor, being under the beginning stage of decommissioning. Activities of ⁹⁰Sr, ¹³⁷Cs, ²³⁸Pu, ^239+240Pu, ²⁴¹Am and ²⁴⁴Cm were determined. It was established that the main source of the revealed actinides and fission products was an impurity of natural thorium and uranium in TC source material on the level of several tenths of ppm. Impurity analysis of TC source material was performed by neutron activation analysis (NAA) and inductively coupled plasma mass spectrometry (ICP-MS). Fission product and transuranium element activities measured were compared with the results of prognostic calculations performed with the help of the NAAPRO code.     

Reactor physic and reprocessing scheme for innovative molten salt reactor system

The molten salt reactor is one of the six concepts retained by the Generation IV forum in 2001. Based on the MSRE and MSBR concepts developed by ORNL in the 60s which involve a liquid fuel constituted of fluorine molten salt at a temperature close to 600 °C, new developments with innovative approach and technology have been realized which contribute to strongly improve the concept. The thorium breeder potentiality is closely related to the use of a liquid fuel which is able to be periodically treated. A reprocessing scheme has been established to treat used fuel by extraction of fission products. According to the Gen IV philosophy for closed cycle nuclear reactor, the actinides are sent back in the reactor core. In this way, the wastes radiotoxicity is strongly decreased and the use of natural resource is optimized. This paper describes an innovative reactor concept, the TMSR-NM (non-moderated thorium molten salt reactor), from the nuclear physic point of view and the different steps involving in the reprocessing scheme from the chemical point of view.     

Utilization of a mixed-bed column for the removal of iodine from radioactive process waste solutions

A boron carbide capsule was previously designed and tested by Pacific Northwest National Laboratory (PNNL) and Washington State University (WSU) for spectral-tailoring in mixed spectrum reactors. The presented work used this B₄C capsule to create a fission product sample from the irradiation of highly enriched uranium (HEU) with a fast fission neutron spectrum. An HEU foil was irradiated inside of the capsule in WSU’s 1 MW TRIGA reactor at full power for 200 min to produce 5.8 × 10¹³ fissions. After 3 days of cooling, the sample was shipped to PNNL for radiochemical separations and analysis by gamma and beta spectroscopy. Fission yields for products were calculated from the radiometric measurements and compared to measurements from thermal neutron induced fission (analyzed in parallel with the non-thermal sample at PNNL) and published evaluated fast-pooled and thermal nuclear data. Reactor dosimetry measurements were also completed to fully characterize the neutron spectrum and total fluence of the irradiation.     

Hot particles analysis originating from failed and damaged fuels

The increase of activities of fission products and transmutation products in the primary coolant of a nuclear power plant indicates the presence of fuel rod failures. The measurement of the activity concentration of the primary coolant was able to detect fuel failures in the reactor core. Microanalytical methods for examining individual hot particles have been developed and applied to fuel failure detection under normal operation conditions as well as during the severe fuel damage that occurred in the cleaning tank incident at Unit 2 of NPP Paks in April 2003. Several faulty fuel rods can be detected simultaneously by the characterization of individual hot particles originating from the primary water. The analysis of particles originating from the damaged fuels provides information relating to the dissolution process of the fuel debris.     

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 Après 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.Après 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 107 Pd.Rh creation from Ru is not feasible under the neutron irradiation of typical nuclear reactors.     

Nuclear decay studies of fission-product nuclides using an on-line mass separation technique

An isotope-separator-on-line (ISOL) system has been developed at the Idaho National Engineering Laboratory to enable a wide variety of nuclear decay studies to be made for fission-product radionuclides. The system is unique in that it utilizes the spontaneous fission source,²⁵²Cf, as the source of fission-product radioactivity. Fission products are transported to the ion source of the mass separator by the helium gas-jet technique. Mass-separated beams of previously unattainable rare-earth nuclides are produced with this system because of the higher yield of fission products with A>150, relative to that for thermal-neutron fission of²³⁵U, and the use of a relatively efficient ion source. Recent decay studies reported here include systematic measurements of rare-earth nuclide half-lives and comparison of them to theoretical prediction, a decay scheme investigation for¹⁵⁴Nd, and -strength function measurements for¹⁴⁰Cs.     

Technetium chemistry in a novel group actinide extraction process

A newly developed method for advanced reprocessing of used nuclear fuel is the Group ActiNide EXtraction (GANEX) process. It is a liquid–liquid extraction process that aims at extracting all the actinides as a group from dissolved used nuclear fuel. This extraction can either be performed after a removal of the bulk uranium or directly on the dissolution liquor. At Chalmers University of Technology in Sweden a solvent that utilizes tributyl-phosphate (TBP) and a molecule from the bis-triazine bipyridine (BTBP) class of ligands dissolved in cyclohexanone has been developed for the use in a GANEX process. Previously the system has not been tested with the presence of technetium that is one of the major fission products. Technetium is often considered a problem within reprocessing since it has a chemical behaviour that differs from most other elements in the spent fuel. Therefore, a special emphasis was put on the investigation of technetium in the selected GANEX system. It was shown that technetium is readily extracted by the GANEX solvent and that cyclohexanone is the main extractant when no other metals were present in the system. It was also found that the presence of uranium decreased the overall technetium extraction despite a slight co-extraction with TBP, while irradiation of the GANEX solvent to large doses (>1 MGy) increased its technetium extraction capability. It was also discovered that an increased nitrate concentration in the aqueous phase and an addition of other fission products both inhibited the technetium extraction even though fission product loading most likely changed the extraction mechanism to co-extraction by BTBP.     

Burn-up measurements on nuclear reactor fuels using high performance liquid chromatography

Burn-up measurements on thermal as well as fast reactor fuels were carried out using high performance liquid chromatography (HPLC). A column chromatographic technique using di-(2-ethylhexyl) phosphoric acid (HDEHP) coated column was employed for the isolation of lanthanides from uranium, plutonium and other fission products. Ion-pair HPLC was used for the separation of individual lanthanides. The atom percent fissions were calculated from the concentrations of the lanthanide (neodymium in the case of thermal reactor and lanthanum for the fast reactor fuels) and from uranium and plutonium contents of the dissolver solutions. The HPLC method was also used for determining the fractional fissions from uranium and plutonium for the thermal reactor fuel.     

Entwicklung von Methoden zur Abbrandbestimmung von Kernbrennstoffen in Schnellen Reaktoren

The burn up determination of fast reactor fuel is more difficult than for thermal reactors, because a greater number of isotopes fissions. The burn up measurement by fission products needed the determination of fission yields in a fast flux of isotopes contributing considerably to fission. To ease the analysis the sample conditioning was eliminated if applicable. The measurement techniques and the evaluation was automatised. A isotope correlation to check the analysis is mentioned.      

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Determination of the fast fission yield of141Pr using neutron activation analysis

A method is described for the determination of the fission yield of¹⁴¹Pr. This method was developed to determine the fast fission yield of¹⁴¹Pr in the Mark III loading (enriched uranium with about 2% zirconium) of the fast fission breeder reactor, EBR-1. The burnup of the fuel sample was determined using the previously reported fission yield of¹³⁷Cs. Praseodymium was separated from uranium, plutonium and other fission products by a combination of precipitation and ion exchange stages. Thereafter,⁵⁵Mn was added to serve as an internal flux monitor and praseodymium determined by neutron activation analysis. A precision of ±2% was obtained. Presented at the 15th Annual Meeting of the American Chemical Society, Miami Beach, Florida (USA), April 1967.     

Fission in the reaction of12C with209Bi at 73.4 and 84.2 MeV

Fission product yield studies in the reaction of 73,4 and 84.2 MeV¹²C with²⁰⁹Bi have been carried out using gamma-ray spectrometry. The cross sections for the production of fission products have been determined. The yield distribution of fission products is found to be symmetric and broad with FWHM around 25 mass units and peak near mass 107.7 and 107.0. The average number of neutrons emitted per fission have been found to be around 5.5 at 73.4 MeV and 7.1 at 84.2 MeV, respectively.