Short-lived Fission Products Research Articles

Rapid determination of strontium radionuclides in plants, fodder and foodstuffs

AbstractThe fission yield, the transfer factors in the food chain and the dose coefficient are large for the nuclear fission product Sr-90. The surveillance of Sr-90 in the food chain is therefore important in precautionary radiation protection and in assessing the radiation dose to the public especially after a nuclear incident. Prior to analysis, as it is a pure β-emitter, Sr must be separated from the sample by procedures which, for complex organic samples, are lengthy, laborious and dependent on operator skill. Ubiquitous natural radionuclides and short-lived fission products in samples contaminated with fresh fallout may interfere. Here we describe a fast, reproducible and effficient method for extracting Sr from grass, clover, maize, wholemeal rye, baby food, and total diet. The method depends on obtaining an ash free of traces of organic interferences. Sr may be separated from a dilute nitric acid leachate of such ash with a solution of dicyclohexyl-18-crown-6 in chloroform. Interfering radionuclides are removed with a special manganese (IV) oxide (active, precipitated from Merck). Sr is precipitated as carbonate then dispersed in a cocktail for liquid scintillation spectrometry. This allows simultaneous counting of Sr-89 (a short-lived β-emitter in fresh fallout) and Sr-90. The chemical yields of Sr determined with the gamma-emitting Sr-85 tracer are reproducible and greater than 75 % in all cases. The sample ashing requires 18 h and the extraction 4.0 to 4.5 h. Thus, a duplicate analysis may be completed within 2 days of receipt of the sample.

Feasibility of Eliminating the Use of Highly Enriched Uranium in the Production of Medical Radioisotopes

Significant quantities of highly enriched uranium (HEU)—more than enough to make a Hiroshima bomb—are used annually as neutron target material in Canadian, European, and South African reactors to produce the short-lived fission products used in nuclear medicine. The most important of these fission products is 99Mo, which decays into 99mTc, which is the most widely used medical radioisotope. The U.S. supplies weapon-grade uranium to the Canadian radioisotope producer and might in the future provide it to the European producers as well. As a condition for receiving U.S. HEU, the 1992 Schumer Amendment to the U.S. Atomic Energy Act requires that a foreign producer cooperate with the United States in converting to low-enriched uranium (LEU) targets. Some smaller producers have already done so. The Canadian producer has asserted, however, that the cost of conversion would be too high. The 2005 Burr amendment therefore exempted radioisotope producers in Canada and Europe from the Schumer amendment's requirements but requested a National Academy of Sciences study of the feasibility of conversion, setting as a feasibility test that the production cost be increased by no more than 10 percent. We show that paying for the conversion for the largest European production facility would increase the cost of 99Mo production there by only a few percent. For the Canadian facility the production cost could be more than 10 percent but the increase in the cost of the final 99mTc-containing radiopharmaceutical would be only about 1 percent. It is also pointed out that savings in security could well dwarf the costs of converting to LEU if HEU were no longer present at the production and radioactive waste sites.

Dating the age of a recent nuclear event by gamma-spectrometry

The age of a recent, i.e., fresh, nuclear event can be determined by measuring the activity of short-lived parent and daughter fission products. The event studied was a short irradiation, of a small sample of uranium, in a nuclear reactor. Two clocks were investigated, 92Sr-92Y and 135I-135Xe. Measurements of the source by gamma-spectrometry yielded very good agreement between true and measured ages. The upper and lower age limits of applicability for the clocks in question were defined. The half-life of 92Sr was found 2.635±0.008 hours and of 135I 6.65±0.04 hours.

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

The "nuclear car wash": a scanner to detect illicit special nuclear material in cargo containers

There is an urgent need to improve the reliability of screening cargo containers for illicit nuclear material that may be hidden there for terrorist purposes. A screening system is described for the detection of fissionable material hidden in maritime cargo containers. The system makes use of a low-intensity neutron beam for producing fission and the detection of the abundant high-energy /spl gamma/ rays emitted in the /spl beta/-decay of short-lived fission products and /spl beta/-delayed neutrons. The abundance of the delayed /spl gamma/ rays is almost an order of magnitude larger than that of the delayed neutrons normally used to detect fission, and they are emitted on about the same time scale as the delayed neutrons, i.e., /spl sim/1 min. The energy and temporal distributions of the delayed /spl gamma/ rays provide a unique signature of fission. Because of their high energy, these delayed /spl gamma/ rays penetrate low-Z cargoes much more readily than the delayed neutrons. Coupled with their higher abundance, the signal from the delayed /spl gamma/ rays escaping from the container is predicted to be as much as six decades more intense than the delayed neutron signal, depending upon the type and thickness of the intervening cargo. The /spl gamma/ rays are detected in a large array of scintillators located along the sides of the container as it is moved through them. Measurements have confirmed the signal strength in somewhat idealized experiments and have also identified one interference when 14.5-MeV neutrons from the D, T reaction are used for the interrogation. The interference can be removed easily by the appropriate choice of the neutron source.

Analysis of fissile materials by high-energy neutron-induced fission decay gamma-rays

Author(s): Firestone, R.B.; English, G.A.; Reijonen, J.; Gicquel, F.; Leung, K.-N.; Perry, D.L.; Garabedian, G.; Bandong, B.; Revay, Zs.; Molnar, G.L. | Abstract: Thermal neutrons from the Budapest Research Reactor and fast neutrons from the Berkeley Neutron Generator Facility have been used to analyze uranium. It has been shown that both prompt and delayed gamma rays from neutron capture and fission product decay can be used to analyze uranium concentrations and 235U enrichment. Detection of neutrons from the spontaneous fission of 238U has also been demonstrated for uranium analysis. The observation of high-energy gamma rays following the decay of short-lived fission products is a sensitive indication of fissile material, and the ratio of fission product gamma ray intensities can uniquely determine the concentrations of fission isotopes.

Use of γ Radiation from Short-Lived Fission Products for Detecting 235U and 239Pu

It is suggested that γ radiation with E γ > 4900 keV from short-lived fission products produced by thermal neutrons be used to detect 235U and 239Pu in samples. A time regime is substantiated: 120 sec irradiation, 60 sec holding time, and 120 sec measurement time. The contribution of the reaction (n, p) on fast neutrons is studied.

Monte Carlo models for the production of β-delayed gamma-rays following fission of special nuclear materials

A Monte Carlo method for the estimation of β-delayed γ-ray spectra following fission is described that can accomodate an arbitrary time-dependent fission rate and photon collection history. The method invokes direct sampling of the independent fission yield distributions of the fissioning system, the branching ratios for decay of individual fission products and the spectral distributions for photon emission for each decay mode. Though computationally intensive, the method can provide a detailed estimate of the spectrum that would be recorded by an arbitrary spectrometer, and can prove useful in assessing the quality of evaluated data libraries, for identifying gaps in these libraries, etc. The method is illustrated by a first comparison of calculated and experimental spectra from decay of short-lived fission products following the reactions 235U(n th, f) and 239Pu(n th, f). For general purpose transport calculations, where detailed consideration of the large number of individual γ-ray transitions in a spectrum may be unnecessary, it is shown that an accurate and simple parameterization of a γ-ray source function can be obtained. These parametrizations should provide high-quality average spectral distributions that should prove useful in calculations describing photons escaping from thick attenuating media.

Miss Piggy, a californium-252 fission fragment source as a generator of short-lived radionuclides

Carrier-free short-lived nuclides are employed in many different fields of modern nuclear chemistry. The two main production strategies are either thermal neutron-induced fission of 235U or 239Pu at nuclear reactors or spallation neutron sources or charged particle-induced nuclear reactions at accelerator facilities. An alternative method is to use a spontaneously fissioning nuclide. A facility applying this technique (“Miss Piggy”) was built at the University of Berne (Switzerland). Californium-252 ( 252Cf), which has a 3% fission branch and a half-life of 2.645 a, is used for the production of short-lived fission products that are stopped in an adjacent recoil chamber. Short-lived nuclides are transported out of the recoil chamber using the well-known gas-jet technique. Over 100 nuclides have been identified so far and used in different applications. Since such a device does not require any large facility and is easy to operate it serves well the needs of typical university laboratories.

A reconsideration of the external dose assessment for the Techa river population.

A new methodology is employed to infer the time course of the external dose rate that affected the Techa riverside population. This was accomplished using information on the radionuclide composition of released radioactive waste and from earlier and current data on the degree of the river floodplain contamination with (137)Cs. The new approach is proposed for use in the dose reconstruction. Compared to current assumptions it indicates a considerably higher dose contribution due to short-lived fission products from the predominant peak of contamination in 1951. Relative to the present Techa river dosimetry system (TRDS-2000) this information may increase the external dose estimates several fold and correspondingly reduce the solid cancer risk estimate.

The Southern Urals radiation studies. A reappraisal of the current status.

In the late 1940s and early 1950s the nuclear workers of the Mayak Production Association in the Southern Urals were exposed to high doses from gamma-rays and from incorporated plutonium. In addition, the population of the Techa riverside downstream of the plutonium-production sites received continued exposures from external gamma-rays due to fission products released into the river and from the internal radiation due to incorporation of the fission products. Based on two international coordination meetings in 1998 and 2000, a synopsis has been given recently in this journal of the radioepidemiological studies on these exposed populations. This commentary describes the current status of these singular investigations with regard to the dosimetry, the assessment of late health effects, and the risk estimation both for the Mayak nuclear workers and the Techa riverside population. A central issue are newly published reduced estimates of the external dose to the Techa riverside population which imply substantially increased risk coefficients for solid cancer. Unless the new dosimetry system, TRDS-2000, has missed a major dose contribution, there is now conspicuous disagreement with current risk estimates. Unaccounted doses from atmospheric releases of fission products and from radiological screening of the Techa riverside population need to be explored, but underestimation of the short lived fission products released into the river appears to be a more critical factor. It is furthermore argued that even if TRDS-2000 were confirmed it would remain questionable whether risk estimates can be based on organ-specific doses when they are obtained in a population with a much higher bone-marrow exposure that may possibly have caused an 'abscopal' radiation effect.

Development of the Diamex Process for Treating PHWR High-Level Liquid Waste

The extraction behavior of actinides like U(VI), Pu(IV), and Am(III) as well as of fission products like Tc(VII), Zr(IV), Eu(III), and structural material Fe(III) using nitric acid/simulated pressurized heavy water reactor-high-level waste solution as aqueous phase and N,N’-dimethyl-N,N’ dibutyl tetra decyl malonamide (DMDBTDMA) in n-dodecane as solvent was investigated. The present work contributes significantly toward the development of nonphosphorous (environmentally friendly) extractant capable of partitioning long-lived actinides and fission products like 99Tc from relatively short lived fission products like 90Sr and 137Cs as well as inactive components present in high-level waste. The DAm values determined in the temperature range between 15 and 45°C showed a gradual decrease with increase in temperature throughout the acidity range. The effect of N,N di-2-ethylhexyl acetamide (D2EHAA) as phase modifier on the physicochemical properties of DMDBTDMA/n-dodecane solvent was investigated and could be correlated with parameters like the limiting organic concentration value of HNO3, DAm, or phase disengagement time. Overall comparison of diamide extraction (Diamex) and transuranium extraction (Truex) solvents has been made.

JEFF-3T: Decay Data and Fission Yield Libraries

Comprehensive decay-data and fission-yield libraries provide important input to a wide range of nuclear physics codes for nuclear applications. A new initiative has begun under the auspices of the NEA/OECD to generate improved data sets that will constitute the JEFF-3 libraries in ENDF-6 format, primarily for nuclear power, fuel reprocessing and waste management needs. Various sources of decay data have been accessed in order to assemble these files: NUBASE, ENSDF, UKPADD-6 and UKHEDD-2. Efforts have also focused on the evaluation of decay data for a number of important short-lived fission products, so that artificial adjustments to some of the relevant decay data and fission yields are not required to accommodate a previous lack of such data. Fission yields were adopted from UK evaluations recently undertaken to create the UKFY3 library. Decay-data files for 3 755 nuclides have been prepared, including sets of data for the stable nuclides (i.e. mass, natural abundance, spin and parity). Problems in the assignment of ENDF material numbers were addressed, while format and consistency tests were made using CHECKR and FIZCON, respectively. The assembly processes are discussed and reviewed, and the contents of the JEFF-3T starter libraries are described.

Determination of Gamma-Ray Exposure Rate from Short-Lived Fission Products under Criticality Accident Conditions

For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 s or less of time lapse after an initial power burst.

Short-lived fission products as a diagnostics tool for studying atom and ion behavior in a gas-based laser ion source

A striking difference between gamma spectra of neutron-rich Rh isotopes obtained with the laser ion source at the Leuven Isotope Separator On-Line facility has been observed depending on the mode of operation. Although the global production rate of 112Rh g,m nuclei decreases considerably, the ratio between the productions of 112Rh g and 112Rh m increases strongly when no laser ionisation is used. This effect is caused by 112Ru atoms which decay during gas evacuation of the ion source gas cell, thereby producing 112Rh g ions. The comparison in time behaviour of reaction produced ions, β-decay produced ions and laser produced ions makes it possible to study and characterise the different processes in the gas cell. The influence of these processes has to be considered when extracting nuclear information such as the relative feeding of different short-lived isomers and isotopes and fission cross-sections in a particular mass chain.

EXTRACTION STUDY FOR TRUEX PROCESS USING SHORT-LIVED RADIONUCLIDES PRODUCED BY NEUTRON IRRADIATION OF URANIUM

The extraction behavior of short-lived fission products and neptunium was studied by using octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide under the conditions of the transuranic elements extraction (TRUEX) process. The short-lived fission products and neptunium were produced by neutron irradiation of UO2 of natural uranium, and the extraction behavior of 93Y, 99Mo, 97Zr, 122Sb, 132Te, 133I, 143Ce, and 239Np was simultaneously studied, where 122Sb was produced by neutron irradiation of antimony metal. The extraction of fission products and Np under the conditions of the PUREX process was also studied for comparison. The extraction of nuclides in the presence of large amounts of uranium(VI), and the presence of oxalic acid was examined. The conditions and performance of the TRUEX extraction were discussed by considering the obtained results.

Determination of Gamma-Ray Exposure Rate from Short-Lived Fission Products under Criticality Accident Conditions.

For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 s or less of time lapse after an initial power burst.

Possibility of activation γ-spectrometric measurements for determining the masses of nuclear materials in samples

Measurements of γ-ray spectra performed on uranium and plutonium samples irradiated with 252Cf source neutrons are described. Peaks due to 138Cs and other short-lived fission products are observed in the spectra. The intensity of the peaks characterizes the content of fissioning isotopes in the samples.

The fractal nature of the surface of uranium dioxide: a resolution of the short-lived/stable gas release dichotomy

The framework for analysis of fission product release in the Halden gas flow rigs has been developed over a period of 20 years. The predominant mode of release is single gas atom diffusion to free surfaces with a small, but burn-up-dependent contribution from direct recoil. Measurements of longer-lived fission products indicated that their release appeared to be controlled by a diffusion coefficient, which, at low temperatures, was several orders of magnitude smaller than that of the shorter-lived fission products. The enhanced release of the shorter-lived products was attributed to a near-surface enhancement effect. However, a theoretical study of a two-zone diffusion model reveals that it is incapable of accounting for the observations. A model based on the fractal nature of the uranium dioxide surface circumvents these difficulties and provides a framework in which a single three-term diffusion coefficient can be used to describe the release of both short- and long-lived fission products.

A code to simulate nuclear reactor inventories and associated gamma-ray spectra

A computer code has been developed to simulate the gamma-ray spectra that would be measured by airborne gamma spectrometry (AGS) systems from sources containing short-lived fission products. The code uses simple numerical methods to simulate the production and decay of fission products and generates spectra for sodium iodide (NaI) detectors using Monte Carlo codes. A new Monte Carlo code using a virtual array of detectors to reduce simulation times for airborne geometries is described. Spectra generated for a short irradiation and laboratory geometry have been compared with an experimental data set. The agreement is good. Spectra have also been generated for airborne geometries and longer irradiation periods. The application of this code to generate AGS spectra for accident scenarios and their uses in the development and evaluation of spectral analysis methods for such situations are discussed.