Management of personnel exposure data with a personal computer

This paper describes a software package that manages dosimeter inventory and exposure data with a personal computer. The computer, connected with the access control system, deals with both records of the film badge (FB) supply to workers and the ordering of new FB for a company, processes exposure data and produces reports. The software not only ensures accurate reports but also significantly reduces the tedious labour involved in managing the control system.     

Determination of technetium-99 in soils and radioactive wastes using ICP-MS

Three methods have been used for the determination of ⁹⁹Tc in soils and solid radioactive wastes using ^99mTc as a yield monitor. In the method one and three the samples were leached in low concentrated nitric and sulphuric acid. Many contaminants were then co-precipitated with Fe(OH)₃ in alkali media and Tc in the supernatant was separated using anion-exchange extraction chromatography. There were made also some studies how to improve the chemical recovery of ^99mTc in the process of chromatography. In the method two the sample was ashed and then leached in 8 mol dm⁻³ HNO₃ and after iron precipitation, technetium was separated on chromatographic column. The chemical recovery of ^99mTc was optimized in the process of chromatography and leaching. Typical recoveries of technetium determined with ^99mTc tracer for all these methods were in the range 39 %–87 %. The ⁹⁹Tc activity was measured using proportional low-background beta detector after one week of staying to allow decay of ^99mTc activity. ⁹⁹Tc was also determined by the non-radiometric method using inductively coupled plasma mass spectrometer.     

Melt densification of radioactive plastic wastes

Summary Volume reduction studies were carried out on low level radioactive plastic wastes containing polyethylene, PVC and neoprene by melt densification. The optimized temperature for melting of plastics was between 170 and 180 °C. Based on laboratory scale studies. Plant scale studies were planned and conducted. The volume reduction factors obtained were around 30, which was 6-10 times higher than that of the conventional compaction process. Thermogravimetry and infrared spectroscopy were used to investigate the thermal and structural properties of the given materials. The effect of the presence of salts like potassium permanganate and hydrazine sulphate on the thermal properties of the materials was also evaluated. Leaching studies were also performed on melt-densified specimens in the laboratory. The average leach index was observed to be around 9, which was higher than the minimum stipulated value.     

Control system of stockers in radioisotopes storage room using a personal computer

This paper describes a control system of stockers for radioactive source in storage room at laboratory for tracer experiment. The system is composed of a personal computer, a locker controller, three card readers, a monitor TV, and a video tape recorder (VTR). The personal computer controls other equipment with a registered user's number. When a user inserts an identification card into the card reader, the computer memorizes assigned gate number, the user's number and the time; it processes those data and prints out a document. The locker controller releases the electric key of user's locker which is designated by the computer. The VTR records the person entering into the storage room to identity if he uses his card. This system proved to effectively prevent intrusion into the storage room of an unregistered person or to use carelessly other user's source; in addition it can record precisely the stock of radioactive source.     

Determination of alpha-emitting isotopes in radioactive wastes

An analytical procedure was developed for the alpha-spectrometric determination of uranium (²³⁸U, ²³⁵U, ²³⁴U) and transuranium isotopes (^239+240Pu, ²⁴⁴Cm) in liquid radioactive wastes (sludges, evaporation residues) of low and medium activity from the VVER-440 reactor of the nuclear power plant in Paks, and in waste waters to be released into the environment. Radioactive sludge samples were separated to a liquid phase and a wet suspension (solid) phase by centrifuging, and the two phases were treated and analyzed separately, in different ways. A sample preparation procedure based on chemical separation was worked out for the isolation of the alpha-emitting components of radioactive evaporation residues which were saturated with boric acid. To the separate determination of the low activity alpha-emitting isotopes appearing in waste waters to be released into the environment, a microvawe concentration technique was applied. The accuracy of the results obtained using the above chemical separation and alpha-spectrometry was tested in the framework of the international U. S. DOE EML Quality Assessment Program.     

Radiological characterization of low-and intermediate-level radioactive wastes

An analytical procedure for the determination of activation products ²³⁸Pu, ²⁴¹Pu, ²³⁹Pu/²⁴⁰Pu, ²⁴¹Am, ²³⁷Np, and a fission product ⁹⁰Sr in radioactive wastes is presented. Samples were decomposed using Fenton’s reaction. The separation was performed by anion-exchange chromatography, extraction chromatography, using TRU and Srresin, and precipitation techniques, followed by α-spectrometry and LSC counting. Tracer solutions and pure ion exchange resins were used to prepare artificial samples and trace nuclides during the analytical procedure. Some real samples of spent ion-exchange resins originating from our TRIGA Mark II research reactor were analyzed.     

Pre-concentration and determination of 90Sr in radioactive wastes using solid phase extraction techniques

This paper describes the use of IBC??s AnaLig^®Sr-01 molecular recognition technology product to effectively and selectively pre-concentrate, separate and recover strontium from radioactive waste samples. The use and effectiveness of AnaLig^®Sr-01 gel was successfully validated by analysis of International Atomic Energy Agency (IAEA 375) reference soil and National Physical Laboratory (NPL)?CHigh Alpha?CBeta (2003) liquid sample. The second part of this paper focuses on analysis of radioactive waste samples from nuclear power plant A1 Jaslovske Bohunice in Slovak Republic (NPP A1).     

Statistical estimate for evaluation of vitrified radioactive wastes

The evaluation of experimental results by methods of mathematical statistics enabled us to derive a number of conclusions on the leachability of vitrified radioactive wastes. Practical application of this procedure requires that the ratio of Na and K concentrations in the solution should be independent of the leaching time. The actual value of this ratio is influenced, above all, by the properties of the glass matrix. These results confirm the assumption that the Na/K correlation found could be extended for the determination of the Na/¹³⁷ Cs concentration ratio. This finding was used for the application of a In-In correlation, while evaluating the quality of vitrified radioactive wastes products.     

Analysis of organics in highly radioactive nuclear wastes

Analytical methods are being developed at Pacific Northwest Laboratory (PNL) for the organic analysis of nuclear wastes. In this study our laboratory analyzed the organic content of two highly radioactive wastes, a neutralized cladding removal waste (NCRW) and a volume reduction, double-shell slurry (DSS) waste. In-house methods were developed and U.S. Environmental Protection Agency (EPA) methods were modified for isolating and analyzing volatile, semivolatile and hydrophilic organics in the hot cell and radiation hood. The procedures worked well, particularly for the NCRW, in which 94.9% of the waste's total organic carbon (TOC) has been characterized. In contrast, only 1.2% of the DSS waste's TOC has been identified to date. Quite a variety of volatile, semivolatile and hydrophilic organics, e.g., chelating agents, were identified in the wastes, occasionally at relatively high mM, or mg/g, concentrations. Most of the compounds are either source-term organics associated with nuclear operations or their degradation products, indicating that organic diagenesis in nuclear wastes can be quite vigorous and extensive.     

A data acquisition system for mass-analyzed ion kinetic energy spectra using a personal computer

An inexpensive data acquisition system has been developed for mass-analyzed kinetic energy experiments which involve scanning the electrostatic analyzer of a reversed geometry mass spectrometer. The various hardware and software design features are described. Results for data obtained with a commercially available VG ZAB-2F mass spectrometer are presented and discussed.     

Digital radiation monitoring system with a personal computer

This paper describes a data processing system of monitors with a personal computer for a radioisotope laboratory. A monitoring system includes 6 area, 6 dust, and 3 gas monitors. Monitor signals through a multiplexer were fed to the personal computer which has analogue ports. Processed data were displayed on a color CRT, stored on a 5 inch mini floppy disk, and printed out as a report. We developed a data processing program with BASIC language. This system largely reduced manpower required to estimate radiation levels in the laboratory. The personal computer has sufficiently enough ability to construct a digital monitoring system for a small radioisotope laboratory.     

Treatment of low level radioactive liquid wastes using composite ion-exchange resins based on polyurethane foam

Composite ion-exchange resins were prepared by coating copper-ferrocyanide (CFC) and hydrous manganese oxide (HMO) powders on polyurethane (PU) foam. Polyvinyl acetate/Acetone was used as a binder. The foam was loaded with about five times its weight with CFC and HMO powders. The distribution coefficients of CFC-PU foam and HMO-PU foam for cesium and strontium respectively were estimated. Under similar conditions the HMO-PU foam showed higher capacity as well as better kinetics for removal of strontium than CFC-PU foam for Cs. The pilot plant scale studies were conducted using a mixed composite ion-exchange resin bed. About 1000 bed volumes could be passed before attaining a DF of 10 from an initial value of 60–80. The spent resin was digested in alkaline KMnO₄ and the digested liquid was fixed in cement matrix. The matrices were characterized with respect to compressive strength and leach resistance.     

Decontamination of radioactive liquid wastes by hydrophytic vegetal organisms

This paper deals with the bioaccumulation of some radioactive ions from contaminated waste solutions, on hydrophytic vegetal organisms. In order to follow the distribution of radioactive ions ¹³⁷Cs⁺, ⁶⁰Co²⁺ and ⁵¹Cr³⁺ in various cell components extracted from Spirulina platensis, Porphiridium cruentum, Scenedesmus quadricauda, Lemna minor,Elodea canadensis, Pistia stratiotes and Riccia fluitans, the plants were cultivated in radioactive solutions. The resulting complexes were extracted with acetone or acetic acid and separated chromatographically. The results show an intense activity of the polysaccharide and lipoid fractions in the bioaccumulation process. The bioaccumulation varies in the series: Spirulina>Scenedesmus>Porphiridium>Riccia>Pistia> Lemna ³ Elodea for ¹³⁷Cs⁺ and ⁶⁰Co²⁺; Spirulina>Porphiridium>Scenedesmus>Riccia>Pistia>Lemna>Elodea for ⁵¹Cr³⁺.     

Diffusion of actinides in glasses containing simulated radioactive wastes

Diffusion coefficients of radionuclides²³⁷Np,²³⁹Pu and²⁴¹Am in simulated alumina phosphate and alumina borosilicate glasses at temperatures lower than their transformation temperature were determined. Actinides are known to be the least mobile elements. In particular, the diffusion coefficients of actinides in alumina phosphate glasses at 673 K are about 10^–18 m²·s^–1, and in alumina borosilicate at 773 K about 10^–19 m²·s^–1. It is shown that crystallization of glasses leads to increasing²³⁷Np diffusion mobility. It is also shown that a rather small quantity of water absorbed by a crystallized alumina phosphate glass in tensifies low-temperature migration of²³⁷Np.     

Separation of boric acid from radioactive wastes by liquid-liquid extraction

The extraction of boric acid from a solution modeling radioactive waste has been studied. Different aliphatic alcohols were used as extractants. Factors affecting the distribution ratio of boric acid and the behaviour of some other components of the solution under conditions of boric acid extraction were investigated. The composition of species extracted was determined from distribution data of n-hexanol and H₃BO₃.     

Treatment of radioactive and industrial liquid wastes by Eichornia crassipes

Chemical factors such as pH, concentration and temperature affecting the removal of UO₂ ²⁺, Th⁴⁺, Fe³⁺, Cu²⁺, Pb²⁺,Cd²⁺, Ni²⁺, MnO₄ ^- and phenol by Eichornia crassipes aquatic plant from their solutions were examined. Maximum uptake of ions by Eichornia crassipes occurred at pH 4 to 6±0.5 at 25±3 °C. An initial rapid uptake phase for the first 6 hours followed by a slower near linear one was observed. One gram of Eichornia crassipes can accumulate about 25 mg UO₂ ²⁺, 5 mg Th⁴⁺, 30 mg Fe³⁺, 10 mg MnO₄ ^-, 15 mg Cu²⁺, 1.0 mg Pb²⁺, 1.5 mg Ni²⁺, 0.7 mg Cd²⁺ and or 25 mg of phenol.     

Activity reduction of radioactive mixed wastes for subsequent carboxylate determinations

Instrumental neutron activation analysis (INAA) based on a nuclear reactor and proton induced X-ray emission (PIXE) based on a 2 Me V Van de Graaff accelerator were used to analyse different environmental samples including coal, sawdust, fly ash and landfill materials. These samples represent a large component of the solid waste being buried in the ground, and may be a potential source of toxicity which can have not only adverse effects on the soil and ground water, if diffused during leaching processes, but may also be transferred to humans through the food chain. Both techniques were employed to determine trace element concentrations in these matrices. The concentration of 30 elements were detemined, namely Na, Mg. Cl, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Se, Br, Rb, Cd, Ba, La, Sm, Eu, Tb, Ho, Yb, Hf, Ta and Th. The detection limits in the various sample matrices were calculated and a comparison was made for those elements commonly detected by both techniques.     

A comparative study using liquid scintillation counting and X-ray spectrometry to determine 55Fe in radioactive wastes

The radionuclide ⁵⁵Fe was determined in samples of radioactive wastes from the water cleanup system of the IEA-R1 nuclear research reactor. In order to validate the results, the ⁵⁵Fe activity concentration was measured in eight waste samples and in six simulated samples containing the most important interfering radionuclides. A simple method was employed to separate and purify ⁵⁵Fe from other radionuclides present in these samples, combining co-precipitation with ammonium hydroxide and purification with anionic ion-exchange resin, which enables ⁵⁵Fe to be quantified either by liquid scintillation counting (LSC) or by X-ray spectrometry using a low-energy germanium spectrometer (LEGe). Both measurement methods were used so that the separation and purification process could be confirmed by comparison of spectra with and without the utilization of anionic ion-exchange resin. Activity and interferences were compared in the results obtained from LSC and LEGe measurement methods.     

Uptake of cesium and strontium by crystalline silicotitanates from radioactive wastes

Crystalline silicotitanate inorganic ion exchanger, with a sitinakite structure is candidate material for remediation of aqueous nuclear waste streams. The syntheses of crystalline silicotitanate (CST) and Nb-substituted crystalline silcotitanate (Nb-CST) were carried out under hydrothermal conditions and the products were characterized using techniques viz., XRD, SEM/EDS, DTA/TGA, surface area respectively. Batch experiments were carried out to study the kinetics of uptake of ¹³⁷Cs and ⁹⁰Sr, to estimate the decontamination factor (DF) values and distribution coefficients (K _d) for the above synthesized CST and Nb-CST samples from actual radioactive waste solutions. The DF values for uptake of Cs and Sr by Nb-CST after 24 h of equilibration was 355 and 136 whereas for CST it was found to be 40 and 176 respectively. The K _d values for uptake of Cs and Sr for Nb-CST after 24 h of equilibration was found to be 35,490 and 13,500 mL/g respectively whereas the K _d values for uptake of Cs and Sr for CST was found to be 4,025 and 17,525 mL/g respectively. The ion exchange capacity of Nb-CST towards ⁹⁰Sr and ¹³⁷Cs was estimated to be 11.8 and 3.2 meq/g respectively whereas the ion exchange capacity of CST towards ⁹⁰Sr and ¹³⁷Cs was estimated to be 14.6 and 4.4 meq/g respectively.     

Online retrieval of patient information by asynchronous communication between a general purpose computer and a stand-alone personal computer

Asynchronous communication was made between host (FACOM M-340) and personal computer (OLIBETTIE S-2250) to get patient's information required for RIA test registration. The retrieval system consists of a keyboard input of six numeric codes, patient's ID, and a real time reply containing six parameters for the patient. Their identified parameters are patient's name, sex, date of birth (include area), department, and out- or inpatient. Linking this program to RIA registration program for individual patient, then, operator can input name of RIA test requested. Our simple retrieval program made a useful data network between different types of host and stand-alone personal computers, and enabled us accurate and labor-saving registration for RIA test.