This paper reports on the method, results, and experiences of the Sr-89 and Sr-90 analysis performed on low-and intermediate-level radioactive liquid waste samples from the Paks Nuclear Power Plant. The analysis is based on a combined separation method, which results in the separation and detection of radiostrontium (Sr-89 and Sr-90) and of Y-90. After the suitable pretreatment of the samples, radiostrontium is concentrated using the Sr-Spec(R) (EIChroM Industries, Darien, IL, 60561 USA) chromatographic column. The eluted radiostrontium is precipitated in a thin layer and its activity measured by a low-background, gamma-ray-compensated, alpha-beta detector. The chemical yield is determined by atomic emission spectroscopy. The Y-90 activity of the samples is also measured by the procedure described elsewhere. This measurement is based on the selective adsorbing property of the yttrium-ammonium-oxalate and the yttrium-oxalate double salt mixture of yttrium. Under the proper conditions ( pH, oxalate anion activity, mass ratio of the salts) the quantitative yield of Y-90 can be achieved, as determined by using the Y-88 inner standard. As a result, the radioactive concentration of Sr-90 in the samples could be calculated. It was concluded that Sr-89 was not present in the samples because of the relatively long tin years) storage life of these samples. However, this combined method can be easily adapted to matrices sampled at nuclear test sites.

The majority of the radioactive waste that is generated in Hungary arises from the operation of the Paks Nuclear Power Plant. Fuel for Hungarian NPP just as for all other East European V VERS has in the past been supplied by the ex-Soviet Union. As part of Hungary’s agreement on fuel supply the Soviet Union was obliged to take back and dispose of all spent fuel. Due to the likely interruption of the current spent fuel disposal route in 1993, decision has been made to construct at Paks a Modular Vault Dry Store. This interim storage facility can hold spent fuel after it has completed 5-year cooling in the ponds. As insurance against the waste remaining in Hungary or being returned after reprocessing it is highly desirable to proceed by planning for possible disposal of spent fuel some 50 years or more in the future. Based on preliminary assessments and technical judgements, the use of the Permian Boda Claystone Formation in the Mecsek Hill area is being considered for high level waste disposal. To evaluate the suitability of this formation as a location for a nuclear fuel waste repository investigations have started. In 1992, after two previous siting attempts had ended in failure a National Project has been launched aiming at establishment of a repository for low and intermediate (LLW/ILW) level radioactive wastes. Between 1993 and 1995, some 300 geological objects were identified as potentially suitable for either near-surface or tunnel-type/mined cavity/disposal. In the second stage of the site selection on a designated area suitable geological objects were screened to identify and rank potential formations. By 1996 three areas have been selected where preliminary site investigations have started. Two of the selected areas are subject to near-surface disposal, while the third region candidate for subsurface disposal is in granitic terrain. The paper overviews the problem areas and development priorities in the field of the nuclear waste management.

The topic of this paper is the method, results and experiences of Sr-90 analysis performed on aqueous radwaste samples from the NPP Paks. The analysis is based on a combined separation method, which results in the separation and detection of radiostrontium (Sr-89 and Sr-90) and or Y-90. After suitable pretreatment of the samples, radiostrontium is concentrated using a Sr.Spec(R) chromatographic column. The eluted radiostrontium is precipitated in thin layer and its activity measured in low-background, gamma-ray compensated alpha-beta detector. The chemical yield is determined by AES. The Y-90 activity of the samples is also measured using the procedure described elsewhere. This is based on the selective adsorption properity of the yttrium-ammonium-oxalate and yttrium-oxalate double salt mixture towards yttrium. In proper conditions (pH, oxalate anion activity, mass ratio of the salts) quantitative yield of Y-90 can be achived which is determined using Y-88 internal standard. As a result, the radioactive concentration of Sr-90 in the samples could be calculated. It was concluded that Sr-89 was not present in the samples, which is due to the relatively long (years) storage life of these samples.

A method was developed for the detection and quantitative determination of the alpha-emitting radionuclides Pu-238, Pu-239, Pu-240, Am-41, Cm-244, U-234, H-238 occuring in low- and medium-activity radioactive liquid wastes. The procedure developed for chemical sample preparation and subsequent analysis by alpha-spectrometry is described in detail, and some results yielded by the method are given. In the case of such sample types for the separation of borate ions an effective method is, to farm volatile alkyl borates and to use evaporation residues in the next step of the analysis. This method seems to be advisable because it is simple and fast. The transuranium activities measured by the new method in radioactive liquid wastes. of the VVER-440 type (PWR) reactor at the NPP Paks, are: A((239+240)Pu) = 72+/-6 mBq/dm(3) and A (Pu-238+Am-241) = 90+/-9 mBq/dm(3). The estimated uncertainty is +/-20 %.

In the execution of disposal of low and intermediate level radioactive wastes, it is important to evaluate accurately the kind and quantity of each radionuclide in the wastes. For such an evaluation, correlation of non-gamma-emitting nuclides based on gamma-emitting nuclides is recommended and regarded as a practical method. This method necessitates a completion of a highly accurate and reliable nondestructive assay system of gamma-emitting nuclides for practical use. In 1992, in support of the new waste disposal program in Hungary, Paks NPP initiated a waste characterization program to determine the radiological properties of its radwastes. A segmented gamma scanning system has been set up to measure the gamma-emitting nuclides in 200 litre low level drums following in-drum compaction. In the framework of the program a radiochemical analysis sub-program was started to determine the long-lived non-gamma emitting radionuclides, mainly those listed in the US regulatory document (10CFR61). The radionuclides of interest have been3H,14C,90Sr,55Fe,59Ni,99Tc,129I and TRUs. Sample preparation techniques and measurement methods have been selected and used. Newly developed or adopted methods have been tested on real liquid radwaste streams such as concentrates, ion-exchange resin and sludge. The measurements taken so far have revealed brand new information and data on radiological composition of waste of WWER-type reactors. In the next stage of the characterisation program attempt will be made for providing correlation factors between the gamma and non-gamma-emitting radionuclides in different waste streams. Short description of the methods and results on waste inventory are given by highlighting the problem areas.

During the operation of nuclear reactors, various alpha-emitting isotopes and activation products, coming from the surface contamination of fuel elements or from the fuel itself due to leaking through possible cracks or damage of the enclosure, may appear in nuclear power plant wastes.It is known that in the case of uranium-based fuel elements, isotopes of uranium, neptunium and plutonium constitute the greater part, while those of americium and curium the smaller part of the alpha-emitting isotopes occurring in the wastes, These also appear in the regeneration wastes of primary coolant purifiers in concentrations depending on their amounts originally present and their solubilities, the latter being determined by the chemical nature of the radionuclides.Several papers have been published on the analysis of radioactive nuclear power plant wastes for transuranium elements, In these articles different individual techniques are described for measuring alpha-emitting radionuclides used at various nuclear reactors.(1-3) It is quite understandable that different analytical procedures are needed in the case of different types of reactors for the selective detection of or-emitters.In this paper a new technique is described in detail, which was developed for the analysis of alpha-emitting radionuclides in low-activity liquid wastes of VVER-440 type (PWR) reactors, The main feature of the procedure is its suitability for treating solutions saturated with respect to boric acid and also containing several organic and inorganic compounds.

Safety studies related to the disposal of low- and intermediate waste indicate that the long term risk is determined by the presence of long-lived nuclides such as C-14, Ni-59, Ni-63, Tc-99, I-129 and the transuranium elements.As most of these nuclides are difficult to measure, the correlation between these critical nuclides and some other easily measurable key nuclides such as Co-60 and Cs-137 has been investigated for typical waste streams of Paks Nuclear Power Plant (Hungary) and scaling factors have been proposed.An automated gamma-scanning monitor has been purchased and calibrated to determine the gamma-emitting radionuclides.Radiochemical methods have been developed to determine significant difficult-to-measure radionuclides. The radionuclides of interest have been H-3, C-14, Sr-90, Fe-55, Ni-59, Tc-99, I-129 and TRUs. The measurements taken so far have revealed brand new information and data on radiological composition of waste of WWER-type reactors.The reliability of the radioanalyitical methods was checked by an international intercomparison test. For ail radionuclides the Hungarian results were in the average range of the total data set.

We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1 ey−1 for hydrocarbons and 0.06 TBqGW−1 ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.

The Paks Nuclear Power Plant in Hungary runs with four pressurized water reactors, each of 440-MWe capacity. Sampling systems have been developed and used to determine the 14C of various chemical forms (14CO2, 14CO, 14CnHm) in the airborne releases. The average normalized yearly discharge rates for the time period 1988-1991 are equal to 0.77 TBq GWe-1 y-1 for hydrocarbons and 0.05 TBq GWe-1 y-1 for CO2. The contribution of 14CO was less than 0.5% of the total emission. The 14C discharge rate is estimated to be four times higher than the corresponding mean data of Western European pressurized water reactors. The calculated effective dose equivalent to individuals living in the vicinity of the power plant, due to 14C release, was 0.64 microSv in 1991 while the effective dose equivalent due to the natural 14C level was 15 microSv y-1. The long-term global impact of the 14C release in the operational period of the plant (1982-1991) was 1,270 man-Sv. The 14C excess in the environmental air has been measured since 1989 by taking biweekly samples at a distance of 1.7 km from the nuclear power plant. The long-term average of radiocarbon excess coming from the power plant was 2 mBq m-3. The local 14C deposition was followed by tree ring analysis, too. No 14C increase higher than the uncertainty of the measurement (four per thousand = 0.17 mBq m-3) was observed.

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

Several codes and measuring technics have been elaborated for shut-down dose reduction research at Paks nuclear power plant /NPP/. Main principles and some results of these codes as well as comparison to measurements at Paks and at other NPP are given.