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Summary  

During the period of 1993-2001 chemical decontaminations of 24 SGs in the units 1-3 of the Paks NPP were carried out by a non-regenerative version of AP-CITROX technology, even in two or three consecutive cycles. A comprehensive investigation of the above decontamination method have revealed that the fundamental issues of analytical chemistry and corrosion science were not taken into consideration during the elaboration of AP-CITROX procedure. Therefore, the non-regenerative version of the technology utilized at Paks NPP can be considered to be not an adequate method for the chemical decontamination of any reactor equipments having large steel surfaces (e.g., SGs). As a consequence of the lack of the appropriate decontamination method, initiation of a R&D project focused on the elaboration of the required technology should not be postponed. In this paper, we present a brief overview on the fundamental issues of the technology development. Selected findings obtained in our laboratory on the field of the improvement of the AP-CITROX technology are also reviewed in order to demonstrate the crucial role of some selection criteria.

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Journal of Radioanalytical and Nuclear Chemistry
Authors:
K. Varga
,
Z. Németh
,
J. Somlai
,
I. Varga
,
R. Szánthó
,
J. Borszéki
,
P. Halmos
,
J. Schunk
, and
P. Tilky

Abstract  

During the optimization of the AP-CITROX decontamination technology the effect of the different flow rates of the decontamination solutions on the radioactive contamination and corrosion state of stainless steel tube samples originating from steam generators of Paks NPP were studied by a pilot-plant circulation system. The results have proved that a significant increase (up to 2.89 m/s) in the flow rate of the decontamination solution in the 1-5 steps is highly recommended and in order to improve the passivity of the surfaces it should be kept as low as possible (0.5 m/s) during the passivation.

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Low velocity surface layers can significantly increase ground accelerations during earthquakes. When saturated sandy sediments are present, because of pore pressure increase, decrease of soil strength or even liquefaction can occur. Some volume change follows the dissipation of excess pore pressure after the earthquake resulting surface settlements. To determine the liquefaction probability and post-liquefaction settlement is very important for critical facilities e.g. for the site of Paks Nuclear Power Plant, Hungary. Pore pressure increase and so the liquefaction and surface settlements depend on the characteristics of seismic loading and soil parameters. To quantify the extent of these phenomena is rather difficult. Uncertainties arise both from the probabilistic nature of the earthquake loading and from the simplifications of soil models as well. In the paper, the most important semi-empirical and dynamical effective stress methods for liquefaction and post-liquefaction settlement assessment are summarized. Most significant contributors to the uncertainties are highlighted, and particular examples through the investigation of Paks NPP site are given. Finally, a probabilistic procedure is proposed where the uncertainties will be taken into account by applying a logic tree methodology. At the same time, the uncertainties are reduced by the use of site-specific UHRS and stress reduction factors.

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Abstract  

Within the frame of a joint project, the accumulation of the uranium and transuranium (TRU) species on some structural materials used at Soviet made VVER-type pressurized water reactors (such as heat exchanger tube of steam generators and stainless steel canister material) has been studied. The experiments were carried out in a laboratory model system. During the sorption studies, boric acid coolants provided by the Paks Nuclear Power Plant (Paks NPP) were circulated for a period of 30 h. Solution and tube samples obtained in the course of above experiments were analyzed by independent methods (α- and γ-spectrometry, ICP-MS, SEM-EDX, voltammetry and XPS). The experimental results reveal that: (i) the surface excess of the TRU nuclides studied is extremely low (less than 1% of a monolayer coverage); (ii) the surface excess of uranium species measured on the SG tube surfaces is significantly higher, after 30 h sorption period (Γsample = 1.0 μg cm−2 U ≅ 3.7 × 10−9 mol cm−2 UO2) exceeds a monolayer coverage; (iii) the mechanistic features of the contamination processes (specific or non-specific adsorption, deposition of colloidal and/or disperse particles) depend decisively upon the nature of the studied radionuclides and the chemical structure and composition of the oxide layer formed on stainless steel surfaces.

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Abstract  

The surface contamination by uranium and transuranium (Pu, Am, Cm) nuclides in the primary circuit of pressurized water nuclear reactors is a fairly complex problem as (i) different chemical forms (molecular, colloidal and/or disperse) of these atoms can be present in the boric acid coolant, and (ii) only limited information about the extent, kinetics and mechanism of uranium and transuranium (TRU) accumulation on constructional materials is available in the literature. A comprehensive program has been initiated in order to study the accumulation of uranium and TRU species on some structural materials used at Soviet made VVER-type pressurized water reactors (such as heat exchanger tube of steam generators and stainless steel canister material). This paper, which is the first part of a series of two, provides a comprehensive view on the main experimental parameters influencing the extent and character of the surface contamination by uranium and TRU nuclides. Specifically, we give a brief summary of relevant literature data on the chemistry of uranium and TRU elements and review the dominant chemical forms and their relative sorbability on austenitic stainless steel and Zr(Nb) alloy surfaces. Moreover, some findings on the distribution of uranium, plutonium, americium, and curium species in a model solution of boric acid coolant obtained by the VisualMINTEQ computer code are also presented and discussed.

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Abstract  

A real specimen originating from the primary circuit of a VVER-440 type pressurized water cooled nuclear reactor has been studied by Conversion Electron Mössbauer Spectroscopy (CEMS) in order to find out how the AP-CITROX decontamination procedure modifies the structure and composition of the surface oxide layer of stainless steel which is used in the steam generator. Other methods like voltammetry, gravimetry, and SEM-EDAX were also applied to characterize the samples and to help the interpretation of CEMS results. It was found that, in contradiction with expectations, the presence of the surface magnetite layer could not be convincingly identified even on the non-decontaminated sample. This finding together with the relatively weak Mössbauer signals indicated that the surface oxide layer is strongly Fe-depleted. It was also concluded that the upper layer of the bulk steel (under the oxide layer) has an altered composition probably due to irradiation-enhanced diffusion of the metallic constituents. It was established that the AP-CITROX decontamination procedure does not exert detrimental effects on the thickness and composition of the surface oxide layer.

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Abstract  

The aim of the present work was to reveal the kinetics of the accumulation of some possible contaminant on the surfaces of structural materials (zirconium alloys and 08H18N10T stainless steel) in the primary circuit of Paks NPP. The kinetics of adsorption and desorption of iodide, caesium and cerium ions were investigated by quartz crystal microbalance (QCM) installed into a flow cell. The results on thin layers were confirmed by immersing experiments, using radiotracer technique and γ-spectrometry to detect the traced ions on the surfaces. Experiments were carried out in electrolyte solution which was similar to the cooling water. All measurements were carried out at room temperature. Both adsorption and desorption were found to be fast, taking only several seconds; time constants were also evaluated.

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Journal of Radioanalytical and Nuclear Chemistry
Authors:
Zoltán Németh
,
Bernadett Baja
,
Krisztián Radó
,
Emese Deák
,
Kálmán Varga
,
Andrea Nagy
,
János Schunk
, and
Gábor Patek

Abstract  

Decontamination technologies are generally developed to reduce the collective dose of the maintenance and operation personnel at nuclear power plants (NPP). The highest efficiency (i.e., the highest decontamination factors) available without detrimental modification of the treated surface of structural material is the most important goal in the course of the application of a decontamination technology. At the Paks NPP the AP-CITROX procedure has been utilized for the decontamination of the primary coolant circuit’s components (e.g., main circulating pump (MCP) and steam generators (SGs)). Our previous studies have revealed that a ‘hybrid’ structure of the amorphous and crystalline phases was formed in the outermost surface region of the austenitic stainless steel tubes of SGs as an undesired consequence of the industrial application of the AP-CITROX decontamination technology during the period of 1993–2001. In this paper, we report some comparative findings on the corrosion and surface chemical effects of the AP-CITROX procedure and the novel decontamination technology elaborated at our institution. On optimizing the operational parameters the latter technology may become suitable for the effective decontamination of both dismountable (e.g., MCP swivel) and separable (e.g., SGs) equipments. For this purpose experiments were performed. In this laboratory scale experiments, the passivity, morphology and chemical compositions of the treated surfaces of tube specimens were investigated by voltammetry, and SEM–EDX methods, respectively. The SEM–EDX results have revealed that the oxide removal is surprisingly uniform even after 2 or 3 consecutive cycles. The electrochemical studies have provided evidences that no unfavorable tendencies in the general corrosion state of the tube samples can be detected in the course of the chemical treatments.

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Journal of Radioanalytical and Nuclear Chemistry
Authors:
P. Ormai
,
A. Fritz
,
J. Solymosi
,
I. Gresits
,
E. Hertelendi
,
Z. Szúcs
,
N. Vajda
,
Zs. Molnár
, and
P. Zagyvai

Abstract  

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.

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] Ove Arup & Partners, Paks NPP site investigation for site response and liquefaction potential , Report, Archives, Paks NPP, 1995 . [18] Bán Z

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