ABSTRACT 1. INTRODUCTION

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1 IMPLEMENTATION OF DAVIES & GRAY/NBL METHOD FOR POTENTIOMETRIC TITRATION OF URANIUM IN THE SAFEGUARDS LABORATORY OF CNEN BY THE USE OF A DL - 67 METTLER TITRATOR Radier Mário Silveira de Araújo and Pedro Dionisio de Barros Laboratório de Salvaguardas - Comissão Nacional de Energia Nuclear. Avenida Salvador Allende s/n , Rio de Janeiro RJ. radier@ird.gov.br, pedrodio@ird.gov.br. ABSTRACT To meet the requirements of the Brazilian State System of Accounting for and Control of Nuclear Materials SSAC, the Safeguards Laboratory of CNEN - LASAL has been applying the Davies & Gray/NBL method for potentiometric determination of total uranium concentration in uranium samples taken during safeguards inspections at nuclear facilities since 1984, using a RADIOMETER ETS 822 titrator. In order to improve the analytical capability and the procedures related to the titration methodology, the same method was also implemented by using a METTLER DL - 67 titrator. This equipment is microprocessor - controlled and can be connected to additional devices such as printers, analytical balances, etc. It also provides accurate and reproducible results for end-point titrations, providing analytical performance according to the current international safeguards requirements. The implementation of the method in such equipment included the addition of analytical data as well as the improvement of the equipment parameters for uranium determination. Parameters like predispensing volume; titrant data and end-point value were studied. Some uranium samples (solids and solutions) were used during the initial tests with the titrator. A solution of pure uranyl nitrate was used as reference sample for this paper. From this, aliquots were analyzed in both Radiometer ETS-822 and METTLER DL-67. Results obtained from each equipment were compared with the reference value of the sample. The comparison showed that results from METTLER DL-67 meets the precision and accuracy requirements for this kind of analysis and led to the conclusion that the performance of this titrator is adequate for the determination of total uranium content in samples of nuclear materials for safeguards purposes. 1. INTRODUCTION One important activity conducted by the Brazilian SSAC is the independent verification of nuclear materials by non-destructive and destructive analysis. For destructive analysis, the Davies & Gray/NBL method [1,2,3] for determination of total uranium concentration has been used by LASAL since Along these past years LASAL has obtained accurate and precise results of total uranium content in nuclear materials, in order to verify operators declarations as well as their measurement systems. In addition, LASAL has also been participating in international intercomparison programs [3,4,5,6,7], sponsored by the several international organizations such as Agência Brasileiro-Argentina de Contabilidade e Controle de Materiais Nucleares - ABACC, Direction de L'Energie Nucleaire CEA and New Brunswick Laboratory NBL/US-DOE. A Radiometer ETS-822 titrator has been used for uranium analysis of all kinds of sample along this period. However, this is semiautomatic equipment with some handling difficulties during the titration procedure. For instance, it may cause excessive delay on the titration, causing reoxidation of U (IV). Thus, a new fully automatic titratror the METTLER DL-67 has been acquired in order to improve the measurement procedure and analytical performance. The equipment uses software that permits to store the entire method parameters for analysis of uranium as well as other methods. First, all characteristics of the equipment were

2 investigated. Then, the Davies & Gray/NBL method was loaded in the titrator by using recommended parameters. Several analyses of different uranium compounds were made to verify its performance. Both titrators analyzed aliquots taken from the same solution. Results obtained were compared with the reference value for the uranium concentration in the reference solution. 2. DESCRIPTION OF THE METHOD Potentiometric Davies & Gray/NBL [1,2,3] method for determination of uranium is a selective method based on the reduction of U (VI) to U (IV) in a concentrated solution of phosphoric acid by excess of Fe (II) in sulfamic acid media. This excess of ferrous ions is oxidized with nitric acid in the presence of Mo (VI) and sulfamic acid. Then U (IV) is titrated with a standard solution of K2Cr2O7 until a preset end point potential of 130 mv. To precise this potentiometric end-point, vanadyl sulfate is also added. In this method it is essential to avoid any reoxidation of U (IV) before the titration. Thus, all reactions times must be rigorously controlled and the time between the preparation of the aliquot and the titration. This time shall not be longer then 5 minutes. Besides, the presence of As (III), Sb (III), halides and organic material should also be avoided. Another important feature to assure the good performance of the method is the range of uranium concentration, which must be kept between 90 and 110 mgu/g of solution. A wire of Pt-Rh (90:10) was used as indicator electrode and the reference electrode was a mercurous one. This method may be applied for uranium analysis of oxides, nitrates, ADU, AUC of any physical form [5]. 3.OPERATION OF METTLER DL-67 The Mettler DL-67 is a microprocessor-controlled analytical titrator [8]. It can control up to two burette drivers. One method can be loaded at a time. The operation of the titrator is menu driven. It stores, for instance, titrant identification and the corresponding concentration. Besides, it interrupts the titration after an equivalence point has been found, rinses burette and can send sample s data and results to a printer. Burette and stirrer are installed in defined drives. Titrant data and the type of electrodes used for indication and reference are stored in the proper menu. For analysis, either methods developed by Mettler or a new one loaded in the equipment memory can be used. For this purpose, firstly the method data must be installed in the proper menu of the equipment. Samples data must be loaded every time they are to be analyzed. The determinations results as well as statistical analysis are presented in a sheet, printed at the end of the analysis. 4. ANALYSIS AND RESULTS For the determinations, a solution of pure uranyl nitrate was used. From this, four aliquots were analyzed in the Radiometer ETS-822 titrator and three aliquots in the Mettler DL-67. In the Mettler titrator it was also necessary to load instrument parameters such as: predispensing volume which may be about 60% of the theoretical volume to be added to perform the titration; information about addition of volume increments just before the endpoint; maximum dispensing volume which was pre-fixed in 40 ml; end-point (pre-fixed in 130 mv), etc. The solution concentration must be adjusted to be in the range recommended

3 for this determination. Aliquots were taken and prepared according to the conditions mentioned in section 2. In order to avoid any calibration error from volumetric flasks and pipettes, this method is performed in a weight basis. As a consequence, the burettes must be calibrated in a weight basis to convert the volume in mililiters of potassium dichromate dispensed into grams of the titrant. The results were calculated by the expression: C=V.F.T/m (1) Where: C=concentration of uranium in each aliquot, expressed in mgu/gsol. V=volume of the titrant, in ml. T=titer of the titrant, in mgu/gsolution. m=weight of the aliquot, in grams. The results obtained are showed in the Table 1. Table 1 Results of uranium concentration obtained under the same conditions for aliquots from the same sample in both titrators METTLER DL-67 Results (mgu/gsolution) RADIOMETER ETS-822 Results (mgu/gsolution) Mean: Mean: Standard deviation: Standard deviation: Coefficient of variation: 0.089% Coefficient of variation: 0.035% The results in table 1 were obtained at the same analytical conditions including room temperature, analytical solutions and analytical balance. Four aliquots were titrated in Radiometer ETS-822 and three aliquots in Mettler DL-67 equipment. Outlier tests of Dixon and Grubbs applied to the results of both equipments did not any statistical outlier [9]. The reference value provided by DIRECION DE L'ENERGIE NUCLEAIRE for this solution is (189.26±0.19) mgu/gsolution.

4 5. DISCUSSIONS AND CONCLUSIONS The value of the titrant s concentration must be expressed as titer (in mgu/gsolution) instead of normality and the value which can be loaded in the titrator is normality with four decimal places. To achieve the precision and accuracy adequate to this kind of analysis it was necessary to perform the calculations using concentration expressed as a titer (in mgu/gsolution) with five decimal places in an electronic calculator because it affects the results performance in a significant manner. The means and standard deviations of the results of the uranium solution analysis were statistically compared. The means were compared by use of a Student t-test and the standard deviations were compared by F-test. The two tests were performed at 95% probability level. Neither the means nor the standard deviations have presented significant statistical differences at this probability level. The statistic criteria for this analysis say that any aliquot analysis result shall be compared with the others of the same sample. The relative error of the results between them should be less or equal to 0.14% relative. Results lying in these limits are accepted and reported. The uncertainty was calculated for both titrators at a confidence interval of 95% [9,10] by the equation 2: Where: I: uncertainty; t: Student factor; s: standard deviation; n: number of determinations. I = ± t.s/ n (2) The uncertainty values calculated from the values of Table 1 were: mg/g for Mettler DL-67 and mg/g for Radiometer ETS-822. Detailed results are presented below: Mettler DL-67: ( ± )mg/g Radiometer ETS-822: ( ± )mg/g The values of random and systematic components were: Mettler DL-67: random= 0.05%; systematic= 0.05% Radiometer ETS-822: random= 0.02%; systematic= 0.02% These values are according to the international target values for uncertainties associated to safeguards measurements [11]. Considering all the results obtained and its evaluation, the titrator Mettler DL-67 can be considered adequate for potentiometric analysis of uranium for safeguards purposes. REFERENCES 1. A.R. Eberle, M.W. Lerner, C.G.Goldbeck, C.J. Roden NBL-252, p.1 (1970). 2. E. Kuhn, S. Deron, H. Aigner, A. Zoigner Destructive Analysis of Safeguarded Materials, IAEA/RL/62, pp (1979). 3. R.M.S.Araújo, S.G.de Almeida, S.V.Gonçalves, J.H.B.Bezerra, M.M.Duarte, Avaliação dos Resultados Obtidos em Análises de Amostras de Materiais Nucleares Colhidas em Inspeções de Salvaguardas e em Amostras de Intercomparação.

5 Livro de Resumos do XXXVI Congresso Brasileiro de Química, São Paulo, (1996). 4. I. Frank, M.A. Legel. Determination of Uranium by Ferrous Reduction in Phosphoric Acid and Titration with Dichromate (NBLTitrimetric Method) pp.1-16 (2004). 5. R.M.S.Araújo, J.H.B.Bezerra, The Brazilian Safeguards Analytical Laboratory. Proceedings of the VI ENAN, Rio de Janeiro (2002). 6. V.Verdingh, Y.Le Duigou, Interlaboratory Comparission Exercise for the Determination of uranium by Potentiometric Titration (first phase), Technical Report Joint Research Center GEEL, Belgium (1990) C. Roche, Programe D'Evaluation de la Qualite du Resultat D'Analyse Dans L'Industrie Nucleaire Programe EQRAIN Uranium N o 10, Note Technique DRCP/CETAMA/2005/02 pp.1-24 (2005). 8. Mettler Toledo DL67 Titrator Operating Instructions pp.1-40 (1993) 9. A. M. A. Pereira Estatística Básica Aplicada ao Laboratório pp.1-95(2004) C. Roche, Programe D'Evaluation de la Qualite du Resultat D'Analyse Dans L'Industrie Nucleaire Programe EQRAIN Uranium N o 10, Note Technique DRCP/CETAMA/2005/02 pp.1-24 (2005). 11. Aigner, R. Binner, E. Kuhn, U. Blohm-Hieber, K. Mayer, S. Guardine, C. Pietri, T. Adachi, B. Rappinger, B. Mitterrand, J. Reed, O. Mafra-Guidicini, S. Deron, International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials, IAEA/STR-327, pp.24 (2001).