Behaviors of Cementitious Materials in Long Term Storage and Disposal

1. Research Coordination Meeting on CRP “Behaviour of Cementitious Materials in Multipurpose Packaging for Transportation, Long Term Storage and Disposal” , 24-28 Nov 08, Bucharest, Romania Behaviors of Cementitious Materials in Long Term Storage and Disposal Chief Scientific Investigator: I. Plećaš Additional Scientific Staff: S. Dimović, I. Smičiklas, D. Kićević Vinca Institute of Nuclear Sciences, Belgrade-Vinca, Serbia Radiation and Environmental Protection Laboratory

2. Institute of Nuclear Sciences "Vinca"Belgrade, Serbia Development of Solidification Techniques for Radioactive Sludge Produced by a Research Reactor ILIJA PLECAS,PhD

3. Abstract The reportspresent the results on removal of sludge from the bottom of the spent fuel storage pool in RA reactor, mechanical filtration of the pool water, sludge immobilization by cement , conditioning and storage.

4. The processing of radioactive wastes may be done for : economic reasons (e.g. to reduce the volume for storage or disposal, or to recover a "resource" from the waste), or safety reasons (e.g. converting the waste to a more "stable" form, such as one that will contain the radionuclide inventory for a long time). Typically processing involves reducing the volume of the waste solidifying non-solid wastes to make them physically stable, and packaging the waste to isolate it from the environment.

5. The RA research reactor, (6.5 MW) was shut down in 1984 in order to reconstruct and improve all vital reactor systems. However, for a number of political, administrative, economical and technical reasons, this reconstruction has never been completed.SNF in the storage water pools inside the RA building is leaking !!! To solve the problems, Government decide to proposeVIND Project, (VINČA INSTITUTE NUCLEAR DECOMMISSIONING PROGRAM) Three Sub Projects: 1. Spent Fuel Transport 2. Decommissioning of RA Reactor 3. Radioactive Waste Management at the Vinča site

6. RA reactor SNF room

7. The spent fuel storage pool (Fig. 1a) on the RA research reactor in the “Vinca” Institute consists essentially of four, six meters deep, inter-connected rectangular basins.

8. Pool water radioactivity

9. Sludge sedimentation in the pool

10. Presently, water in the reactor RA- spent fuel storage pool is in very bad condition. Water in the pool is dirty and its chemical parameters are not maintained to minimize corrosion process. Following the recommendations obtained from (IAEA) the “Vinca” Institute elaborated a project incorporating the following steps: preliminary removal of sludge and other debris from the bottom of the pool in RA reactor, washing of deposits from all the surfaces in contact with the pool water, venting of the aluminum barrels, mechanical filtration of the pool water, final removal of the sludge, sludge conditioning and storage at the waste repository at the “Vinca” Institute site. Newer hangar has an space for radwaste materials storing only for 1-2 years. So, attempts are made in the “Vinca” in developing the immobilization process for conditioning low and intermediate level radioactive waste materials and their safe disposal into the appropriate disposal system.

11. The cementation, as an immobilization process, for the certain radwaste materials origin and composition is investigated. Developed immobilization processes have, as a final goal, production of the solidified radwaste matrix mixture form, that is easy for handling and that satisfies safety and QA requirements, according to radionuclide inventory, decay heat, radiation dose rate and contamination, identification, configuration and weight, mechanical integrity for interim storage and the final disposal of such materials on the appropriate sites. Radwaste materials that were immobilized in the inactive matrixes are to be placed into the concrete containers, for the further management and disposal.

12. 2. Characteristics and quantities of radioactive sludge in the spent fuel element storage pool In order to estimate storage conditions for the spent fuel elements in the storage pool and characteristics and quantities of the sludge on the bottom of the pool, water and sludge samples have been taken out from different locations in the pool. Analysis of the water from the pool (pH = 8.4, electrical conductivity = 446 µS/cm, [Cl] = 66 mg/L, [Cu] = 0.05 mg/L, [Zn] < 0.01 mg/L, [Fe] = 0.15 mg/L, [SO4 ] = 55 mg/L) shows that the water is highly corrosive to aluminum alloys . Activity concentration of the water from the pool of about 80 - 90 kBq/L of 137Cs nuclide, although not of grave concern, is certainly significant, and is incontestable proof that some amount of the fission products is leaking.

13. 3. Sludge conditioning and storage Total quantity of sludge on the bottom of the RA research reactor spent fuel storage pool was estimated to be about 3 m3. Estimation was made on the basis of the average sludge height on the bottom of the pool and pool surface. The sludge color has been a dark red - brown, like an iron oxide corrosion products. Gamma spectrometry analysis showed that the specific activity of the sludge is about 1.8 ± 0.2 MBq/L from 137Cs nuclide and about 15 kBq/L from 60Co nuclide

14. Cascs Based on the previous experience, a technology was developed for sludge immobilization and conditioning in a cement matrix, inside casks, produced using the standard 200 L metal barrels which have lids supplied with screws. Casks have been produced as concrete shielded containers in standard metal barrels. Thickness of the concrete walls is from 8 to 10 cm. Entire inner side of the cylindrical concrete wall is covered by plastic tube with wall thickness of 1 cm, which has been used as a model in forming cylindrical concrete wall. This plastic tube serves as a first barrier in preventing radionuclides leaching from radioactive sludge immobilized in a cement matrix. The bottom cask concrete wall is also 6– 7 cm thick. In order to prevent or reduce radionuclide leaching, this wall has been covered with epoxy resin. The useful volume of such designed cask is about 75 L.

15. The existing pilot cement mixer was reconstructed to enable placing a barrel containing the planned quantity of sludge on its platform without a risk of spilling. About 60 – 65 l of sludge are poured at a time from the sedimentation vessel into a previously prepared cask. As soon as a cask is filled up, it is hermetically covered with a lid supplied with screw and transported to the laboratory for sludge conditioning. There, additional settling of sludge is allowed. Separated water is pumped into a plastic can and taken back to the RA reactor spent fuel storage pool. Through the second stage of the sludge settling, volume of the sludge in the cask has been reduced to about 40 l. Fig. 2. A new mechanical manipulator, which provides mixing of the cement matrix with the sludge in the entire volume of the barrel, was constructed.

16. Fig. 3. Modified pilot mixer with concrete container made in metal barrel. Fig. 2. Apparatus for sludge settling

17. Casks have been produced as concrete shielded containers in standard metal barrels. Thickness of the concrete walls was 10 cm.

18. Well experience Dr Ilija Plećaš, and my team published more than 40 papers in International Journals and more than 90 papers on International Conferences in the field od Radioactive Waste Management, especialy, on the problemas of Immobilization of Rad. Waste in liquid and solid forms with cement.

20. 2. RADIONUCLIDE MIGRATION THROUGH POROUS MATERIALS The dispersion of radionuclides in porous materials, such as grout or concrete, is described using a one dimensional differential model.(Burns,1971, Lu,1978,Moriyama,1977). where: KF - retardation factor (=)1 D - diffusion coefficient (cm2/d) or (cm2/s) A - concentration in liquid (mol/l) or (Bq) X - length (cm) Vv - velocity of leachant fluid (cm/d) f - porosity (=)1 T - bulk density (g/cm3) kd - distribution coefficient (ml/g) t - time variable (d). Using Laplace transformation method, Eq.(1') becomes:

21. from which we can calculate a retardation factor, KF. The coefficient of distribution, kd, can be calculated: in which: Vv, X, T, t and Ao are known. An and De can be determined experimentally using a leaching test procedure.(Hespe,1971) . For the interpretation of the results of leach tests shown in the following figures and tables, leach coefficient D, is used, and it is defined as: where: D - leach coefficient (diffusion) (cm2/d) or (cm2/s); m - (An/Ao)(1/t), slope of the straight line (d-1/2); Ao - initial sample activity at time zero (Bq); (Table I) An - activity leached out of sample after leaching time t, (Bq); t - duration of leaching renewal period (d); (1,2,3,4,5,6,7,15,30,60) V - sample volume (cm3); S - sample surface (cm2).

22. Table I. Representative formulations of concrete compositions as grams, for 1000 cm3 of concrete. The cement specimens were prepared with a standard Portland cement * .

23. Table II Leach coefficients De(cm2/d) in different concrete samples after 60 days, using Eq.(4) Table III Retardation factor KF and coefficients of distribution kd(ml/g), after 60 days, T=2,5 (g/cm3). f=0,15-0,30

24. Compressive strength M(MPa) of samples after 28 days

25. About 60 – 65 L of sludge are poured at a time from the sedimentation vessel into such a previously prepared cask. As soon as the cask is filled up, it is hermetically covered with a lid supplied with screw and transported to the laboratory for sludge conditioning. When the cask with the settled sludge is placed on the platform of the mixer for further conditioning, the necessary amount of cement, according to the established formula of cement matrix and the cement-sludge ratio, are poured into the cask. The formula of cement matrix and the cement to sludge ratio, are defined in accordance with previous experience and experimental investigations on radwaste cementation and and experiments made with this sludge. The best sludge to cement mass ratio for appropriate mechanical strength was approximately 1:1.8

26. Homogeneous substance could be obtained by less than one hour mixing. This technology for sludge conditioning eliminates all contamination and radiation risks related to pouring the sludge into the concrete mixer and pouring the cement-sludge mixture into the metal barrel. The barrel with the homogenized mixture is removed from the mixer platform and placed in a separate room for concrete to harden. It was experimentally determined that the time needed for concrete hardening is about 48 h. Due to dynamics of the of the sludge removal from the spent fuel storage pool, and sludge settling, the final stage of radioactive waste conditioning has took place 7 – 10 days after the sludge cementation.

27. Taking into account the measured sludge activity concentration, two stage sedimentation process (the first one in the vessel for sedimentation and the second in the concrete cask – container) and the conditioning technology, it is estimated that each cask with conditioned sludge contains about 150 - 200 MBq from 137Cs nuclide and about 7 – 10 MBq from 60Co nuclide, i.e., specific volume activity of the conditioned radioactive waste in radioactive waste packages is about 0.7 - 1 GBq/m3 from 137Cs nuclide and about 35 - 50 MBq/m3 from 60Co nuclide. Taking into account composition of radioactive waste packages, the effect of self-absorption in homogeneously dispersed radioisotopes in the cement matrix, and concrete cask walls radiation absorption capability, the contact gamma-ray dose rates, measured on the casks surface, were in the range from 0.1 to 0.15 mSv/h, i.e., much less than 2 mSv/h, which is an acceptable limit value for the radioactive waste packages.

28. Practically, Reduction Factor, RF , was about 33 ( 200m3 of liquid radioactive waste from RA basins, we transferred in 31 concrete casks in metal barrels, each 200 lit., approximately 6 m3 of solidified sludge with concrete protection).

29. 31 concrete casks in metal barrels are disposed in hangar H2, at the existing radwaste repository at “ Vinca “ Institute site. 31 concrete casks

30. 4. Conclusion After the operations, explained above, have been performed, necessary elements for planning further stages of pool and water cleaning and treatment of the spent fuel should be obtained. Through many years research and development in radioactive waste immobilization and conditioning performed experimental experience gave the possibility to choose the best formulation for cement mixture and results gave us certainty to claim that described methods and used matrix materials will serve as a barriers to preserve radionuclides migration to the surroundings for at least 300 years. Optimization of the processes and matrix-radwaste mixtures is in further progress and we hope that this work will influence the design of the future Serbian storage center, shallow land burial type for low and intermediate level radioactive wastes. All performed steps have been done in accordance with all relevant requirements for radiation safety and radiation protection .