logo2In the presented paper, the basic characterization of metal melting, described as a perspective decommissioning waste management method for radionuclides effective immobilizing and decontamination of metals up to clearance levels, is given. The materials with radioactivity below the clearance levels are releasable to the environment without any further restrictions. To evaluate the impact of metals melting application on the clearance process, it is necessary to monitor the physical and nuclide resolved radiological characteristics of the individual material items during the whole decommissioning.

These characteristics are implemented into the analytical calculation code OMEGA, especially into the integrated material flow tool that could be used for the evaluation of various options of melting application in the decommissioning waste management process. The possibilities of reaching the clearance levels, by applying the decontamination character of melting, are the main results discussed in the paper.


One of the characteristic feature of nuclear installation (NI) decommissioning process is production of large amount of various radioactive and also non-radioactive waste that have to be managed and released from the former NI area, taking into account their physical, chemical, toxic and radiological properties. Therefore, the safe and economic implementation of the waste management plan is considered to be one of the key issues in the frame of the decommissioning. In the waste management process several methods and technologies are used to reach the two main goals:

  • Release the materials to the environment (ENV) for further use and then minimize the volume of radioactive waste (RAW) – decontamination techniques;
  • Safe isolation of the non-releasable materials from the environment within the radioactive waste repository barriers – waste treatment and conditioning techniques.

The melting technology, specified by international organizations as a promising method for metal waste (materials) processing (IAEA, 2008), could be included into both of two mentioned groups. In the paper the melting technology, applied as a decontamination method used for reducing the radioactivity of metals up to clearance levels, is analysed.


Melting is a high-temperature technology that completely destroys the metal components and redistributes the radioactivity among:

  • Ingots as a primary product representing the main mass flow. Ingots are further managed according their characteristics.
  • Slag as a secondary solid radioactive waste representing 1-4% of melted scrap weight. The slag has to be further treated and immobilized as a radioactive waste.
  • Dust as a secondary radioactive waste is exhausted into the air conditioning system and before discharging to the ENV is absorbed on the high-efficiency filters.

The distribution of radionuclides during melting is a complex process that could be influenced by numerous chemical and physical factors, including the composition, solubility of an element in molten metal, density of the oxides, composition and basicity of the slag former, melting temperature and melting practices such as the furnace type and size or melting time.

Melting is considered to be the final and usually irreversible step in the metal radioactive waste treatment. The advantages of melting application in the decommissioning waste management process should be summarized as follows (IAEA, 2008; IAEA 2001; Min, et al., 2009):

  • Decontamination of the melted metals reached by the effective separation of radionuclides from the metal waste and their redistribution to the slag and to the dust. The decontamination efficiency varies widely, depending on the radionuclides present. It is also possible to reach the clearance levels and release the ingots to the ENV for further use.
  • Homogenous distribution and effective immobilizing of radionuclides in the ingots volume (no surface contamination), thus the possibility of contamination spreading is significantly reduced.
  • Volume reduction, thereby the storage or disposal capacities are preserved.
  • Precise determination of specific mass radioactivity and easier comparison with clearance levels by sampling each ingot.
  • Simplified monitoring procedures for characterization of radioactive materials with complex geometry. The problem of inaccessible surfaces is eliminated.
  • Suitable form and shape (ingots) for further use of metals e.g. re-melting and reuse outside or within nuclear industry, decay storage.


Clearance is defined as the removal of radioactive materials within authorized practices from any further regulatory control applied for radiation protection purposes. The concept of clearance expects that once cleared materials are subject to no further radiological restriction, may be treated as a normal waste and recycled/reused in any other industrial area (OECD/NEA, 2008).

Clearance is based on the concept of triviality of exposure, generally taken to mean that radiation risks to individuals and collective radiological impact, caused by the cleared material, is sufficiently low. In quantitative terms, the mentioned state is related to the stipulation that the effective dose expected to be incurred by any member of the public due to the cleared materials is of the order of 10 \mu Sv or less during the one year (OECD/NEA, 2008).

In the Slovak legislation – Statutory Order No. 345/2006 on the Basic Safety Requirements on Personnel and Public Health Protection against Ionizing Radiation, based on the 10 \mu Sv principle, the surface contamination and mass radioactivity limits for unconditional release of radioactive materials to the ENV are defined. These limits could be understood as a clearance levels (Table 1.) defined for individual classes of radionuclides radiotoxicity (Table 2.).

Table 1. The clearance levels for releasing the materials to the environment

Radiotoxicity category
1 2 3 4 5
Clearance level for mass radioactivity [kBq.kg-1] 0,3 3 30 300 3000
Clearance level for surface contamination [kBq.m-2] 3 30 300 3 000 30 000

Table 2. Categories of radionuclides radiotoxicity

Category Radionuclide
1 22Na, 24Na, 54Mn, 60Co, 65Zn, 94Nb, 110Ag, 124Sb, 134Cs, 137Cs, 152Eu, 210Pb, 226Ra, 228Ra, 228Th, 232Th, 234U, 235U, 238U, 237Np, 239Pu, 240Pu, 241Am, 244Cm
2 58Co, 59Fe, 90Sr, 106Ru, 111In, 131I, 192Ir, 198Au, 210Po
3 51Cr, 57Co, 99Tc, 123I, 125I, 129I, 144Ce, 201Tl, 241Pu
4 14C, 32P, 36Cl, 55Fe, 89Sr,90Y, 99Tc, 109Cd
5 3H, 35S, 45Ca, 63Ni, 147Pm


The analytical calculation code OMEGA (Oracle Multicriterial General Assessment of Decommissioning), developed in the DECOM Company, is characterized as a planning tool for evaluation of decommissioning parameters (costs, manpower, exposure, consumptions, duration of individual decommissioning tasks, effluents volume or radioactivity, amount of released or disposed materials etc.) and optimization of individual decommissioning options in the decision making process (Daniška and Nečas, 2000).

Analysis and control of material and radioactivity streams during the whole decommissioning process (from dismantling up to release to the ENV or disposal within the radioactive waste repository barriers) is in the OMEGA code performed by using an integrated material flow tool where the material parameters are linked together with the nuclide resolved radiological characteristics of each material item. In the calculations, following attributes are taken into consideration (Daniška, et al., 2008; Zachar and Nečas, 2008):

  • Physical parameters of material items from inventory database (mass, volume, surface, category etc.);
  • Radiological parameters of materials (inner and outer surface contamination, induced or mass activity, dose rates);
  • Nuclide vectors characterising the relative shares of individual radionuclides on the total contamination, induced activity or dose rate;
  • Distribution coefficients characterizing the impact of all decommissioning activities on the materials and radioactivity distribution and on the secondary waste and effluents generation;
  • Limits and conditions for waste treatment and conditioning facilities, for materials to be released to the environment, for radioactive waste disposal;
  • Radioactive decay of nuclides.


The OMEGA code and the database of the primary circuit technological equipment (without reactor and its internal parts) and the reactor building auxiliary systems of VVER-440 nuclear power plant (NPP) was used for the model calculation.

5.1 Input parameters of the calculation assessment

To evaluate the possibilities of melting application in the decommissioning materials clearance process various model scenarios were calculated. Following input parameters of decommissioning were changed:

  • Surface contamination nuclide composition (nuclide vector) characterising the radiological situation of the primary systems at the end of NPP operation:
    • Nuclide vector characterising the NPP with standard ended operation without any serious accident that caused spreading of the fission product to the primary circuit. The dominant contaminant is activation product 60Co (marked as “SO” – standard operation).
    • Nuclide vector characterising the NPP with fuel accident where the fuel elements were damaged and primary circuit systems was, except of activation products (60Co), contaminated also with fission products as 137Cs, 90Sr and also alpha active nuclides e.g. 241Am, 238Pu, 239Pu (marked as “AO” – operation with accident)
  • Time period of melted ingots long term storage. To achieve the clearance levels, the time decontamination principle applied for following time periods were used in the calculation options:
    • No storage period of ingots (marked as “0Y);
    • 5 years storage period (marked as “5Y”);
    • 10 years storage period (marked as “10Y”);
    • 20 years storage period (marked as “20Y”);
    • 30 years storage period (marked as “30Y”);
    • 50 years storage period (marked as “50Y”).

In the calculation, the following initial assumptions were considered:

  • No radioactivity or mass (volume) restrictions for metal melting technological facility.
  • The 10 years long transition period after final NPP final shutdown (Fig. 1). During this time spent fuel have to be removed to interim storage facility, the parts of non-radioactive civil building should be decommissioned and the preparation activities before the radioactive systems dismantling are also done. Within the transition period (as for ingots long term storage) the radioactive decay of nuclides is considered.
  • No radioactive decay of nuclides between dismantling and melting (Fig. 1). It means that all database items are dismantled, melted, released to ENV or disposed in the radioactive waste repositories in one day.
  • Distribution coefficients for melting technology to ingots and secondary waste and ingots clearance levels (according to Slovak legislation as mentioned in chapter 3) for representative radionuclides as shown in Table 3.

Table 3. Characteristics of dominant contaminants

Radio-nuclide Nuclide vector Distribution coefficients Clearance level [Bq.kg-1]
Ingots Secondary waste
60Co SO,AO 0,15 0,85 300
63Ni AO 0,1 0,9 3 000 000
137Cs AO 0,99 0,01 300
90Sr AO 0,99 0,01 3 000
Alpha nuclides AO 0,99 0,01 300

5.2 Output parameters of the calculation assessment

For the melting application analysis, in terms of metals clearance process, the mass and radioactivity distribution of steel (carbon and stainless) items from the inventory database was evaluated. The total mass of steel was, based on the radiological parameters and subsistent limits, distributed into the following material flow streams (Fig. 1):

zachar09 1
Fig. 1 The materials flow streams and time aspects of model calculation

  • Release of steel to the environment just after dismantling. The characterization applied after dismantling unambiguously demonstrates that the clearance levels for surface contamination and mass activity are fulfilled (SC
  • Release the ingots to the environment after melting application. The decontamination character of melting or homogenization of surface contamination into the ingots volume could lead to achieving the clearance levels for mass activity (MA
  • Release the ingots to the environment after melting and long term storage application (if the calculation scenario covers the storage). Using the time decontamination principle for melted ingots, the clearance level for mass activity could be achieved.
  • Disposal of melted ingots to the radioactive waste repository. If the characterization after melting or long term storage does not demonstrate the fulfilment of the legislatively prescribed conditions for releasing the steel ingots to the environment (MA>L_ENV), they have to be disposed to the RAW repository.


Considering the initial assumptions and various decommissioning input parameters mentioned in the previous chapter, the model calculation results evaluating the possibilities of releasing the materials by melting application are presented and further discussed in this chapter.

Analysing the calculation scenarios with different surface contamination nuclide vectors (Fig. 2), it is possible to conclude, that melting application leads to releasing more materials in the case of worse nuclide composition due to fuel accident. The reason is higher decontamination factors for volatile nuclides such as 137Cs (dominant contaminant after 10 years transition period). Almost all radioactivity of caesium is after melting application transferred to slag or dust, the radioactivity of partially decontaminated ingots is below clearance level (about 14% of total steel mass before dismantling) and could be released to the environment.

On the other side, the better radiological situation at the end of operation period in scenario S0_0Y do not caused the higher ratio of cleared materials (almost 50% of steel has to be disposed as a RAW). The reason is that, 85% of 60Co (dominant contaminant for SO nuclide vector) radioactivity is fixed in the ingots volume and then the clearance levels could not be fulfilled.

zachar09 2
Fig. 2 Steel distribution for various surface contamination nuclide composition

The increasing ratio of released ingots for SO nuclide vector could be seen in the case of long term ingots storage application (Fig. 3). Because of 60Co relatively short half-time (5,27 years), the radioactivity of ingots goes down relatively fast. Just after 5 years ingots storage period, the ratio of cleared material is at the same level as for the scenario AO_0Y. The storage period extension leads to accrual of cleared ingots ratio and after 50 years, the ratio of RAW disposed to the repository is reduced to about one third of the value calculated for the scenario without ingots storage (SO_0Y).

zachar09 3
Fig. 3 Steel distribution for various time periods of long term ingots storage (for SO nuclide vector)

The different situation is for the scenarios where nuclide vector AO and long term ingots storage is considered (Fig. 4). Due to relatively long half life of dominant radionuclide 137Cs, the time decontamination principle could not have the significant impact on ingots clearance process. Comparing the scenarios without and with 50 years long storage period, the increase of cleared ingots is only 3,5%.

zachar09 4
Fig. 4 Steel distribution for various time periods of long term ingots storage (for AO nuclide vector)

In the Table 4., the results of calculation analysis for each scenario are summarized. The mass of cleared dismantled steel parts, cleared melted ingots and disposed RAW are shown. In the last column, the estimated number of 0,2 m3 steel drums (the capacity of each drum is 350 kg of steel) that have to be disposed to the RAW repository is given. From this data, it is possible to determine the repository capacity that could be preserved applying the melting and long term storage in the decommissioning materials management and clearance process.

Table 4. Summarization of cleared materials and disposed RAW for each calculated scenario

Scenario Cleared metals Cleared ingots RAW
Mass [tons] Mass [tons] Mass [tons] Number of Drums
SO_0Y 4 040 227 4 027 11 505
SO_5Y 4 040 1 131 3 123 8 924
SO_10 4 040 1 314 2 940 8 400
SO_20 4 040 1 325 2 930 8 370
SO_30 4 040 1 492 2 762 7 893
SO_50 4 040 2 552 1 703 4 864
AO_0Y 4 030 1 141 3 123 8 924
AO_5Y 4 030 1 325 2 940 8 400
AO_10 4 030 1 331 2 934 8 382
AO_20 4 030 1 331 2 934 8 382
AO_30 4 030 1 331 2 933 8 381
AO_50 4 030 1 442 2 822 8 063


The possibilities of metal melting technology application in the decommissioning waste management process, especially in terms of metals clearance, are discussed in the paper. The issues of decommissioning materials parameters evaluation, using the analytical calculation code OMEGA, are also shortly introduced.

Based on the model calculation results, presented in the chapter 6, it is possible to conclude, that the application of metal melting in the clearance process strongly depends on the radiological situation at the end of nuclear power plant operation period, represented by the surface contamination nuclide composition. The advantages of melting decontamination character could be used only in the case, if the volatile radionuclides such as 137Cs or 90Sr are occurred in the radiological composition of dismantled material items. Considering the standard operated power plant (60Co is dominant contaminant), the melting could be significantly applied in the clearance process only with the combination with other process like long term ingots storage within the appropriate facility. Another option could be using the chemical, electro-chemical or dry (blasting) post-dismantling decontamination technology to decrease the contamination level. Followed by melting, the clearance of metals, in suitable form and shape (ingots) for further reuse in the environment, could be achieved.

Finally, it is necessary to stress the importance of radiological characterization of the power plant inventory in the decommissioning planning period. The inadequate fulfilment of characterization tasks could lead to significant obstruction in the clearance process. Homogenization of the radioactivity in the ingots volume allowed precise determination of specific mass radioactivity. Knowing the radioactivity nuclide composition, it is easy to predict the necessary period of ingots storage until the clearance levels are reached.


This project has been partially supported by the Slovak Research and Development Agency through the grant No. APVV-0761-07 and by the Slovak Grant Agency for Science through the grant No. VEGA 1/0685/09.


  1. Daniška, V., Rehák, I., Vaško, M., Krištofová, K., Bezák, P., Ondra, F., Zachar, M., Tatranský, P., Schultz, O., Nečas, V. Comparative Analysis of Decommissioning Technologies Based on Model Calculations and Multi-attribute Analysis of Specific Decommissioning Cases of Nuclear Facilities. In: Innovative and Adaptive Technologies in Decommissioning of Nuclear Facilities: Final report of a coordinated research project 2004-2008. IAEA-TECDOC-1602. IAEA, Vienna, 2008. p.217-249.
  2. Daniška, V., Nečas, V. Calculation modelling of the decommissioning process of nuclear installations. In: Journal of Electrical Engineering, Vol. 51 (2000), No. 5-6, p. 156-167.
  3. International Atomic Energy Agency. Managing Low Radioactivity Material from the Decommissioning of Nuclear Facilities: Technical report series No.462. IAEA, Vienna, 2008.
  4. International Atomic Energy Agency. Methods for the Minimization of Radioactive Waste from Decontamination and Decommissioning of Nuclear Facilities: Technical report series No.401. IAEA, Vienna, 2001.
  5. Min, B.Y., Choi, W.K., Oh, W.Z., Jung, CH.H., Lee K.W. Partition characteristics of radionuclides during a melt decontamination of a contaminated metal waste. In: Journal of Industrial and Engineering Chemistry. Vol. 15 (2009), p. 31–35.
  6. Organisation for Economic Co-operation and Development/The Nuclear Energy Agency. Release of Radioactive Materials and Buildings from Regulatory Control: A Status report. OECD/NEA, Paris, 2008.
  7. Zachar, M., Nečas, V. The Optimization of Radioactive Waste Management in the Nuclear Installation Decommissioning Process. In: International Youth Nuclear Congress 2008 [USB memory stick]. Interlaken, Switzerland, 2008. p.432.1-9.

Co-author of this paper is Vladimír Nečas Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Ilkovičova 3, 812 19 Bratislava.

Napísať príspevok