Radiological characteristics of activated in-vessel structures, reactor

After final reactor shutdown, the project of the BN-350 decommissioning is required for fulfilment of preparatory work, unloading of all fuel assemblies and control rods from the reactor, drainage of the coolant and washing of the equipment and pipelines from the residues of sodium. Further it will be possible to commence the initial stages of dismantling of some reactor structures. It is necessary to take into account that the unloading of the fuel subassemblies is accompanied by their replacement with the non-radioactive steel dummy subassemblies, which should require in the future washing from the sodium residues and decontamination.

The main amount of solid radioactive wastes is caused by the activated structural materials of the reactor and shielding materials, located in the reactor vessel cavity. Reactor vessel is located in the concrete cavity and is separated from the concrete cavity walls by the lateral radiation shielding, consisting of steel plates and the reinforcing cage filled by the iron ore concentrate. On the foundation concrete beam under the reactor vessel the bottom shielding is located, which is made of materials, similar to those used for the lateral shielding. In the neck of the reactor vessel rotating plugs made of steel sheets, graphite and thermal insulation are located, playing a role of thermal and biological shielding. The upper section of the cavity beyond the reactor vessel neck is blocked by the upper stationary shielding, consisting of steel sheets, space between them being filled with the serpentinite concrete.

The activity of this structure was estimated for 2003 as a conditional date of expiration of the reactor life. Until 1^st of December 1997, the real history of the reactor operation was taken into account, while after this date average capacity of the reactor for the previous period was used.

5.6.2.1. Calculation codes of spatial-power distributions of neutrons and activation of materials [22–24]

The spatial power distribution of neutrons in the reactor and in the ex-vessel shielding structures was calculated by the TWODANT code in RZ-geometry with the use of 26-group data base library ABBN-93. Analytical studies on accumulation of the radionuclides in the main structures of the reactor, shielding and building materials were carried out by the FISPACT code, taking into account irradiation history of materials during 30 years of the reactor operation.

5.6.2.2. Initial contents of elements in structural materials

Impurities were taken into account in the calculations of the radionuclides accumulation in the materials. The experimental data on the impurities contents in the iron ore filling and Cr18Ni9 steel of the BN-350 reactor are presented in Tables 16 and 17. These tables also give expertise estimated data on the contents of some impurities, which were not determined by the experiments, although they can influence the total value of the induced activity.

TABLE 16. THE CONTENTS OF BASIC ELEMENTS AND IMPURITIES IN IRON ORE FILLING (wt%)

* Elements which were taken into account in calculation in addition to those measured.

TABLE 17. THE CONTENTS OF BASIC ELEMENTS AND IMPURITIES IN

* Elements which were taken into account in calculation in addition to those measured.

Experimental studies were not made on the contents of the impurities in standard and serpentinite concrete, thermal insulation of plugs and graphite and 3-grade steel of the BN-350 reactor. Table 18 shows the most complete data on the contents of the impurities in concrete were published in the article [26].

TABLE 18. THE CONTENTS OF BASIC ELEMENTS AND IMPURITIES IN SHIELDING CONCRETE (1.0E-4 wt%) These data on the impurities were taken as the basis for the serpentinite concrete and thermal insulation of the plugs, the composition of the main elements for these materials being presented in Table 19.

TABLE 19. THE CONTENTS OF BASIC ELEMENTS IN SERPENTINITE CONCRETE AND IN MINERAL COTTON WOOL (THERMAL INSULATION IN THE PLUG), wt%

H O Na Mg Al Si S K Ca Fe Dens., g/cm³ Mineral

cotton wool (heat ins.)

- 39.9 2.04 5.04 5.47 19.76 - 2.28 15.8 3.96 0.15 Serpentinite

concrete [6] 1.27 48.5 - 15.5 1.45 18.6 0.036 - - 5.68 2.2 Tables 20 and 21 show the expertise estimated contents of the impurity elements in graphite and 3-grade steel are presented.

TABLE 20. THE CONTENTS OF BASIC AND IMPURITIES ELEMENTS IN GRAPHITE, wt%

Elements Concentrations Elements Concentrations Elements Concentrations

Li 5.8E-6 Co 5.9E-6 Sm 2.5E-5

C ∼ 100 Ni 1.8E-3 Eu 7.6E-7

N 4.7E-4 Nb 7.7E-4 Hg 6.7E-5

Cl 1.3E-3 Ag 9.0E-6 U 2.0E-4

Ca 1.4E-2 Sn 4.9E-5 - -

Fe 2.0E-3 Ba 1.1E-3 - -

TABLE 21. THE CONTENTS OF BASIC ELEMENTS AND IMPURITIES IN 3-GRADE STEEL, wt%

Elements Concentrations Elements Concentrations Elements Concentrations

Cr 0.25 Co* 0.021 Mo* 0.11E-2

Mn 0.25 Ni 0.25 Eu* 0.1E-4

Fe ∼ 99 Nb* 0.25E-2 - -

* Concentrations of elements assumed to be equal to those of Cr18Ni9 steel

5.6.2.3. Results of calculation of materials activity

Values of activity induced in the reactor materials and compositions of radionuclides produced under irradiation, vary in a wide range. For the purpose of convenience, analysis of calculation results is given for individual structural elements of the reactor and shielding, apart from analysis of activity values summed up for all radioactive reactor materials.

5.6.2.3.1. Activity of steel of in-vessel lateral shielding (band), shielding subassemblies and core diagrid

Tables 22 and 23 represent average specific activity values of nuclides contained in the steel of the band with the shielding subassemblies and in the steel of the core diagrid. The total activity values of steels depending on the duration of storage after irradiation are also presented in these tables.

TABLE 22. AVERAGE SPECIFIC ACTIVITIES OF NUCLIDES IN STEEL OF BAND AND SHIELDING SUBASSEMBLIES (Bq/kg) AND TOTAL ACTIVITY OF THE STEEL (Bq)

Storage duration, years Nuclide

composition 0 10 50 100 150 10 000

60Co 2.52E+11 6.77E+10 3.50E+8 4.86E+5 2.39E+3 -

63Ni 7.21E+10 6.73E+10 5.09E+10 3.60E+10 2.55E+10 -

59Ni 1.19E+9 1.19E+9 1.19E+9 1.19E+9 1.19E+9 1.09E+9

94Nb 1.26E+7 1.26E+7 1.26E+7 1.26E+7 1.26E+7 8.93E+6

55Fe 3.56E+12 2.74E+11 1.30E+7 - - - Sum, Bq 6.13E+17 1.41E+16 1.80E+15 1.28E+15 9.15E+14 3.76E+13 TABLE 23. AVERAGE SPECIFIC ACTIVITY OF NUCLIDES IN THE STEEL OF THE

CORE DIAGRID (Bq/kg) AND TOTAL ACTIVITY OF STEEL (Bq) Storage duration, years

Nuclide

composition 0 10 50 100 150 10 000

60Co 2.33E+11 6.27E+10 3.23E+8 4.50E+5 7.48E+2 -

63Ni 3.60E+10 3.33E+10 2.53E+10 1.79E+10 1.27E+10 -

59Ni 2.60E+8 2.60E+8 2.60E+8 2.60E+8 2.60E+8 2.38E+8

94Nb 2.13E+6 2.13E+6 2.13E+6 2.13E+6 2.13E+6 1.50E+6

55Fe 6.29E+11 4.84E+10 - - - - Sum, Bq 8.97E+16 2.42E+15 4.30E+14 3.04E+14 2.16E+14 4.01E+12 The analysis of results of calculation of induced activity of the bond steel, shielding subassemblies and core diagrid has shown that the value of activity of steel of these structures in the whole range of storage duration values exceeds the allowable level, at which no regulation of the radioactive material by the RSS (radiation safety standards) is required [25].

After several years of storage the activity is determined by ⁶⁰Co, ⁶³Ni, ⁵⁹Ni and ⁹⁴Nb. All these steel structures should be isolated from the human environment.

5.6.2.3.2. Activity of steel of thermal shielding, main and guard reactor vessels and supporting ring

Tables 24 and 25 present the average specific activities of nuclides in the steel of structures mentioned above. The total activities of steels depending on time of storage after irradiation are also given in these tables. The analysis of results of calculation has shown, that the induced activity of the significant portion of steel mass of the main reactor vessel and guard vessel (~ 80%) and all steel of the supporting ring exceeds the allowable level, at which the regulation of the radioactive material by the RSS is not required.

TABLE 24. AVERAGE SPECIFIC ACTIVITIES OF NUCLIDES IN STEEL OF THERMAL SHIELDING OF THE VESSEL (Bq/kg) AND TOTAL ACTIVITY OF STEEL (Bq)

Total, Bq/kg 3.31E+12 9.61E+10 2.03E+10 1.43E+10 1.02E+10 2.21E+8 Total, Bq 1.78E+17 5.17E+15 1.09E+15 7.69E+14 5.47E+14 1.19E+13 TABLE 25. AVERAGE SPECIFIC ACTIVITIES OF NUCLIDES IN STEEL OF THE

MAIN REACTOR VESSEL, GUARD VESSEL AND SUPPORT RING (Bq/kg) AND TOTAL ACTIVITY OF STEEL (Bq)

Storage duration, years Nuclide

composition 0 10 50 100 150 10 000

60Co 2.32E+10 6.39E+9 3.31E+7 4.59E+4 6.30E+1 -

63Ni 3.59E+9 3.13E+9 2.37E+9 1.67E+9 1.19E+9 -

59Ni 3.52E+7 3.52E+7 3.52E+7 3.52E+7 3.52E+7 3.20E+7

94Nb 2.13E+5 2.13E+5 2.13E+5 2.13E+5 2.13E+5 1.52E+5

55Fe 5.12E+10 3.93E+9 1.37E+5 - - -

Total, Bq/kg 4.35E+11 1.35E+10 2.44E+9 1.71E+9 1.22E+9 3.21E+7 Total, Bq 4.42E+16 1.37E+15 2.47E+14 1.74E+14 1.23E+14 3.26E+12

The analysis of results of calculation has shown, that the induced activity of the significant portion of steel mass of the main reactor vessel and guard vessel (~ 80%) and all steel of the supporting ring exceeds the allowable level, at which the regulation of the radioactive material by the RSS is not required. After several years of storage the activity is determined by ⁶⁰C, ⁶³Ni, ⁵⁹Ni and ⁹⁴Nbisotopes.

5.6.2.3.3. Activity of steel of rotating plugs, plugs shielding structure and central column The central column penetrates through the rotating plugs and shielding structure the activity at the bottom part making the main contribution to the total activity of steel of the reactor structures. Some steel sheets located in the bottom part of rotating plugs have the activity higher than the permissible limits of RSS.

The induced activity of graphite and thermal insulation is lower than the permitted value, i.e.

no RSS regulation of radioactive material is required. Fifty years later the activity of rotating plugs will be lower than the permitted value, i.e. no RSS regulation of radioactive material is required. A hundred years later ~ 80% of steel shielding of rotating plugs will have an activity level, indicated above. Average specific activity values of nuclides in the steel of structures indicated above are presented in Table 26. In this Table the total activity values of steel depending on time of storage after irradiation are also presented. Value of the activity after several years of storage is determined by ⁶⁰C, ⁶³Ni, ⁵⁹Ni and ⁹⁴Nbisotopes.

TABLE 26. AVERAGE SPECIFIC ACTIVITIES OF NUCLIDES IN THE STEEL OF THE ROTATING PLUGS, SHIELDING STRUCTURE AND CENTRAL COLUMN (Bq/kg), AND TOTAL ACTIVITY OF THE STEEL (Bq)

Storage duration, years

Total, Bq/kg 3.78E+10 1.44E+9 4.74E+8 3.35E+8 2.39E+8 5.64E+6 Total, Bq 4.73E+15 1.80E+14 5.92E+13 4.19E+13 2.99E+13 7.06E+11 5.6.2.3.4. Activity of steel shielding, iron ore filling and steel lining of concrete

The activity of all shielding steel, located inside the iron ore filling, will exceed the level, which is critical from the standpoint of the radioactive material regulation by the RSS.

A hundred years from then only steel of the concrete lining and that of the back wall of the reinforcing cage of the iron ore filling will have activity below the RSS limit value. Average specific activities of the nuclides in the steel of the structures mentioned above are given in Table 27. In this table the total activities of steel depending on time of storage after irradiation are also presented. Value of activity of 3 grade steel after several years of storage is determined by ⁶⁰Co, ⁶³Ni, ⁵⁹Ni, ⁹⁴Nb, ¹⁵²Eu, and ¹⁵⁴Euisotopes.

TABLE 27. AVERAGE SPECIFIC ACTIVITIES OF THE NUCLIDES IN THE STEEL OF SHIELDING, IRON ORE FILLING AND CONCRETE LINING (Bq/kg), AND TOTAL ACTIVITY OF STEEL (Bq)

Storage duration, years Nuclide

composition 0 10 50 100 150 10 000

60Co 7.65E+9 2.05E+9 1.06E+7 1.46E+4 2.04E+1 -

63Ni 3.55E+7 3.3E+7 2.50E+7 1.77E+7 1.25E+7 -

59Ni 3.88E+5 3.88E+5 3.88E+5 3.88E+5 3.88E+5 3.53E+5

94Nb 8.99E+4 8.99E+4 8.99E+4 8.99E+4 8.99E+4 6.39E+4

152Eu 7.69E+6 4.58E+6 5.71E+5 4.24E+4 3.15E+3 -

Total, Bq/kg 6.33E+10 4.26E+9 3.66E+7 1.81E+7 1.29E+7 4.17E+5 Total, Bq 2.18E+16 1.47E+15 1.26E+13 6.23E+12 4.44E+12 1.44E+11 5.6.2.3.5. Activity of the iron ore filling

Iron ore filling serves as concrete shielding against radiation. After 50 years of storage, about 65% of the iron ore filling will have activity above limiting RSS values. A hundred years after the reactor shutdown the portion of the filling having such activity would not exceed 15% of its initial volume. Average specific activities of the nuclides in iron ore filling are shown in Table 28 In this table the total activity values depending on the time of storage after irradiation are also presented. Value of activity of the iron ore filling after several years of storage is determined by ⁶⁰Co, ¹⁵²Eu, ¹⁵⁴Eu, ⁹⁴Nb, ¹⁵¹Sm, ¹⁴C and ⁴¹Ca isotopes [27].

TABLE 28. AVERAGE SPECIFIC ACTIVITIES OF THE NUCLIDES IN IRON ORE FILLING (BQ/KG) AND TOTAL ACTIVITY OF IRON ORE FILLING (BQ)

Storage duration, years

Total, Bq/kg 3.29E+8 2.64+E7 7.88+E4 1.62+E4 1.34+E4 2.73E+3 Total, Bq 4.28E+14 3.44+E13 1.03+E11 2.12+E10 1.74+E10 3.55E+9

5.6.2.3.6. Activity of the concrete

Just after the reactor shutdown the activity of about 381 m³ of concrete exceeds the RSS limit values. After 100 years of storage the amount of radioactive concrete decreases down to ~ 8%

of the initial value. After 120 years of storage the whole amount of concrete will have activity not exceeding the limiting RSS values. The average specific activities of nuclides in concrete are given in Table 29. In this Table the total activities depending on time of storage after irradiation are also presented. The concrete activity value after several years of storage is determined by Eu¹⁵², Eu¹⁵⁴, Co⁶⁰, H³, Ni⁶³, C¹⁴ and Ca⁴¹ isotopes.

TABLE 29. AVERAGE SPECIFIC ACTIVITIES OF THE NUCLIDES IN CONCRETE (BQ/KG) AND TOTAL ACTIVITY OF THE CONCRETE (BQ)

Storage duration, years Nuclide

composition 0 10 50 100 150 10 000

60Co 3.55E+4 9.53E+3 4.92E+1 - -

-55Fe 4.04E+5 3.09E+4 - - -

-3H 2.11E+5 9.91E+4 1.27E+4 7.64E+2 4.61E+1

-152Eu 3.52E+4 2.29E+4 2.86E+3 2.12E+2 1.57E+1

-154Eu 3.55E+3 1.75E+3 6.87E+1 - -

-41Ca 1.69E+3 1.69E+3 1.69E+3 1.69E+3 1.69E+3 1.58E+3

63Ni 2.87E+2 2.68E+2 2.03E+2 1.43E+2 1.02E+2

-14C 3.11E+2 3.10E+2 3.09E+2 3.06E+2 3.05E+2 9.30E+1

134Cs 1.79E+4 1.30E+2 - - -

-Total, Bq/kg 7.90E+5 1.65E+5 1.78E+4 3.12E+3 2.20E+3 1.69E+3 Total, Bq 6.93E+1 1.45E+11 1.56E+10 2.74E+9 1.93E+9 1.48E+9 5.6.2.3.7. Activity of the graphite

Graphite placed in the plugs, has activity below the limiting RSS value. About 50% of graphite located in the ionization chambers unit (ICU), also has activity below RSS limit.

After ~ 150 years of storage all graphite will have activity below RSS limit values. Average specific activities of the nuclides in graphite of ICU are presented in Table 30.

TABLE 30. AVERAGE SPECIFIC ACTIVITIES OF THE NUCLIDES IN GRAPHITE OF THE ICU (BQ/KG) AND TOTAL ACTIVITY OF GRAPHITE (BQ)

Storage duration, years

94Nb 4.38E+2 4.38E+2 4.38E+2 4.38E+2 4.38E+2 3.12E+2

63Ni 6.85E+4 6.37E+4 4.84E+4 3.42E+4 2.42E+4 9.30E+1

14C 1.05E+4 1.05E+4 1.04E+4 1.03E+4 1.03E+4 3.13E+3

59Ni 6.27E+2 6.27E+2 6.27E+2 6.27E+2 6.27E+2 5.71E+2

41Ca 6.32E+2 6.32E+2 6.32E+2 6.32E+2 6.32E+2 5.91E+2

134Cs 3.67E+6 1.26E+5 - - - -

55Fe 1.04E+5 7.78E+3 - - - -

Total, Bq/kg 4.36E+6 4.29E+5 8.41E+4 4.78E+4 3.63E+4 4.61E+3 Total, Bq 8.99E+10 8.85E+9 1.73E+9 9.85E+8 7.49E+8 9.49E+7

The total activities depending on time of storage after irradiation are also presented in this table. Graphite activity value after several years of storage is determined by¹⁵²Eu, ¹⁵⁴Eu, ⁶⁰Co,

94Nb, ⁶³Ni, ⁵⁹Ni, ¹⁴C and ⁴¹Ca isotopes. Total activities of the nuclides for all reactor materials are presented in the Table 31.

TABLE 31. TOTAL ACTIVITY OF THE NUCLIDES OF THE REACTOR

MATERIALS (Bq)

Storage duration, years Nuclide

composition 0 10 50 100 150 10 000

60Co 2.57E+16 6.89E+15 3.57E+13 4.96E+10 1.32E+8 -

63Ni 5.04E+15 4.68E+15 3.55E+15 2.50E+15 1.78E+15 -

59Ni 6.26E+13 6.26E+13 6.26E+13 6.26E+13 6.26E+13 5.27E+13

94Nb 5.90E+11 5.90E+11 5.90E+11 5.90E+11 5.90E+11 4.18E+11

152Eu 2.62E+12 1.58E+12 1.97E+11 1.46E+9 1.08E+8 -

154Eu 3.23E+12 1.44E+12 5.75E+10 1.02E+9 1.81E+7 -

55Fe 1.69E+17 1.30E+16 5.28E+11 - - -

56Mn 4.21E+17 - - -

58Co 7.96E+16 - - -

58mCo 3.94E+16 - - -

51Cr 8.80E+16 - - -

59Fe 3.12E+16 - - -

60mCo 1.49E+16 - - -

55Cr 3.31E+15 - - -

52V 2.72E+15 - - -

65Ni 2.32E+15 - - -

95Nb 2.75E+13 - - -

64Cu 1.01E+13 - - -

54Mn 8.70E+12 - - -

Total, Bq 8.82E+17 2.47E+16 3.65E+15 2.56E+15 1.84E+15 5.77E+13 5.6.2.3.8. Total activity of reactor materials and total volumes of radioactive waste

Table 32 gives total volumes and weights of the radioactive waste, formed as a result of neutron irradiation of reactor materials, having induced activity above RSS limit values. The main contribution to the amount of radioactive waste is made by the iron-ore filling and reactor structures made of stainless steel and 3 grade carbon steel. During 100 years of storage the amount of the radioactive waste having induced activity higher than the RSS limit value, will be decreased from ~ 860 m³ (~ 2.900 t) down to ~ 150 m³ (~ 680 t). After storage during 120-150 years the whole mass of concrete and graphite will have activity, at which these radioactive materials are free from the RSS regulation. The radiological conditions in the reactor vessel cavity will be determined by the products of activation of Co, Eu and Nb, contained as the impurities in structural and shielding materials of the reactor.

TABLE 32. VOLUMES AND WEIGHTS OF THE REACTOR MATERIALS, HAVING INDUCED ACTIVITY EXCEEDING RSS LIMITS

Storage duration, years

0 50 100 150 10 000

Stainless m³ 42.5 35.6 25.6 25.6 18.8 Steel Cr18Ni9 tons 331.5 277.7 199.7 199.7 146.6 3-grade steel m³ 44.2 44.2 30.1 30.1 30.1

tons 344.8 344.8 234.8 234.8 234.8 Iron-ore m³ 376.3 240.3 56.0 56.0 42.0 Filling tons 1303.0 720.9 168.0 168.0 126.0 Graphite m³ 12.9 6.0 3.0 0.0 0.0

tons 20.6 9.6 4.8 0.0 0.0 Concrete m³ 381.0 114.3 30.0 0.0 0.0

tons 876.3 263.0 69.0 0.0 0.0

Total for the m³ 856.9 440.4 144.7 111.7 90.9 whole reactor tons 2876.2 1616.0 676.3 602.5 507.4 Maximum dose rates of gamma-radiation on the surface of activated structures of the reactor are presented in Table 33.

TABLE 33. DOSE RATES OF GAMMA-RADIATION ON THE SURFACE OF ACTIVATED STRUCTURES, μSV/S

It was assumed that dose rate of gamma-radiation was formed only due to activation of the material of the above indicated structure. The possible contribution to the dose rate of gamma-radiation made by the adjacent structures is not shown. When estimating the radiological conditions at the reactor top structures, it is necessary to take into account the increased amount of radioactive products deposits on the reactor structural elements having been in contact with the primary cover gas. The total values of activity of the materials after several years of storage are determine by ⁶⁰Co, ⁶³Ni, ⁵⁹Ni, ¹⁵²Eu, ¹⁵⁴Eu, ⁹⁴Nb and ⁵⁵Feisotopes.

Dose rate of gamma-radiation on the surface of structural elements only due to this contamination can reach ∼ 2 μSv/s value, which is comparable to the effect of induced activity of structural materials located above the “floating” shielding. The analysis of data presented in Tables 29-33 shows that in case of short storage duration it is impossible to carry out dismantling of the reactor installation without application of robotics. The complete dismantling of the reactor installation without application of robotics can be carried out only after 100 years of storage.

5.6.2.4. Protection techniques

It follows from the analysis of the radiological characteristics of the equipment and technological media of the shutdown reactor, that in order to assure radiological safety of the personnel during the installation decommissioning it is necessary to meet the following two requirements:

— Maximum use of the operating systems of the reactor installation;

—

Development of additional systems, allowing safe implementation of dismantling work

.

The ratio between these directions depends on the chosen scenario of the reactor installation decommissioning. Below are given some specific recommendations, directed to assurance of the radiological safety of the BN-350 reactor decommissioning.

5.6.2.4.1. Use of operating systems of the reactor

On the stage of the BN-350 reactor decommissioning, the following systems required for the radiological safety assurance, should be used to the maximum extent:

— System of the radiological control;

— System of fuel element cladding failure detection and location;

— System of special ventilation;

— Systems of sodium fire fighting;

— System of special drains;

— Systems of fuel transport;

— Systems of reprocessing and disposal of the liquid and solid radioactive waste;

— “Hot” chambers for works with high activity products;

— Protection barriers and sanitary barriers;

— System of monitoring of the radiological situation on the NPP site.

The majority of these systems should keep their serviceability, practically at all stages of decommissioning. After removal of spent fuel and liquid metal coolant of the primary and secondary circuits and spent fuel storage cooling circuit from the BN-350 NPP, some systems, assuring radiological safety or their elements, such as the system of fuel element

cladding failure detection, the monitoring system of inert radioactive gases and the system of sodium fire fighting can be dismantled.

5.6.2.4.2. Additional systems for assurance of radiological safety for putting in prolonged storage and dismantling of the reactor components

In order to assure radiological safety on the stage of putting in prolonged storage and dismantling of the reactor installation, development of new systems or modification of some existing systems is required:

— Creation of the additional sanitary barriers to prevent spreading of the radioactive contamination (for example, sanitary lock in the area of cells of the primary circuit loops, reactor dome, etc.;

— Replacement of closed ventilation systems of the primary circuit cells and reactor vessel cavity by a special ventilation systems having aerosol filters, gas being released to the atmosphere through the stack (it is necessary to exclude the ingress to the attended rooms of ozone and nitrogen oxides, caused by gamma-radiation of the high activity structural elements of the reactor). On the stage of putting reactor cavity structures in prolonged storage it is expedient to have ventilation system based on the natural draft provided by the stack;

— Creation of local systems of collecting, sorting, reprocessing, storage (putting in prolonged storage) and disposal of the solid radioactive waste, including high activity items (control rods and their sleeves, photo neutron sources, cesium absorbers, etc.);

— Development of the remote tools, robotics and devices for realization of dismantling works, including those carried out under water;

— Creation of an additional network of stationary instruments, intended for the control of superficial contamination of personal protection means, equipment, etc. by the radioactive products;

— Organization of system for analysis of radionuclide composition of material samples from the dismantled equipment (including α and β control). All these activities should be taken into account in the project of installation decommissioning and, as a part of it, in the report on the nuclear and radiological safety on the stage of the BN-350 reactor decommissioning.

REFERENCES

[1] Safety Requirements For transportation of Radioactive Materials, PBТ RV-73, Moscow, Russian Federation (1973).

[2] Guidebook on Extended Sources Radiation Protection, Atomizdat, Moscow, Russian Federation (1965).

[3] Guidebook on Ionizing Radiation Protection, Atomizdat, Moscow, Russian Federation (1966).

[4] Radiation Safety Standards (NRB-96), Goskomsanepidnadzor of Russian Federation, Moscow, Russian Federation (1996).

[5] LOKSHIN, E.P., On Production of High Purity Materials by Vacuum Gauge Methods,

[5] LOKSHIN, E.P., On Production of High Purity Materials by Vacuum Gauge Methods,

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