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The main sources of radioactive substances generation at the plant are the uranium-235 fission products at the core fuel neutron irradiation, neutron activation of structural materials, primary coolant additives and air in the at-reactor space.

The restriction of the radioactive gases and aerosols spread over the plant and their discharge into the environment is provided due to consistent implementation of the in depth protection principle based on application of the barrier system. The restricting barriers are as follows:

-              fuel matrix;

-              fuel element cladding;

-              primary coolant circuit;

-              containment enclosing the primary coolant circuit.

Additionally, the process circuits and equipment containing radioactive products restrict the uncontrolled spread of radioactive substances over the plant and beyond it.

Under normal operation conditions, all the barriers and their protection facilities are in the operating state. If one of the specified barriers or its protection facilities is detected to be non-operable, according to the safety operation conditions, the NPP power operation is stopped (Chapter 16 of PSAR).

For all the NPP service conditions, the safety report establishes the service limits and the safe operation limits describing conditions of the systems (elements) and the entire NPP (reactor power, pressure in the first circuit, coolant temperature during operation of the reactor with different number of working loops, activity of the coolants of the primary and secondary circuits, etc.). Introduction of such limits ensures the control of the barriers integrity and, first of all, the integrity of the fuel element claddings, thus preventing from the significant release of fission products from the fuel to the primary coolant and further into plant premises with the basic process equipment.

The assessment of the plant radioactive waste generation during operation of the power unit under normal conditions was performed in compliance with the Terms of Reference for the AES-2006 for the fuel element leakiness level corresponding with the service limit established in
НП-082-07. The Sections 10.2.3 and 10.3.3 of PSAR analyze the quantitative nuclide content of the gas and aerosol emission and discharge of the fission and activation products from the plant during continuous operation of the unit with the fuel which has gained the safety level in terms of the fuel element leakiness according to НП-082-07.

Only a little part of all radioactive fission and activation products generated during power operation of the unit can be released to the plant waste or to the environment which is even less likely. Much more than 99.9 % of the fission products accumulated in the fuel remain in the used fuel elements. Content of the fission fragments in the reactor EP during power operation of the unit at the end of the stationary fuel load V-491 RP is performed within the project of LNPP-2 [26] using the code “Radionuclide” [27] which certification is scheduled for 2009–2010. The calculation results are presented in Section 15.7.5 of PSAR and lay at the following level, TBq: IRG – 4.4×107, iodines – 3.6×107, cesium-134, 137 – 1.1×106.


The rates of release of various nuclide groups into the process circuits of the unit within the calculation models accepted in the project while in service amount to [1], TBq/year: IRG – 8.6×104, iodines – 3.5×104, cesium-134, 137 – 3.5.

The source of argon-41 generation at the plant is the neutron activation of the stationary argon-40 present in the atmosphere of the hermetic containment rooms next to the reactor vessel. The discharge of argon-41 from the power units VVER and PWR is estimated [2, 3] at 0.4 TBq/year·GWе.

The most important source of carbon-14 inthe primary coolant is the thermal neutron activation of oxygen-17 included in the natural mixture. In addition, the carbon-14 generation is considered at the neutron activation of nitrogen-14.

The rate of the global nuclides generation at the unit with the V-491 RP is estimated at [1, 4], GBq/year·GWе:

-              carbon-14        3.0×102

-              krypton-85      3.6×102

-              tritium             1.3×104

-              iodine-129      1.0×10–3

10.1.2 PRIMARY COOLANT ACTIVITY Mathematical calculation models

The sources of contamination of the primary coolant with the fission products during unit operation at the rated power are the following:

-              defective fuel elements with the gas leakiness and considerable damages;

-              surface contamination of external claddings of the fuel elements;

-              contamination of structural materials.

During the initial operating period of the reactor (without the fuel element manufacturing defects), the contamination level of the coolant with the fission products is defined by the discharge into the circuit of the uranium-235 fission fragments present on the external surfaces of the fuel elements as contaminations occurring when manufacturing, and in the natural uranium impurity in the zirconium shells, due to the kinetic energy.

Under normal operation conditions of the reactor, the tightness failure of the fuel element claddings may occur due to various corrosion-fatigue processes. With that, at first microfissures and then macroscopic defects occur in the claddings accompanied by growth of transfer of the fission products from the fuel elements into the primary coolant.

The diffusion model is accepted as the justifying model of contamination of the coolant with the fission products. This model describes the migration and leakage of the radioactive fission products from the uranium dioxide and from under the fuel element cladding under different irradiation conditions [5].

Within the given model, there are used the following mechanisms of the fission products discharge from the uranium dioxide into the fuel element gas gap:

-              discharge of the fragments due to kinetic energy at the uranium-235 fission (discharge due to the kickback);

-              discharge due to “knock-out” of the fission products from the fuel surface by flying out fission fragments (knock-out effect);

-              discharge of the fission products from the structural changes zone (high-temperature area of the fuel).


During operation of the reactor having non-tight fuel elements in the core, the activity growth (emission) of some fission products in the primary coolant is frequently observed after reduction or increase of the reactor power. This activity emission (spike effect) conditioned by the additional discharge of gaseous and volatile fission products accumulated in the gas gap of the non-tight fuel elements is taken into account in transient modes of the power unit.  

The functional diagram of the fission products migration for the fuel rods’ fuel elements in zirconium alloy claddings with the nuclear fuel on the base of sintered uranium dioxide is presented in Figure 10.1.1.

It is assumed that the fuel burn up in the VVER-1200 reactor will achieve 70 MW·days/kg of uranium [Item of Terms of Reference for RP VVER-1200]. In this connection, RRC KI had performed adaptation of the existing calculation method and program RELWWER-2.0 [5, 6, 17] intended for calculation of activity of the fission products in the process media and under claddings of the tight and non-tight fuel elements of the VVER reactors certified by GAN RF for burn ups of up to 43 MW·days/kg of uranium.

The authors have proven that the previously used design scheme of migration of the fission products under claddings of the non-tight fuel elements can be conceptually preserved. However, the final adaptation of the method will be possible only after its verification according to the results of the reactor experiments on irradiation of fuel elements with artificial defects within the range of the practically achieved fuel burn up [1]. It was offered to use the empirical correlation obtained from the analysis of the literature data on gas emission in the VVER reactors fuel elements and describing increase of the gas emission from the VVER reactors fuel with the burn ups exceeding 43 MW·day/kg of uranium. Such an approach is conservative with regard to the radioactive fission products. According to the empirical correlation, discharge of the radioactive fission products from the fuel increases by 1 % per each 7 MW·day/kg of uranium starting from burn up of 43 MW·day/kg of uranium. The same correlation was also used for calculation of accumulation of the fission products under the fuel elements claddings outside the fuel.

The RELWWER code was used in the product for calculation of discharge of the fission products into the primary circuit heat carried with the power operation of RP and considering the “spike effect” conditioned by the additional discharge of gaseous and volatile fission products accumulated in the gas gap of the non-tight fuel elements in the transient modes.

Calculations of the fragment activity of the primary coolant under conditions of NO RP at the 100 % power take into account the direct discharge of the fission products into the coolant from the fuel (0.02 % of defect fuel elements having direct contact with the coolant) and the fission products discharge into the coolant from the axial gap of the fuel elements (0.2 % of defect fuel elements having gas leakiness).

The balance equations solutions implemented in the RELWWER [6] calculation code describe accumulation of the fission product with the coefficients specifying the rate of their discharge from the fuel elements/Gd fuel elements depending on power distribution at the end of the stationary fuel load with regard to power density irregularity by the fuel height, temperature distribution by the fuel height for the fuel elements of different power and the digital printout of the FA burn up distribution.

The activated corrosion products in the primary coolant and on the surfaces are generated as a result of the following: 

-              activation and corrosion of materials of various components of inner units and core of the reactor;  


-              corrosion of materials outside the reactor core which corroded parts are activated when passing through the core with the primary coolant.

Additionally, there takes place activation of chemical impurities dissolved in the coolant (40К, 24Na and others).

Calculations of the corrosion products in the primary circuit equipment elements are performed for the project [8, 20] according to the program COTRAN-М [7] certified by GAN RF for such class of the tasks. AR provides the evaluation of specific activity of the corrosion products in the coolant, activity in the filters of the system of low-temperature cleanup of the KBE primary coolant and the surface corrosion activity in different areas of the primary circuit equipment in the stationary operating mode of the unit:

-              hot pipeline;

-              hot steam generator collector;

-              steam generator pipe still;

-              cold pipeline;

-              cold steam generator collector;

-              main circulation pump volute;

-              protective tube unit;

-              reactor pressure vessel;

-              reactor cover.

The COTRAN-М [7] program algorithm is implemented on the base of the developed physical-chemical model considering behavior in the circuit of the soluble and dispersed phases of the corrosion products under water-chemistry conditions in the primary circuit. The main mechanisms of the model used in the program:

-              generation of nonradioactive corrosion products in the primary circuit as a result of corrosion of the circuit structural materials;

-              generation of particles in the coolant flow core in the circuit points where the concentration of saturation with the soluble corrosion products exceeds the solubility product value for the given thermodynamic conditions;

-              generation of the external oxide layer of the corrosion products due to soluble phase crystallization processes, particles deposition, and deposits layer dissolution and erosion;

-              presence of the ion exchange between the coolant, particles, and the deposits layer.

The calculations were performed for the conditions of the optimized WC maintenance of the primary circuit [9.2.12 PSAR]. The diagram of the corrosion products mass transfer processes implemented in the program COTRAN-М is presented in Figure 10.1.2.

Considering the fact that the collective dose for the personnel when servicing the NPP with VVER is substantially defined by the long-lived radionuclides 60Со, 58Со, 54Mn and 59Fe of activation origin, these nuclides were selected as reference when analyzing the calculated and experimental data. 

- 0 +    дата: 27 марта 2015

   Загружено переводчиком: Медведев Алексей Геннадиевич Биржа переводов 01
   Язык оригинала: английский    Источник: Транс-Линк