AKKUYU NUCLEAR POWER PLANT

       Agreement between the Government of the Republic of Turkey and the Government of the Russian Federation on Cooperation on the Establishment and Operation of a Nuclear Power Plant (NPP) Intergovernmental Agreement (IGA) was signed on 16.08.2010 in the Republic of Turkey. The Akkuyu Nuclear Power Plant, planned to be built in the Akkuyu district of the Büyükeceli town of Gülnar province in the south of Turkey, will be composed of 4-Unit Russian type VVER-1200 reactor and each unit will have a power of 1200 MWe. The display on Akkuyu area's map of Turkey is presented below.

Akkuyu Area

Figure 1 Akkuyu Area on Turkey map


      The steps and completion dates for the Akkuyu Nuclear Power Plant licensed by the Turkish Atomic Energy Authority (Türkiye Atom Enerji Kurumu, TAEK) are given below in Table 1.


Table 1 Important dates for Akkuyu NPP [1]

Step Completion Date
Recognition as a founder 28.02.2011
Location License 01.07.1976
Update Location Report 06.12.2013
Field Parameter,Confirmation,09.02.2017 09.02.2017


     The project company applied to TAEA on 03.03.2017 for construction licence of Akkuyu Nuclear Power Plant and is currently under review.


akkuyu npp

Figure 2 Akkuyu Nuclear Power Plant Simulation


     Some of the characteristics of Akkuyu NPP VVER-1200 reactor are given below



Table 2 Akkuyu Nuclear Power Plant Properties

AKKUYU NUCLEAR POWER PLANT PROJECT
Approximate Cost 20 Billion US Dollars
Reactor Type VVER-1200
Number of Units 4 Units
Total Installed,Power 4800 MWe
Operating Life 60 year
Founder Rosatom
Founding Country Russia
Region Mersin Province, Gülnar District,
Büyükeceli Town
The Stage Construction Licensing


Table 3 VVER-1200 General Features

Full Name VVER-1200
Reactor Type Pressurized Water Reactor (PWR)
Coolant H2O
Moderator H2O
Neutron Spectrum Thermal Neutron
Thermal Capacity 3200.00 MWth
Electricity Capacity 1170.00MWe



VVER 1200 Reactor General Features


       The security systems of the VVER-1200 design, which will be installed in Turkey, are often composed of systems that do not require an electrical power supply in the event of an accident. Some of the safety systems in the design include: control rods, active and passive emergency core cooling systems, emergency boric acid injection system, steam generator cooling system, passive residual heat extraction system [2].

       To ensure the safety of Nuclear Reactors, the possible spread of ionizing radiation and radioactive materials must be prevented. In order to prevent the spread, it is necessary to apply the Depth Defense Code in a comprehensive manner to the system barriers.

    VVER-1200 design Russian Regulatory Documents (RRC) in cooperation with the scientific supervisor of the "Kurchatov Institute (Moscow)" in accordance with the requirements of the RRC, European Utilities Requirements (EUR) and International Atomic Energy Agency General Designer "Atomenergoproekt (St. Petersburg)" was developed by the organization. This design was made using the ISO 9001-2000 Quality Assurance International Standard.

      The safety concept of VVER-1200 takes into consideration the safety requirements and current world concepts which are constantly tightened in the field of Nuclear Power Plants' safety. These requirements include the economic efficiency of Nuclear Power Plants.

        Specific characteristics of Nuclear Power Plants are given below:

• The service life of the equipment cannot be changed is 60 years,

• Development of horizontal type steam generators with large water inventory improves the natural circulation of the primary side compared to vertical type steam generators,

• Emergency Core Cooling System (ECCS) applications are based on active and passive system,

• Double enveloped concrete enclosure is applied,

• There is Reliability Enhanced Measurement & Control System (I&C), which is the nervous system of the nuclear plants and determines the health of the whole plant and the need for assistance. This system consist of computers and microprocessors,

• The reactor vessel has been enlarged and the shell of the reactor vessel armature has been produced independently from the longitudinal welds,

• Reactor baffle below the reactor vessel independent of the injector holes and cuts,

• There are passive component, insulation, constraints and evacuations appliances,


     The Reactor coolant system removes the heat from the reactor core by coolant circulation in a closed circuit and provides heat transfer to secondary side. The reactor coolant system comprises a reactor, a pressurizer and four circulation loops, each one comprising a steam generator, reactor coolant pump set and main coolant pipelines that procide the loop equipment-to-reactor connection. A steam generator links the primary and the secondary sides. The steam generator headers and heat-exchange tubes are a barrier between the primary coolant and the working medium of the secondary side and prevent the radioactive substance penetration out of the primary tp the secondary side.

    The pressurizing systemis a constituent part of the primary side and performans the functions of primary-side pressurization, pressure maintenance under the steady-state conditions, pressure control at heatup and cooldown, pressure deviation limitation under transient and accidents.

     The main designand thermal-hydraulic performance of the primary side with reactor plant nominal operation is given in Table 4.



Table 4 Thermal-Hydraulic Performance of Primary Side

Parameter Value
Reactor nominal thermal power, MW 3200
Coolant inventory in reactor coolant system
(PRZ system not considered), m3
290
Coolant inventory in PRZ at nominal power
operation, m3
55
Primary pressure at the core outlet,,absolute, MPa 16.2
Coolant temperature at reactor inlet, °C 298.2
Coolant temperature at reactor outlet, °C 328.9
Coolant flowrate through reactor, m3/h 86000
Primary-side design parameters:
      -   Gauge pressure, MPa;
      -    Temperature, °C;

17.64
350
Pressure of the primary-side hydraulic tests, MPa:
      -    For tightness;
      -    For strengths;

17.64
24.5

Reactor Core and Fuel Design


       The reactor core contains 163 fuel assemblies (FA). The FAs are intended dor heat generation and its transfer from the fuel rod surface to coolant during the design service life without exceeding the permissible design limits of fuel rod damage. The FAs are 4570 mm high (nominal value). When the reactor is in the hot state the height of the power-generating part of the fuel rod is 3750 mm. Each FA contains 312 fuel rods. The FA sketlon is assembled of 18 guide channelsi 13 spacer grids welded to them, an instrumentation channel and a support grid. The fuel rod cladding is a zirconium alloy tube. Sintered UO2 pellets with a 5% (4.95 ± 0.05) maximum enrichment are stacked inside the cladding. The average linear heat rate of a fuel rod is 168.8 W/cm.

      According to the cartogramup to 121 rod cluster control assemblies (RCCAs) are placed inside the core. They are intended for quick chain reaction suppression, maintaining power at assigned level and its level-to-level transition, axial power field leveling, xenon oscillation suppression. Pitch electromagnet drives with pitch position indicators are used for RCCA drive mechanisms. The drives are installed on the reactor top head. The maximum effective time of FA operation between refuelings for a 12-month fuel cycle is 8400 effective hours. The average burnup of unloaded fuel is up to 60 MWD/kg U. Annually 42 fresh FAs are loaded into the core for the basic fuel cycle.



REACTOR


      The reactoris a vertical pressure vessel (a vessel and a top head) that houses the internals (protective tube unit, core barrel, and core baffle), the core, control rods and in-core instrumentation sensors. The main joint of the vessel-to-top head that is structurally integrated into the top unit is sealed with the main joint studs. The reactor is positioned in the concrete cavity with a biological and thermal shielding and a cooling system. The reactor and its components are shown below;

Reactor Drawing

Figure 3 Reactor


STEAM GENERATOR


     Steam generator supports comprises the following components: steam generator, steam header, supports shock absorbers, one- and two-chamber surge tanks, embedded components for supports and shock absorbers.

     The steam generator itself is a single-vessel heat exchange apparatus of horizontal type with submerged heat-transfer surface and comprises the following components:

• A vessel with different-purpose nozzles;

• A heat-exchange bundle with fastener and spacer components;

• Primary coolant collectors;

• Feedwater supply and distribution systems;

• Emergency feedwater supply and distribution systems;

• Distribution perforated plate;

• Submerged perforated plate;

• Chemicals feeder.



REACTOR COOLANT PUMP


    The reactor coolant pump is designed to create the primary coolant circulation in the reactor plant. The RCP set hasan additional function of providing coolant circulation at the coastdown under any loss-of power accidents, which allows a smooth passing to the natural circulation mode. The reactor coolant pump set has the parameters are given in the table below.

Table 5 Reactor Coolant Pump Set

Parameter Value
Nominal supply, m3/h 21500
Pressure head,,MPa 0.588
Coolant,temperature, °C 298.2
Design,temperature, °C 350
Design pressure, MPa 17.64
Rotation rate,,rev/min 1000/750
Current frequency,,Hz 50
Sealing water,leakages, m3/h 1.2


PRESSURIZER


     The pressurizer is a vertically positioned pressurized cylindrical vessel with elliptic bottoms installed on a cylindrical support. The pressurizer performance and main dimensions are given in Table 6.


Table 6 The Pressurizer Performance and Main Dimensions

Parametre Değer
Nominal pressure of the steady-state
conditions, MPa
16.1
Nominal temperature of the steady-state
conditions, °C
347.9
Design pressure, MPa 17.64
Design wall temperature, °C 350
İnternal diameter, mm 3000
External diameter, mm 3330
Capacity (full volume), m3 79
Water level at nominal power operation, m3 55
Water level under steady-state conditions
(level gauge reading), m
8.17± 0.15

Working,medium
Steam and water
Nitrogen under
heatup/cooldown conditions
Total power of PRZ TEHs, kW 2520


SAFETY CONCEPT, MAIN DESIGN PRINCIPLES AND METHODS OF LICENCING


     The design was developed on the basis of the requirements of the up-to-date safety rules and standards in the nuclear power engineering of Russia considering the Safety Guides and other recommendations issued by the International Atomic Energy Agency and the requirements of the European Utilities for the NPPs with LWR.

    The determination of the safety system configuration in the present design is based on the application of the following principles:

• Single failure principle,

• Redundancy principle,

• Diversity principle,

• Principle of physical separation,

• Protection against the operator’s errors,

• RP inherent safety principle.



SAFETY SYSTEMS and ACTIVE-PASSIVE SYSTEMS


     The strategy of coping with the design basis accident is based on using both the active and passive systems. The strategy of coping with the beyond design basis accident is based on using preferably the passive safety systems.

       The following active and passive systems are implemented in VVER-1200;


Table 7 VVER-1200 Design Active-Passive Systems

VVER-1200 Design Active System VVER-1200 Design Passive System
High pressure emergency spray system Emergency core cooling system, passive
part
Low pressure emergency spray system System of passive heat removal from
containment
Emergency gas removal system System of passive heat removal via steam ,
generators
Emergency boron injection system Double-envelope containment and core
catcher.
Emergency feedwater system
Residual heat removal system
Main steamline isolation system.


TURBIN


      The main parameters of the turbine plant operation under nominal conditions are given in Table 8.


Table 8 Reactor Turbine Main Parameter

Parameter Value
Nominal temperature of mainsteam, °C 283.8
Nominal humidity of main steam, % 0.8
Calculated temperature of cooling water, °C 18
Nominal absolute steam pressure in condenser, kPa 4.9
Feedwater temperature, °C 227


EXAMPLE POWER SYSTEM WITH NPPs OF SMILAR TYPE


     So far the operating experience of VVER-reactor equipped NPPs has been about 1400 reactor-years (including the commissioned), of which about 500 reactor years account for the operation of NPPs equipped with VVER-1000. The NPP to VVER-1000 design were implemented at the following sites:

• Kozloduy NPP in Bulgaria,

• Tianwan NPP in China,

• Bushehr NPP in Iran,

• Kudankulam NPP in India

• Rovno NPP , Zaporozhe NPP, Khmelnitskaya NPPve South-Ukraine NPP
   (Ukraine),

• Novovoronezh NPP, Balakovo NPPve Rostoc NPP(Russia)[3].



REFERENCES


[1]:  “TAEK resmi sitesi”, Date of access: 18.09.2017,
“http://www.taek.gov.tr/nukleer-guvenlik/nukleer-enerji-ve-reaktorler/165-
akkuyu-nukleer-guc-santrali/428-akkuyu-vver-1200-nukleer-santrali.html”

[2]:  “TAEK resmi sitesi”, Date of access: 18.09.2017,
“http://www.taek.gov.tr/nukleer-guvenlik/nukleer-enerji-ve-reaktorler/165-
akkuyu-nukleer-guc-santrali/428-akkuyu-vver-1200-nukleer-santrali.html”

[3]:  IAEA VVER1200 REPORT, Web, Date of access: 18.09.2017,
“https://www.iaea.org/NuclearPower/Downloadable/aris/2013/36.VVER-
1200(V-491).pdf”