preliminary safety analysis report (psar) chapter 1

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Fennovoima Oy | fennovoima.fi | +358 20 757 9200 | Salmisaarenaukio 1, FI-00180 Helsinki, Finland | Business ID 2125678-5

FH1 Program PSAR Chapter 1

General Plant Description

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PSAR CHAPTER 1 – GENERAL PLANT DESCRIPTION

CONTENTS

CONTENTS ................................................................................................................... 2

ABBREVIATIONS ........................................................................................................... 5

1.1 Plant location ........................................................................................................ 8

1.1.1 Physical and Geographical Location .................................................................... 8

1.1.2 Climatic Conditions .......................................................................................... 9

1.1.3 Hydrological conditions ................................................................................... 12

1.1.3.1 Hydrological characteristics of the service water supply source ...................... 14

1.1.4 Geological and hydrogeological conditions at the site ......................................... 16

1.1.5 Seismotectonic conditions ............................................................................... 22

1.1.6 References .................................................................................................... 27

1.2 Overview of the plant layout and building arrangements .......................................... 32

1.2.1 Description of buildings .................................................................................. 36

1.3 General Plant description ..................................................................................... 39

1.3.1 Brief description of VVER design evolution ........................................................ 39

1.3.2 Overview of the plant operation ....................................................................... 41

1.3.2.1 Overview of the Nuclear Steam Supply System (NSSS) ................................. 46

1.3.2.2 Overview of the Engineered Safety Features ................................................ 47

1.4 General Safety Design principles ........................................................................... 51

1.4.1 Fundamental safety objectives ........................................................................ 51

1.4.2 Safety Policy ................................................................................................. 51

1.4.3 Defense-in-Depth concept .............................................................................. 52

1.4.4 Event categories ............................................................................................ 56

1.4.5 Technical implementation and ensuring reliability of safety functions fulfillment ..... 57

1.4.6 Autonomy ..................................................................................................... 59

1.4.7 Internal and external hazards ......................................................................... 62

1.4.8 Principles of safety classification ...................................................................... 63

1.4.9 Radiation safety ............................................................................................ 63

1.5 NPP general systems description ........................................................................... 65

1.5.1 Reactor Coolant System ................................................................................. 65

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1.5.1.1 Reactor coolant pumps ............................................................................. 65

1.5.1.2 Steam generators .................................................................................... 66

1.5.1.3 Reactor Coolant Piping .............................................................................. 67

1.5.1.4 Pressurizing and steam discharge system .................................................... 67

1.5.2 Safety Systems ............................................................................................. 69

1.5.2.1 Spray System (JMN) ................................................................................. 69

1.5.2.2 Passive Containment Heat Removal System (JMP)........................................ 70

1.5.2.3 Corium Localization System (JMR) .............................................................. 71

1.5.2.4 System of Emergency Use of Water from Reactor Internals Inspection shaft (JNB90) ............................................................................................................. 72

1.5.2.5 Containment Isolation System (JMK) .......................................................... 73

1.5.2.6 High Pressure Safety Injection System (JND) .............................................. 74

1.5.2.7 Emergency Core Cooling System, passive part (JNG2) .................................. 75

1.5.2.8 Low Pressure Safety Injection System (JNG1).............................................. 76

1.5.2.9 Borated Water Storage System (JNK) ......................................................... 78

1.5.2.10 Emergency Boron Injection System (JDH) ................................................. 80

1.5.2.11 Emergency Gas Removal System (KTP) ..................................................... 81

1.5.2.12 Emergency Feedwater System (LAR/LAS) .................................................. 81

1.5.2.13 System of Passive Heat Removal Through Steam Generators (JNB) ............. 83

1.5.2.14 Residual Heat Removal System (JNA) ....................................................... 84

1.5.2.15 Intermediate Cooling Circuit For Important Consumers (KAA) ..................... 85

1.5.2.16 Containment Hydrogen Removal System (JMT) .......................................... 86

1.5.3 Instrumentation and control systems (I&C) ....................................................... 88

1.5.3.1 General ................................................................................................... 88

1.5.3.2 Basic tasks of I&C systems ........................................................................ 89

1.5.3.3 Design requirements ................................................................................ 90

1.5.3.4 Security .................................................................................................. 91

1.5.3.5 I&C functions ........................................................................................... 91

1.5.3.6 I&C systems architecture and systems ........................................................ 91

1.5.3.7 Control rooms and human system interfaces ............................................... 92

1.5.3.8 HFE ........................................................................................................ 93

1.5.4 Power supply system ..................................................................................... 93

1.5.4.1 General ................................................................................................... 93

1.5.4.2 Operation of power distribution system ....................................................... 96

1.5.4.3 Plant auxiliary power supply system ........................................................... 96

1.5.4.4 Alternating current electrical systems ......................................................... 99

1.5.4.5 Direct current electrical systems ............................................................... 100

1.5.4.6 Location of equipment ............................................................................. 100

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1.5.5 Auxiliary Systems ......................................................................................... 102

1.5.5.1 Fuel Storage and Handling ....................................................................... 102

1.5.5.2 Water Systems ....................................................................................... 105

1.5.5.3 Process Auxiliary Systems ........................................................................ 107

1.5.5.4 Heating, Ventilation and Air Conditioning (HVAC) Systems ........................... 114

1.5.5.5 Radiation Monitoring Systems ................................................................... 115

1.5.5.6 Other Auxiliary Systems ........................................................................... 117

1.5.6 Steam and Power Conversion System ............................................................. 119

1.5.6.1 Turbine Generator ................................................................................... 119

1.5.6.2 Main Steam System (LBA/LBU) ................................................................. 120

1.5.6.3 Main Condenser (MAG) ............................................................................ 121

1.5.6.4 Turbine Bypass System (MAN) .................................................................. 121

1.5.6.5 Main Circulating Water System (PA) .......................................................... 122

1.5.6.6 Condensate and Feedwater systems .......................................................... 122

1.5.6.7 Auxiliary Steam System (LBG) .................................................................. 124

1.5.6.8 High Pressure Steam Line Drains System (MAL30) ...................................... 125

1.5.7 Radioactive Waste Management ..................................................................... 125

1.5.7.1 Liquid Waste Management Systems ........................................................... 125

1.5.7.2 Gaseous Waste Handling Systems ............................................................. 129

1.5.7.3 Solid Radioactive Waste Handling System (KPA) ......................................... 129

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ABBREVIATIONS

AC Alternating current

AES Atomic electric station

AIRDMS Automated Individual Radiation Dose Monitoring System

ALARA As Low As Reasonably Achievable

ANT Auxiliary normal transformers

AOO Anticipated Operational Occurrences

APRMS Automated Process Radiation Monitoring System

ARCMS Automated Radiation Contamination Monitoring System

ARMS Automated Radiation Monitoring System

ASRMS-R&S Automated System of Radiation Situation Monitoring in Rooms and on

Site

AST Auxiliary standby transformers

ATWS Anticipated Transient Without Scram

BRU-A Atmospheric steam dump valve

BRU-D Steam dump valves to deaerator

BRU-K Steam dump valves to turbine condenser

BRU-SN Steam dump valve to the auxiliaries

CCTV Closed Circuit Television

DADS Defective Assemblies Detection System

DBA Design Basis Accidents

DBC Design Basis Condition

DBE Design Basis Earthquake

DC Direct current

DEC Design Extension Conditions

DiD Defense-in-Depth

ECCS Emergency Core Cooling System

ECR Emergency control room

EPSS Emergency Power Supply System

ERCP Emergency response command post

EUR European Utility Requirements

EYT Non-nuclear safety (in regulatory safety classification)

FA Fuel assembly

FENCAT Fennoscandian earthquake catalogue

FHM Fuel handling machine

FRIE Fuel repair and inspection equipment

GMPE Ground motion prediction equations

GSSC Gland seal steam condenser

HFE Human factors engineering

http://www.multitran.ru/c/m.exe?t=2875094_1_2&s1=%E0%F2%EE%EC%ED%E0%FF%20%FD%EB%E5%EA%F2%F0%EE%F1%F2%E0%ED%F6%E8%FF

http://www.google.ru/url?sa=t&rct=j&q=&esrc=s&source=web&cd=1&cad=rja&uact=8&ved=0CBwQFjAA&url=http%3A%2F%2Fwww.europeanutilityrequirements.org%2F&ei=IpaWVPnGBoW2ygOX3oKoDA&usg=AFQjCNFtmnal29DcWKqXtKe-3EF5vyr3WA&bvm=bv.82001339,d.bGQ

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HP High pressure

HSI Human system interface

HVAC Heating, ventilation and air conditioning

I&C Instrumentation and control

IAEA International Atomic Energy Agency

ICRP International Commission on Radiological Protection

IEC International Electrotechnical Commission

INSAG International Nuclear Safety Advisory Group

IP Intermediate pressure

LCCS Leak Check of Claddings System

LOCA Loss of Coolant Accident

LP Low pressure

LPH Low pressure heaters

LWR Light Water Reactor

MCR Main control room

MSH Main steam header

MSIV Main steam isolation valve

MSR Moisture separator reheater

MSSH Mist-spraying sprinkler heads

M.A.S.L. Meters above sea level

NF Moderately cold

NI Nuclear Island

NO Normal Operation

NPP Nuclear power plant

NSSS Nuclear Steam Supply System

PHRS Passive Heat Removal System

PRISE Primary-to-secondary

PRZ Pressurizer

PRZ PORV Pressurizer pilot-operated relief valve

PRZ TEH Pressurizer tubular electric heater

PSAR Preliminary safety analysis report

PSHA Probabilistic seismic hazard analysis

PSS Power Supply System

PWR Pressurized Water Reactor

RCCA Rods cluster control assemblies

RCP Reactor coolant pump

RCPS Reactor coolant pump set

RI Reactor internals

RMS Radiation Monitoring System

http://www.google.ru/url?sa=t&rct=j&q=&esrc=s&source=web&cd=8&ved=0CDMQFjAH&url=http%3A%2F%2Fen.wikipedia.org%2Fwiki%2FInternational_Commission_on_Radiological_Protection&ei=BZWWVOnZEcapyQPn7oLAAQ&usg=AFQjCNEV3NoDHFLx3RxjHQdJ76AcUFPkmQ&bvm=bv.82001339,d.bGQ

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RP Reactor plant

RPSS Reliable Power Supply System

RPV Reactor pressure vessel

SA Severe Accident

SAHARA Safety As High As Reasonably Achievable

SDGS Standby diesel generator station

SG Steam generator

SNSN Swedish National Seismic Network

TI Turbine Island

UDGS Unit diesel generator station

UHS Ultimate heat sink

UKHL Cold macroclimatic area

UPS Uninterrupted power supply

US NRC United States Nuclear Regulatory Commission

VVER Vodo Vodyanoy Energeticheskiy Reaktor

WENRA West European Nuclear Regulators Association

YVL Finnish nuclear regulatory guides

http://www.google.ru/url?sa=t&rct=j&q=&esrc=s&source=web&cd=2&cad=rja&uact=8&ved=0CCkQFjAB&url=http%3A%2F%2Fen.wikipedia.org%2Fwiki%2FNuclear_Regulatory_Commission&ei=1JWWVLu7KeTHygO7zYGgCw&usg=AFQjCNFtIbKmt_cr_63ZrJghG3R2HkSbKg&bvm=bv.82001339,d.bGQ

http://www.multitran.ru/c/m.exe?t=4117417_2_1&s1=VVER

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1.1 Plant location

1.1.1 Physical and Geographical Location

The Fennovoima nuclear power plant (NPP) will be constructed on the Hanhikivi

headland in Pyhäjoki, Finland. The site is located on the west coast of Finland, Northern

Ostrobothnia, in Pyhäjoki and Raahe municipalities. The center of the Pyhäjoki

municipality is located about six kilometers south of the Hanhikivi headland. Distance to

the center of Raahe is about 20 kilometers, to Oulu and Kokkola about 100 kilometers

(Figure 1.1.1.1). There is no previous industrial activity at the plant location. The

nearest industrial activity is mainly steel industry in the Raahe area. The Oulu region

has pulp and paper, chemical and electronic industry. The Kalajoki area focuses on

tourism.

Figure 1.1.1.1 The Location of the Hanhikivi headland plant site.

The immediate surroundings of the Hanhikivi power plant site are sparsely populated.

Approximately 440 permanent residents live within the five-kilometer precautionary

action zone. There are about 11 400 permanent residents within a twenty-kilometer radius of the site. (Tilastokeskus, 2013)

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Traffic connections to the plant site are good. There are five international ports and two

international airports in the coastal area. The nearest port is located in Raahe about 15

kilometers to the northeast; Oulu and Kokkola airports are located about 100 kilometers

from Pyhäjoki. Oulu airport is the second largest airport in Finland. A railway passes

through the region from the south to the north via Oulu–Ruukki–Vihanti–Oulainen–

Ylivieska–Kokkola serving the ports of the coastal region and providing a transport

connection to eastern regions of the country and Russia. The E8 road traverses the precautionary action zone and can also be used for transportation.

1.1.2 Climatic Conditions

The climate in Finland has both continental and maritime characteristics, which are

heightened depending on the direction of airflow. The annual mean temperatures are

raised by the proximity of the Baltic Sea and especially the Gulf Stream in the Atlantic.

However, extreme temperatures may occur especially with Asian continental air reaching Finland, resulting in severe cold in the winter and heat waves in the summer

According to Köppen climate classification, Finland is referred to as a softwood and

mixed forests zone with wet and cold winters. Köppen climate classification subtype for

this climate is Continental Subarctic climate (Dfc). According to the International

classification, defined in GOST 15150-69*, the site area belongs to a moderately cold macroclimatic area (UKHL or NF).

The main climate characteristics are given according to the Finnish Meteorological

Institute (FMI), which has Finnish climatic characteristics for the period of 1981–2010 presented on its official web site.

Air temperature

The average air temperature parameters based on long-term observations near

Hanhikivi (Oulunsalo Oulu airport weather station, observation years 1981–2010) are

presented in Table 1.1.2.1. Annual average air temperature at Hanhikivi headland for

the period of 1981–2010 was 2.7 °С.

Table 1.1.2.1 Air temperature, ºС, Oulu weather Station, 1981–2010.

Month Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec

Absolute max in

1981–2010 (°C) 7.2 7.8 10.1 20.3 28.4 31.7 33.0 29.3 24.2 16.4 10.7 8.0

Mean daily max (°C)

−6.0 −5.7 −0.9 5.6 12.5 17.9 20.9 18.3 12.5 5.8 −0.4 −4

Mean temperature (°C)

−9.6 −9.3 −4.8 1.4 7.8 13.5 16.5 14.1 8.9 3.3 −2.8 −7.1

Mean daily min (°C)

−13.6 −13.3 −8.8 −2.6 3.3 9.0 12.2 10.1 5.4 0.8 −5.5 −10.8

Absolute min in

1981–2010 (°C) -37.5 -36.2 -32.0 -19.2 -6.7 0.2 4.2 -1.5 -5.5 -20.1 -28.4 -37.2

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The absolute temperature records according to Oulu weather station observations from

1961 to 2013 are:

– Maximum +33.0 °C (2010)

– Minimum -41.5 °C (1966)

Air humidity

The average air humidity parameters based on long-term observations for Oulu Airport weather station are presented in table 1.1.2.2.

Air humidity is high due to the prevailing sea air masses. Annual average humidity

value is 80%, reaching its maximum in November and December (89–90%), and

minimum in May and June (67–66%). The maximum value of relative humidity (100%)

is reached frequently. According to some incomplete observation data from Oulu weather station, the minimum value has been 16%.

Table 1.1.2.2. Air Humidity, Oulu Airport Weather Station, 1981–2010. (FMI, 2012)

Month Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Year

Mean relative

humidity (%) 87 86 82 73 67 66 71 76 82 86 90 89 80

Precipitation

The average precipitation in different months in the Hanhikivi area is shown in Table

1.1.2.3, based on the observation data from Hailuoto Ojakylä weather station in 1981–

2014. The weather station is located 60 km north of Hanhikivi in similar coastal

conditions. The precipitation level in Ostrobothnia is lower than for the whole country

(600–700 mm) due to closeness of the seacoast; annual precipitation is 477 mm

according to Oulu meteorological station data.

Table 1.1.2.3 Precipitation, Hailuoto Ojakylä Weather Station, 1981–2010. (Pirinen, Simola, Aalto, Kaukoranta, Karlsson, & Ruuhela, 2012)


Mean

precipitation

(mm)

36 30 31 25 36 41 63 60 48 54 47 37 508

Mean number of

precipitation

days (≥1.0mm)

10 8 8 6 7 7 9 10 9 10 11 10 105

Snow cover

The average snow depth in different months in the Hanhikivi area is shown in Table

1.1.2.4, based on the observation data from Hailuoto Ojakylä weather station in 1981–

2014. The weather station is located 60 km north of Hanhikivi in similar coastal

conditions. Steady snow cover at the site appears at the end of November or beginning

of December. Maximum snow depth is reached at the end of March (45–48 cm). The

snow cover melts at the end of April.

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Table 1.1.2.4 Average monthly snow depths in the Hanhikivi area, based on Hailuoto

Ojakylä observation data.

Month Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec

Snow depth, 15th

of the month (cm) 26 40 47 20 0 0 0 0 0 0 5 11

Snow depth, last day of the month

(cm)

32 43 40 3 0 0 0 0 0 1 10 17

Wind regime

The average wind speeds in the Hanhikivi area based on observations from Raahe

Lapaluoto weather stations in 1996–2013 and presented in Table 1.1.2.5. According to the statistics, fall and winter are the windiest seasons.

Table 1.1.2.5. Average wind speeds in the Hanhikivi area based on Raahe Lapaluoto Weather Station, 1996–2013.


Mean wind speed (m/s)

6.3 6.2 6.0 5.4 5.4 5.1 4.9 5.0 6.3 6.7 6.9 6.4 5.9

Mean max wind speed (m/s)

17.8 16.5 15.4 14.6 14.2 13.9 12.5 13.7 16.8 17.3 17.7 17.4 15.6

Wind speed

fraction (0–10 m/s) 85.5% 85.9% 89.5% 93.4% 94.5% 95.3% 96.6% 95.8% 86.3% 83.2% 82.7% 84.1% 89.5%

Wind speed

fraction (10–20 m/s)

14.4% 14.0% 10.4% 6.6% 5.5% 4.7% 3.4% 4.2% 13.7% 16.7% 17.0% 15.8% 10.5%

Wind speed fraction (>20 m/s)

0.1% 0.0% 0.1% 0.0% 0.0% 0.0% 0.0% 0.0% 0.0% 0.1% 0.2% 0.1% 0.1%

The most common wind directions are south and southwest, whereas northwest is the

most improbable wind direction. (Kauppi, 2014)

Abnormal wind speeds can occur at Hanhikivi Plant site if trombs or downbursts occur.

A tromb is a rotating column of air occurring between a thundercloud and earth surface.

Horizontal wind speeds in the rotating funnel have been observed to reach 100 m/s or more. In total, 296 have been identified in Finland since 1796.

Downburst is the name for an intense descending burst of cold air originating from the

middle parts of a thundercloud. When the rapidly downward moving air (downdraft) hits

the ground, it turns and flows horizontally in all directions as a straight-line wind.

Downburst wind speeds are the greatest at the point where the air hits the ground.

(Mäkelä & Hyvärinen, 2014). Downbursts have a small effect area and short lifetime so the identification of downbursts is not easy.

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Characteristics of air pollution sources

There are no industrial facilities in Pyhäjoki municipality that could impair air quality,

hence no measurements have been carried out in this area. The nearest air quality

monitoring stations are in Raahe (20 km from the site). In 2012, air quality in Raahe

was generally estimated as “good” (Ramboll 2013). Based on this survey, the air quality

at Hanhikivi headland may also be considered “good”, because the distance to Raahe area is only 20 kilometers and there are no industrial facilities located in the area.

1.1.3 Hydrological conditions

The coastline around the Hanhikivi headland is open, and water changes efficiently in

the area. The coastal waters surrounding the headland are very shallow and rocky. The

water depth in the planned dock basin and breakwater area varies between 0 and 3.7

meters. Water depth in front of the cooling water discharge structures is approximately

0.3 meters. The depth of the water around the Hanhikivi headland increases very

slowly, first at the rate of one meter per 100 meters. A depth of 10 meters is achieved

approximately one kilometer from the northwestern tip. Depths exceeding 20 meters

are not found until 10 kilometers to the west of the headland. In the planned marine

spoil area, the water is approximately 15–25 meters deep. (Fennovoima Oy, 2014b). The water depth around Hanhikivi headland is shown in figure 1.1.3.1.

The most important river is the Pyhäjoki river, which runs to the sea approximately six

kilometers to the southwest of the Hanhikivi headland. The average discharge of the

Pyhäjoki river is approximately 30 m3/s.

The NPP has once-through service water supply system. The Gulf of Bothnia is

considered as a source of service water supply. The Hanhikivi NPP waterworks layout is shown in figure 1.1.3.2.

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Figure 1.1.3.1 Water Depth Around the Hanhikivi Headland.

The nearest active sea hydrological stations with long-term water level observations

data (since 1922) are situated in Raahe (18 km north of the site) and Pietarsaari (at

120 km’s distance). Fennovoima has organized occasional seawater quality observations in the area of NPP influence since 2009.

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Figure 1.1.3.2 Layout scheme for the Hanhikivi NPP service water supply system waterworks.

1.1.3.1 Hydrological characteristics of the service water supply source

Sea water level

Sea water level can be given in a geodetic leveling system (N2000 height system is

used in the Hanhikivi 1 project) or in relation to the theoretical mean sea level, which is

a forecast for the long-term sea level mean value and made for practical purposes. The

height of the theoretical mean sea level varies depending on location and time and is

confirmed yearly by FMI. For example, the theoretical mean sea level in Raahe was

+12.2 cm in the N2000 system in 2013, +11.8 cm in 2014 and +11.4 cm in 2015. The

yearly conversion values for different locations and systems can be found in (FMI, 2014).

The marigraph closest to Hanhikivi is in Raahe, and its observations in 1922–2013 are

shown in Table 1.1.3.1.1. The records are different, depending on whether we consider

the extreme values relative to theoretical mean sea level or extreme values in the

N2000 height system. The maximum water level +162 cm relative to the mean sea level corresponding 207 cm in N2000 height system was registered on 1962.

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Table 1.1.3.1.1 Sea water level records in Raahe relative to theoretical mean sea level

and in the N2000 height system. (Asp, Södling, Edman, Nerheim, & Sjökvist, 2014)

Year Sea level (cm),

relative to mean level Sea level (cm),

N2000 height system

1962 162 207

1925 136 209

1936 -129 -64

1998 -121 -102

The different factors affecting sea water level have been identified and their effects

quantified in (Asp, Södling, Edman, Nerheim, & Sjökvist, 2014). The results are listed in

Table 1.1.3.1.2 The magnitudes presented are applicable to the northern coast of the

Bothnian Bay (Kemi), whereas the levels at Hanhikivi can be expected to be roughly 0.2 m lower.

Table 1.1.3.1.2 Description of factors affecting water levels at the northern part of the Gulf of Bothnia.

Factor Period Range Comments

Seasonal

mean sea

level

Days,

weeks

0.5 m During periods of prevailing west winds,

the Baltic Sea receives a net inflow of

water through the straits. In combination

with local winds, average level rise of

around 0.5 m are plausible.

Low pressure

and onshore

winds

3–24 hours 1.0–1.5 m During storm events, a combination of on-

shore winds and low pressure can lift the

sea level with up to 1.5 m

Seiche 27 hours 0.5 m The amplitude is given for Kemi and is

somewhat lower further to the south

Tides Daily and

semidiurnal

0.1 m Tides have been considered negligible in

the Baltic Sea, due to the choking effect

in the strait. However, a small tidal

amplitude is found in the Baltic Sea.

Meteotsunami 5–15 min 0.5 m

(assess-

ment)

A meteotsunami is a tsunami-like wave

phenomenon of meteorological origin. It

can be defined as an atmospherically

generated large amplitude seiche

oscillation.

In practice, a high sea level may occur if prevailing westerly winds have increased the

total amount of water in the Baltic Sea, and a low pressure center pushes large

amounts of water towards land together with onshore winds. A tide could also occur

simultaneously, but the occurrence of seiche and meteotsunami together with a low

pressure center can be considered unlikely. Altogether, the water level near Hanhikivi

could be expected to rise by approximately 2 meters (N2000 height system).

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Seawater temperature

According to the water temperature monitoring data of 2009–2013, measured from the

Gulf of Bothnia close to the Hanhikivi NPP, the temperature reaches its maximum in

July and at the beginning of August. Maximum measured temperature of the surface

layer near the NPP water intake was 18.7 ºС. Minimum winter temperature was minus

0.2 ºС, which corresponded to water mineralization of 3‰.

Sea ice

The ice cover usually starts to form in the inner coves of the Bothnian Bay in mid-

November. The ice starts to break away in May. Formation of ice occurs unevenly, but

generally ice forms at around 2.5 mm per degree below zero temperature per day.

Usually, sea ice is in constant motion except for the edge areas. In the Bothnian Bay,

the ice thickness has been observed to be between 45–60 cm in mild winters, 75–90

cm in normal winters and 105–120 cm in harsh winters (Kärnä, 2008). Total number of days with ice cover in a year has ranged from 127 to 199 (Pöyry Oy, 2009).

Bottom soils, turbidity

Bottom soils in the NPP water intake area are sands of different particle sizes, with

inclusions of small amounts (5–20%) of gravel with 2–10 mm gradation. (VitusLab

(Research and consultancy on ocean and climate), 2012).

Water turbidity is defined by the presence of suspended matter, phyto and zooplankton.

Average turbidity of seawater varied from 0.3 to 3.1 FTU during monitoring in 2009–2013.

Seawater quality

The quality of seawater in front of the Hanhikivi headland corresponds to the water

quality typically found along the Bothnian Bay coast. In the ecological assessment of

the Finnish environmental administration (2013), the water quality of the sea in front of

the Hanhikivi headland was classified as moderate. Further (over 300 meters) from the coast, the quality of the water was classified as good.

1.1.4 Geological and hydrogeological conditions at the site

Geological conditions

The Hanhikivi site is situated in a continental intraplate setting, in the northern part of

the Fennoscandian Shield in the Pohjanmaa Belt. The main rock types in this Belt are

turbiditic mica schists and gneisses with mafic and intermediate volcanic rocks, black

schists, metacherts and carbonate rocks as interlayers. The conglomerates and arkosic

rocks in the northern part represent the youngest sedimentation in the belt.

Metamorphic grade increases in the center of the belt towards granulite facies.

Granitoids of 1880 Ma age are crosscutting the supracrustal rocks. (Kähkönen, 2005); (Vaasjoki, 2005).

Ground level in the Hanhikivi area is from 0 to +4 m above sea level (m.a.s.l.)

(Pohjatekniikka Oy, 2009). The overburden thickness varies between 0 m and 6 m, up

to 8–12 m in bedrock depressions. The overburden layers mostly consist of sand and

moraine, though layers of silt or clay are also encountered. In topographic lows, the upper layer may be peat. (Pitkäranta, 2012).

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The bedrock in the Hanhikivi site area (Hanhikivi headland) is part of the supracrustal

Pohjanmaa Belt (Kähkönen, 2005). The Hanhikivi headland consists mostly of

volcaniclastic metaconglomerate. In the west, the conglomerate unit continues under

the sea. (Elminen, 2008) The metaconglomerate is strongly schistose; the schistosity

dips in a 50 to 90 degree angle to the south or southeast. The conglomerate contains

mainly volcanogenic or subvolcanogenic spherical or elongated pebbles and fine grained matrix material. (Melcher, 1953); (Salli, 1964); (Elminen, 2008).

The bedrock surface elevation varies typically between -3…+1 m.a.s.l. Ground

topography is generally flat, but locally, bedrock topography shows sharp depth

changes and over 10 m overburden thicknesses (Valjus, 2008). It is usual that the

bedrock topography varies more than the ground surface, because the overburden

tends to smoothen the bedrock relief. There are two larger areas where the bedrock lies

mostly above the sea level: the central and western parts of the cape and in the

mainland, immediately east of the cape proper. In these elevated areas, the till layer is

usually thin and this is the area where most of the outcrops are found (Geobotnia Oy, 2008).

Interpretations on the exact locations of the subsea deformation zones differ from study

to another, but the general view is that Hanhikivi is located inside a bedrock block

which is bordered from south and north by NW-SE and NE-SW trending deformation

zones (Figure 1.1.4.1). These lineaments are classified as Class 2 structures (length 5–

20 km, width hundreds of meters) and they are located 5–10 km from the proposed

plant site (Kuivamäki, 2011). A relatively small number of lower class (Class 3 or 4) internal lineaments suggests that the bedrock block is intact.

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Figure 1.1.4.1 Morphological lineaments in the Hanhikivi area plotted on a topographic

elevation model and bathymetric map. The lineaments are classified based on their

lengths (Kuivamäki, et al., 2011). Pink lines = length 5–20 km, blue lines = length 1–5 km, black line = length 0.5–1 km, and grey lines = length < 0.5 km

The mineral potential of the Hanhikivi area was examined using the Geological Survey

of Finland's nationwide database. There are no indications of any type of mineral

showings in the Hanhikivi metaconglomerate. Geochemical mapping and airborne

geophysical measurements indicate that the area is not particularly potential for metallic or industrial minerals. (Kärkkäinen & Lohva, 2008)

Hydrogeological conditions

The nuclear power plant location has no registered ground water deposits. The nearest

classified deposit (Kopisto, category I, ID 11625001) is located approximately 10 km to the southeast of the power station location.

According to preliminary studies, ground water lies near the surface within the site.

During the investigation, the groundwater table level varied from 0.1 m to 0.6 m

(between +0.2 and+1.0 m abs.). During the year, variation of the groundwater table level was in the range between -0.5 and +1.5 m abs.

Figures 1.1.4.2 and 1.1.4.3 show locations of water wells that were used for short-term

observation of ground water dynamics and chemical composition in 2012 and

hydroisobath map compiles as of 8/14/2012.

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The groundwater level is generally near the surface on the Hanhikivi headland. The

groundwater level is approximately 0–1.5 m above sea level. The formation of

groundwater is estimated to be about 20% of precipitation. In the Hanhikivi area (a

surface area of 350 hectares), the formation rate of the groundwater is 1000 m3 per

day in average, when precipitation is about 500 mm per year. The groundwater flow at

the soil layer is very slow due to low gradient and low water conductivity of the soil

layer in the area. Groundwater discharge areas have not been identified in the Hanhikivi

area. The groundwater is most probably infiltrated to bedrock groundwater and

discharged to the sea. On the other hand, seawater can also be discharged into

groundwater. The groundwater reservoir at Hanhikivi area is small and has no influence to the construction of the power plant. (Pitkäranta, 2012).

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Figure 1.1.4.2 Piezometer and observation hydrogeological well location diagram.

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Figure 1.1.4.3 Near-surface aquifer hydroisobath sketch map. Dots show wells used for data for map compilation.

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1.1.5 Seismotectonic conditions

The seismotectonic conditions near the Hanhikivi site have been presented in (Korja & Kosonen, 2014) and probabilistic seismic hazard estimation in (Saari, et al., 2015).

Area seismicity and earthquake catalogue

Finland is located on the Baltic Shield—one of the least active seismic areas in the

world. Global seismotectonic structures nearest to the NPP location site, with a stable

seismic setting and high energies of regular strong earthquakes, is a system of mid-

oceanic rifts and transform faults of the North Atlantic. Other factors, such as post-glacial elevations are local by nature.

Earthquake recurrence rate in Fennoscandia is very low compared to plate boundary

regions of the world. Nevertheless, Fennoscandia is a seismically active region, though it has a low earthquake recurrence rate with relatively low magnitudes.

The preparation of a declustered, homogenized parametric earthquake catalogue is

presented in (Saari, et al., 2015). The catalogue contains tectonic earthquakes within a

radius of 500 km from the Hanhikivi site. The earthquakes for 1610–2011 have been

extracted from the Fennoscandian earthquake catalogue (FENCAT), which is the main

source of seismicity data available in the study region. These data have been

supplemented with a preliminary version of the 2012 FENCAT as well as with micro-

earthquakes (M < 1) recorded by Swedish National Seismic Network (SNSN) in 2000–

2013. Non-tectonic events such as rock bursts, mine collapses and suspected non-

seismic events have been removed from the data. Magnitude conversions were

performed from several magnitude units to moment magnitude (MW). The declustered,

manually revised catalogue contains 4587 independent observations. The homogenized

magnitudes of the main shocks vary between -0.2 and 4.3. The homogenized earthquake catalogue is shown in Figure 1.1.5.1.

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Figure 1.1.5.1 Epicenter map of the homogenized earthquake catalogue (1610–2012). (Saari, et al., 2015)

Seismotectonic framework and seismic source areas

Seismic source areas around the Hanhikivi site were determined in (Korja & Kosonen,

2014). Three new spatial models for seismic source areas were identified and described.

In model 1, the analysis focused on the potential reactivation of geologically ancient

features (Figure 1.1.5.2). The model combined data sets bearing on historical and

instrumental seismicity, lithology, deformation zones including brittle components

(faults), lineaments defined on the basis of magnetic and gravity data, and broader

crustal structure including Moho depth. In Model 2 (Figure 1.1.5.3), recently active

structures were analyzed using data sets bearing on the latest high-quality seismicity

data, post-glacial faults, topography, bathymetry, lineaments defined on the basis of

magnetic data, and the current stress field. Model 3 is a slightly modified version of

spatial Model 2 and contains additional polygons. Only models 1 and 2 were included in

the probabilistic seismic hazard assessment to reduce the number of branches in the

logic tree.

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Figure 1.1.5.2. Seismic source areas in spatial Model 1 and their relationship to

seismicity (blue dots) and major deformation zones extracted and modified from the

national bedrock databases at the scale 1:1 M for Sweden and Finland. Hanhikivi – orange dot. Modified from (Korja & Kosonen, 2014).

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Figure 1.1.5.3 Earthquake epicenters (black dots), seismic source areas identified in

spatial Model 2 and their relationship to faults inferred to have been active (red) during

the late Pleistocene or Holocene on the basis of Quaternary geological data. The

bathymetric character of both the Gulf of Bothnia and the Gulf of Finland is also shown. Hanhikivi site: orange dot. (Korja & Kosonen, 2014)

Probabilistic seismic hazard assessment

Based on the seismic source areas, probabilistic seismic hazard assessment (PSHA) for

the Hanhikivi site was made in (Saari, et al., 2015). New ground motion prediction

equations (GMPE) were developed based on Fennoscandian data. Seismic activity

parameters were calculated for the whole study region and for the different seismic

source areas in models 1 and 2. Zone-specific maximum magnitudes were not

assumed, but instead maximum magnitudes of 5.5, 6.0, 6.5 and 7.0 were used in the

PSHA with different weights. The maximum observed magnitude in the project

catalogue was MW= 4.3. A logic tree with five levels (seismic source area model,

seismic activity parameters, maximum magnitudes and GMPEs) was used in the PSHA

calculations.

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The results of probabilistic seismic hazard assessment (PSHA), presented in (Saari, et

al., 2015), include the median seismic hazard curve of the computed acceleration values. The calculated seismic hazard curve is shown in Figure 1.1.5.4.

Figure 1.1.5.4 Seismic hazard curve for the Hanhikivi site. (Saari, et al., 2015)

1,00E-10

1,00E-09

1,00E-08

1,00E-07

1,00E-06

1,00E-05

1,00E-04

1,00E-03

1,00E-02

1,00E-01

0,001 0,01 0,1 1 10

An

nu

al

freq

uen

cy o

f exceed

an

ce

Acceleration (g)

median

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1.1.6 References

Airo, M.-L., Leväniemi, H., & Valjus, T. (2011). Geofysikaalisten aineistojen tulkinta.

Etelä-Suomen yksikkö. Espoo: Geologian tutkimuskeskus.

Alanen, U., Alvi, K., Hämäläinen, J., Kaskela, A., & Rantataro, J. (2013). Pyhäjoen

Hanhikiven edustan merialueen akustis-seismiset tutkimukset 2012. Geologian tutkimuskeskus.

Asp, M., Södling, J., Edman, A., Nerheim, S., & Sjökvist, E. (2014). Evaluation of

Extreme Weather Events in Pyhäjoki, Finland. <In development>. RE-03-0000024.

Swedish Meteorological and Hydrological Institute.

Bogdanova, S., Bongen, B., Gorbatshev, R., Kheraskove, T., Kozlov, V., Puchkov, V., et

al. (2008). The East European Craton (Baltica) before and during the assembly of Rodinia. Precambrian Research, 160(1-2), 23-45.

Donner, J. (1995). The quaternary history of Scandinavia. Cambridge: Cambridge University Press.

Elminen, T. L. (2008). Ydinvoimalaitoksen mahdollisten sijoituspaikkojen geologiset

esiselvitykset. Helsinki: Geological Survey of Finland.

Elminen, T., Lohva, J., & Härmä, P. (2008). Ydinvoimalaitoksen mahdollisten sijoituspaikkojen geologiset esiselvitykset. Helsinki: Geological Survey of Finland.

Fennovoima Oy. (2014b). Environmental Impact Assessment Report for a Nuclear Power Plant. Helsinki: Fennovoima Oy.

FMI. (2012). Tilastoja Suomen ilmastosta 1981-2010.Ilmastotilastoja Suomesta, No. 2012:1. Helsinki: Finnish Meteorological Institute.

FMI. (2014). Finnish Meteorological Institute, Theoretical mean sea level. Retrieved 10

8, 2014, from http://en.ilmatieteenlaitos.fi/theoretical-mean-sea-level

Forssell, T., & Niemi, O. (2012a). Fennovoiman Ydinvoimalahanke, Pyhäjoki.

Pohjantutkimusraportti. Väylä, satama ja jäähdytysvedenottorakenne. Helsinki: SITO Oy.

Forssell, T.;& Niemi, O. (2012b). Fennovoiman Ydinvoimalahanke, Pyhäjoki, Pohjatutkimusraportti. Laitosalue. Helsinki: SITO.

Geobotnia Oy. (2008). Hanhikivi Areal Ground Investigation; Pyhäjoki and Raahe. Helsinki: Fennovoima Oy.

Huusko, K. (2013). Pyhäjoen Hanhikiven laitospaikka, Laitospaikan kehittäminen 2009-

2013. Helsinki: Fennovoima Oy.

Jakobsson, K. (2014). Ydinvoimalaitoksen alustava rakenttavuusselvitys: Kalliotekninen selvitys. Helsinkin: Sipti infra consulting.

Kakkuri, J.;& Chen, R. (1992). On horizontal crustal strain in Finland. Bulletin Geodesique, 12-20.

Kauppi, I. (2014). FH1 wind direction distribution. RE-03-0000039. Fennovoima.

28 (130)

June 1, 2015


Kijko, A. (2004). Estimation of the maximum earthquake magnitude, Mmax. Pure appl

geophys, 161, 1-27.

Kokkinen, J. (2015a). Hanhikiven Ydinvoimalaitoksen satama, jäähdytysveden ottorakenne ja meriväylä, Suunnitelmaselostus v.2.5. Helsinki: SITO.

Kokkinen, J. (2015b). Hanhikiven Ydinvoimalaitoksen jäähdytysveden purkurakenne, Suunnitelmaselostus v. 1.6. Helsinki: SITO.

Korja, A., & Kosonen, E. (2014). Evaluating seismic hazard for the Hanhikivi nuclear

power plant site. Seismotectonic framework and seismic source area models in

Fennoscandia, Northern Europe. FH1-00000255. Institute of Seismology. University of Helsinki.

Korsman, K. (1988). Tectono-metamorphic evolution of the Raahe-Ladoga zone, Introduction. (G. S. Finland, Ed.) Bulletin 343, 5-6.

Kuivamäki, A. W. (2011). Simon Karsikon ja Pyhäjoen Hanhikiven mahdollisilla

ydinvoimalan sijoitusalueilla ja niiden ympäristöissä mahdollisesti tapahtuneiden

nuorten (postglasiaalisten) kallioliikuntojen selvittäminen. Vaihe 1: Kallioperäaineistojen kokoaminen, tulkinta ja analy. Espoo: Geological Survey of Finland.

Kuivamäki, A., Wennerström, M., Vaarma, M., Härmä, P., Nyholm, T., Turunen, et al.

(2011). Simon Karsikon ja Pyhäjoen Hanhikiven mahdollisilla ydinvoimalan

sijoitusalueilla ja niiden ympäristössä mahdollisesti tapahtuneiden nuorten

(postglasiaalisten) kallioliikuntojen selvittäminen. Vaihe 1 - Kallioperäaineistojen kokoaminen, tulkinta ja analy. Etelä-Suomen yksikkö. Geologian tutkimuskeskus.

Kuivamäki, A., Vuorela, P., & Paananen, M. (1998). Indications of postglacial and recent

bedrock movements in Finland and Russian Karelia. Espoo: Geological Survey of

Finland.

Kujansuu, R. (1964). Nuorista siirroksista Lapissa (Summary: Recent faulting in

Lapland). Geology, 16, 30-36.

Kukkonen, I., Olesen, O., & Ask, M. (2010). Postglacial faults in Fennoscandia: targets for scientific drilling. GFF, 132(1), 71-81.

Kähkönen, Y. (2005). Svecofennian supracrustal rocks. In: Lehtinen M., Nurmi, P.A:,

Rämö, O.T. (Eds.) The precambrian Geology of Finland - Key to the Evolution of the Fennoscandian Shield. Amsterdam: Elsevier B.V.

Kärkkäinen, N.;& Lohva, J. (2008). Ydinvoimalaitoksen mahdollisten sijoituspaikkojen

kallioperän malmipotentiaaliselvitys. Espoo: Geological Survey of Finland.

Kärnä, T. (2008). Ice actions on cooling water intakes. Helsinki: Karna Research and Consulting.

Lahtinen, R. (2012). Main geological features of Fennoscandia. In: Eilu, P. (Ed), Mineral

deposits and metallogeny of Fennoscandia. Geological Survey of Finland, Special Paper 53, 13-18.

29 (130)

June 1, 2015


Lahtinen, R., Korja, A., & Nironen, M. (2005). Paleoproterozoic tectonic evolution. In M.

Lehtinen, P. Nurmi, O. Rämö, & (Eds.), The precambrian Geology of Finland - Key to the evolution of the Fennoscandian Shield (pp. 481-532). Amsterdam: Elsevier B.V.

Lehtinen, M., Nurmi, P., & Rämö, O. (2005). The precambrian geology of Finland - Key to the evolution of the Fennoscandian Shield. Amsterdam: Elsevier.

Lund, B. (2005). Effects of Deglaciation on the Crustal Stress Field and Implications for Endglacial Faulting:. Stockholm: Swedish Nuclear Fuel and Waste Management Co.

Lund, B., Schmidt, P., & Hieronymus, C. (2009). Stress evolution and fault stability

during the Weichselian glacial cycle. Stockholm: Swedish Nuclear Fuel and Waste Management Co.

Lunkka, J. P., Johansson, P., Saarnisto, M., & Sallasmaa, O. (2004). Glaciation of

Finland. In Ehlers, J. and Gibbard, P.L. (eds.): Quaternary glaciations - extent and chronology. PART 1: Europe. Amsterdam: Elsevier.

Luoma, R.;& Nuutilainen, O. (2008). Hanhikivi Areal Ground Investigation: Pyhäjoki and

Raahe. Oulu: Geobotnia Oy.

Luukas, J., Kousa, J., Nikander, J., & Ruotsalainen, A. (2004). Raahe-Laatokka -

vyöhykkeen luoteisosan kallioperä Länsi-Suomessa. In J. Kousa, & J. (Luukas, Vihannin

ympäristön kallioperä- ja malmitutkimukset vuosina 1992-2003. Unpublished report M10.4/2004/2 (pp. 6-37). Espoo: Geological Survey of Finland.

Melcher, G. (1953). The Conglomerate of Hanhikivensaari, Pyhäjoki, Finland. Bulletin de

la Commission geologique de Finlande 159.

Mäkelä, A., & Hyvärinen, O. (2014). Return levels of trombs and downbursts in Pyhäjoki region. RE-03-0000023. Finnish Meteorological Institute.

Mäkitie, H. (1999). Structural analysis and metamorphism of Paleaproterozoic

metapelites in the Seinäjoki-Ilmajoki area, western Finland. Bull. Geol. Soc. Finland 71, 305-328.

Mäntyniemi, P. (2008). Assessment of seismicity and design earthquake to possible

nuclear power plant sites in Pyhäjoki, Ruotsinpyhtää and Simo according to the YVL 2.6 regulations. RE-02-0000007. Institute of Seismology, University of Helsinki.

Nironen, M. (2005). Proterozoic orogenic granitoid rocks. Teoksessa M. Lehtinen;N.

P.A;& O. (. Rämö, The Precambrian Geology of Finland - Key to the Evolution of the Fennoscandian Shield (ss. 443-480). Amsterdam: Elsevier B.V.

Ojala, A., Palmu, J.-P., Åberg, A., Åberg, S., & Virkki, H. (2013). Development of an

ancient shoreline database to reconstruct the Litorina Sea maximum extension and the

highest shoreline of the Baltic Sea basin in Finland. Bulletin of the Geological Society of Finland 2013, 85(2), 127-144.

Ojala, V. J., Kuivamäki, A., & Vuorela, P. (2004). Postglacial deformation of bedrock in Finland. Nuclear Waste Disposal Research. Espoo: Geological Survey of Finland.

30 (130)

June 1, 2015


Pezeskh, S., Zandieh, A., & Tavakoli, B. (2011). Hybrid empirical ground-motion

prediction equations for Eastern North America using NGA Models and updated Seismological parameters. Bull. Seism. Soc. Am., 101, 1859-2870.

Pirinen, P., Simola, H., Aalto, J., Kaukoranta, J.-P., Karlsson, P., & Ruuhela, R. (2012).

Tilastoja Suomen ilmastosta 1981-2010. Raportteja 2012:1. Finnish Meteorological

Institute.

Pitkäranta, R. (2012). 2C: Pohjavesiselvitys vesilupaa varten, COM-00004204. Helsinki:

SITO.

Pohjatekniikka Oy. (2009). Soil Investigation Hanhikivi, Pyhäjoki. Helsinki: Fennovoima Oy.

Poutanen, M., Häkli, P., Kallio, U., Nyberg, S., Rouhiainen, P., & Saaranen, V. (2011).

Geodeettisten havaintoaineistojen kokoaminen, käsittely ja analyysi Simon-Pyhäjoen

alueelta mahdollisen ydinvoimalan sijoitusalueen liikuntojen selvittämiseksi. Helsinki:

Fennovoima Oy.

Putkinen, N.;& Valpola, S. (2011). Pyhäjoen Hanhikiven nuoret siirrokset. Länsi-Suomen yksikkö. Kokkola: Geologian tutkimuskeskus.

Pöyry Oy. (2009). Paikanvalintaa edeltävät vesistöselvitykset ydinvoimalaitoshankkeessa. Jääolosuhteet Simossa ja Pyhäjoella. Helsinki: Pöyry Oy.

Rantataro, J., Kaskela, A., & Alvi, K. (2011). Simosel - Simo Merigeologia/vaihe 1 v.2011. Espoo: Geological Survey of Finland.

Rantataro, J., Kaskela, A., Alanen, U., & Hämäläinen, J. (2012). Akustis-seisminen

luotaustutkimus. Espoo: Geologian tutkimuskeskus 1.10.2012.

Rämö, S., & Jokinen, J. (2009). Nuclear Power Plant: Simo Pyhäjoki Ruotsinpyhtää. Helsinki: Pohjatekniikka Oy.

Rämö, S.;& Pulkka, J. (2009). Rock investigation report: Core drilling, video logging,

seismic sounding and rock stress measurement Hanhikivi, Pyhöjoki. Helsinki: Pohjatekniikka Oy.

Rämö, S.;& Pulkka, J. (2009). Soil Investigation Hanhikivi, Pyhäjoki. Helsinki:

Pohjatekniikka Oy.

Saari, J. (2015). Evaluating seismic hazard for the Hanhikivi nuclear power plant site. Summary report. NE-4469. ÅF-Consult.

Saari, J., & Malm, M. (2010). Complementary calculations related to seismic hazard of

the potential nuclear power plant sites in Pyhäjoki and Simo. RE-02-0000006. ÅF-Consult.

Saari, J., Heikkinen, P., Varpasuo, P., Malm, M., Turunen, E., Karkkulainen, K., et al.

(2009). Estimation of Seismic Hazard in Territory of Finland. RE-02-0002602. ÅF-Consult.

Saari, J., Lund, B., Malm, M., P, M., Oinonen, K., Tiira, T., et al. (2015). Evaluating

seismic hazard for the Hanhikivi nuclear power plant site. Seismological characteristics

31 (130)

June 1, 2015


of the seismic source areas, attenuation of seismic signal and probabilistic seismic

hazard analysis. FH1-00000565. ÅF-Consult. Institute of seismology, University of Helsinki. University of Uppsala.

Salli, I. (1964). The structure and stratigraphy of the Ylivieska-Himanka schist area, Finland. Bull. Comm. Geol. Finlande 211.

Strand, K. (2002). Volcanogenic and sedimentary rocks within Svedofennian Domain,

Ylivieska, western Finland - an example of Plaeoproterozoic intra-arc basin fill.

Teoksessa W. Alterman;& P. (. Corcoran, Precambrian Sedimentary Environments,

Modern approach to Ancient Depositional Systems (ss. 339-350). Spec. Publ. 33, Internat. Ass. Sedimentologists.

Sutinen, R.;Kosonen, E.;& Lund, B. (2014). Glaciation cycles in the Quaternary period.

Teoksessa A. Korja;& E. (. Kosonen, Evaluating seismic hazard for the Hanhikivi nuclear

power plant site. PART 2, Seismotectonic framework and seismic source area models in

Fennoscandia, nothern Europe (ss. 89-100). Helsinki: University of Helsinki, Institute of Seismology.

Thorson, R. (2000). Glacial tectonics; a deeper perspective. In: Stewart, I.S., Sauber,

J., Rose, J. (Eds.), Glacio-seismotectonics; icesheets, crustal deformation and seismicity, Vol.19. Teoksessa Quaternary Science Reviews (ss. 1391-1398).

Tilastokeskus. (2013). Väestömäärä Hanhikiven niemen ympäristössä 31.12.2012. Tiedonanto 10.12.2013 Miia Huomo. Helsinki: Tilastokeskus.

Vaasjoki, M. K. (2005). Overview. In Lehtinen, M., Nurmi, P.A., Rämö, O.T (Eds.). The

Precambrian Geology of Finland - Key to the Evolution of the Fennoscandian Shield.

Amsterdam: Elsevier B.V.

Vaasjoki, M., Korsman, K., & Koistinen, T. (2005). Overview. In Lehtinen, M., Nurmi,

P.A., Rämö, O.T (Eds.). The Precambrian Geology of Finland - Key to the Evolution of the Fennoscandian Shield. Amsterdam: Elsevier B.V.

Valjus, T. (2008). Geofysikaaliset kallionpintaselvitykset Ruotsinpyhtäällä, Simossa ja Pyhäjoella. Espoo: Geological Survey of Finland.

Valjus, T. (2008). Geofysikaaliset kallionpintaselvitykset Ruotsinpyhtäällä, Simossa ja

Pyhäjoella. Etelä-Suomen yksikkö. Espoo: Geologian tutkimuskeskus.

Wannäs, K. (1989). Seismic stratigraphy and tectonic development of the upper

proterozoic to lower paleozoic of the Bothinian Bay, Baltic Sea. Stockholm contirbutions in Geology, 40(3), 83-168.

VitusLab (Research and consultancy on ocean and climate). (2012). Hydrodynamical, wave and sediment modelling of Hanhikivi 2011. Denmark.

Wu, P., Johnston, P., & Lambeck, K. (1999). Postglacial cycles in the Quarternary

period. In A. Korja, & E. (. Kosonen, Evaluating seismic hazards for the Hanhikici

nuclear power plant site. PART 2, Seismotectonic framework and seismic source area

models in Fennoscandia, norther Europe (pp. 89-100). Helsinki: University of Helsinki, Institute of Seismology.

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1.2 Overview of the plant layout and building arrangements

Hanhikivi-1 NPP's layout and building arrangements are designed to enhance the

nuclear and radiation safety of the unit. The layout designing also considers and

optimizes the Unit's constructability, operational aspects and costs. Some of the main principles adopted in building arrangements and general layout design are:

– Safety trains are separated from each other either by distance and/or by

structural means.

– Security and physical protection aspects are considered in building disposition

and arrangements.

– The locations of systems handling radioactive mediums as well as locations of

personnel's normal facilities, such as canteens, and transport routes are

defined and designed so as to minimize occupational radiation exposures in

compliance with the ALARA principle. This has been implemented through

zone classification (dividing unit's premises into controlled or non-controlled

areas according to potential radioactive concentration and/or radiation levels)

and controlled area access and exit arrangements.

– The constructability has been taken into account in the building arrangements

and layout design, e.g. by considering modular approach at site construction.

The Nuclear Island consists of the following main buildings and structures:

– Reactor Building 10UJA;

– Steam Cell 10UJE;

– Safety Building 10UKD;

– Pumping Station of the Essential Consumers 10UQB;

– Control Building 10UCB;

– Unit Diesel generator Station Building of Nuclear Island with Diesel Fuel

Intermediate Storage 12UBN

– Standby Diesel generator Station Buildings of Emergency Power Supply

System (EPSS) with Diesel Fuel Intermediate Storage 11-12UBS;

– Auxiliary Building 10UKA;

– Building for Handling and Storage of Solid Radioactive Waste 10UKT;

– Nuclear Service Building with Utility Rooms in Controlled Access Area 10UKC;

– Storage Building for Fresh Nuclear Fuel and Transportation-and-Handling

Equipment 10UKF;

– Ventilation stack 10UKH.

Turbine Island includes the main buildings and structures:

– Turbine Building 10UMA;

– Water Preparation Building 10UGB;

– Chemical Water Purification Auxiliary tanks 10UGD;

– Normal Operation Power Supply Building 10UBA;

– Unit Diesel generator Station Building of Turbine Island with Diesel Fuel

Intermediate Storage 11UBN;

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– Refrigerator Building 10USJ.

Turbine building is oriented so that possible turbine missile impacts to Nuclear Island

are avoided.

Seawater cooling is arranged via the following structures:

– Water Intake Structure (10UPC);

– Supply Tunnel (10UPN);

– Supply Tunnel of the Essential Consumers (10UPP);

– Discharge Tunnel (10UQN);

– Discharge Tunnel of the Essential Consumers (11UQP);

– Water Outlet Structure (10UQQ);

– Standby Discharge Tunnel of the Essential Consumers (12UQP);

– Standby Water Outlet Structure (10UQU);

– Pumping Station of the Essential Consumers (10UQB);

– Pumping Station of the Turbine Building Consumers (10UQA).

Cooling water for the unit is taken from the west shore of the Hanhikivi headland

through one water intake structure (10UPC), and it is routed to pumping stations of

turbine condensers (10UQA) and essential consumers (10UQB) via separate rock

tunnels (10UPN and 10UPP, respectively). The service water is routed once through the

condensers of consumers and returned to the Gulf of Bothnia through discharge tunnels (10UQN and 11UQP ) and common outlet structure (10UQQ).

A schematic NPP Site Plan and a general diagram of layout arrangements are presented

in figures 1.2.1.1 and 1.2.1.2 below. The location and purpose of the unit's most important buildings are shortly described after the diagrams.

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Figure 1.2.1.1 NPP Site Plan.

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Figure 1.2.1.2 NPP General Layout arrangements.

1 – Reactor Building; 2 – Safety Building; 3 – Auxiliary Building; 4 – Control Building; 5

– Nuclear Service Building with Utility Rooms in Controlled Access Area; 6 – Steam Cell;

7 – Building for handling and storage of solid radioactive waste; 8 – Pumping Station of

the Essential Consumers; 9 – Standby Diesel generator Station Buildings of EPSS with

Diesel Fuel Intermediate Storage; 10 – Storage Building for Fresh Nuclear Fuel and

Transportation-and-Handling Equipment; 11 – Turbine Building; 12 – Normal Operation

Power Supply Building; 13 – Water Preparation Building; 14 – Pumping Station of the

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Turbine Building Consumers; 15 – Refrigerators Building; 16 – Building for 400 kV

metal clad gas-insulated switchgear

1.2.1 Description of buildings

Reactor Building (10UJA)

The Reactor Building contains the reactor, nuclear steam supply system, systems for

emergency cooldown, systems for fuel handling and storing as well as some auxiliary

systems for these systems. Spent fuel storage pool is located in the reactor building, and refueling operations are performed inside the building.

Structurally the Reactor Building consists of inner, primary containment and outer,

secondary containment. The primary containment is designed to withstand the pressure

of a design basis accident. It is made of pre-stressed concrete with steel liner on the internal surface. The outer containment is designed of standard reinforced concrete.

A minor vacuum is kept inside the inner containment. Vacuum ventilation system is

equipped with iodine and aerosol filters. The purified air is discharged through the NPP

ventilation stack.

Steam Cell (10UJE)

The Steam Cell is adjacent to the Reactor Building (10UJA), next to the Control Building

(10UCB). The Reactor Building (10UJA) and Steam Cell (10UJE) are located on one

foundation slab and represent one structure. The Steam Cell is separated from the

underground and overground parts of the Control building (10UCB) by an aseismic

joint.

The overpressure protection systems of the secondary circuit are located in the steam

cell as is the emergency feedwater system with water storage tanks. Equipment and pipelines of these systems are divided into four independent safety trains.

Safety Building (10UKD)

The Safety Building is adjacent to the Reactor Building (10UJA), Auxiliary Building

(10UKA), Essential Service Water Building (10UQB), and Building for handling and

storage of solid radioactive waste (10UKT). It is separated from them by aseismic joints.

Equipment and pipelines of Emergency Core Cooling Systems (JNG and JND), Spray

System (JMN), Intermediate Cooling Circuit for Important Consumers (KAA),

Emergency Boron Injection System (JDH) and vacuum ventilation systems for the

annular space of the Reactor Building are located in the safety building. Equipment and pipelines of these systems are divided into four independent safety trains.

Power supply of the safety system equipment installed in the building comes from four isolated trains of the Emergency Power Supply System.

Auxiliary Building (10UKA)

The Auxiliary Building is located in the central part of the plant. It is adjacent to the

Control building (10UCB), Reactor building (10UJA), Nuclear Service Building (10UKC),

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Safety building (10UKD) and Building for Handling and Storage of Solid Radioactive

Waste (10UKT). The Auxiliary Building is separated from these by aseismic joints.

The equipment of the primary circuit auxiliary systems, treatment and storing systems

of radioactive waters, e.g. Purification System of Water in Fuel Pool and Borated Water

Storage Tanks (FAL) and Drain Water Treatment System (KPF), and controlled area

ventilation systems are located in the Auxiliary Building. Additionally, the liquid

radioactive waste handling and storage systems as well as solidification plant are also in the building.

Since the building contains systems handling potentially radioactive liquids and gases,

the building is part of the controlled area and all exhaust air from controlled area is led

into the environment through the Ventilation Stack (10UKH) located on the Auxiliary Building roof.

Control Building (10UCB)

The Control Building is surrounded by the Reactor Building (10UJA), Steam Cell

(10UJE), Auxiliary Building (10UKA) and Nuclear Service Building (10UKC), and is separated from those by aseismic joints.

The Control Building is designed for the main instrumentation and control systems of

the plant. The building contains the Main Control Room and systems for measuring,

controlling and ensuring the plant's safe operation both under normal operating

conditions and in the event of an accident. The electric power supply systems for the I&C systems and other Nuclear Island consumers are also located in the building

The Standby Diesel generator Station Buildings of Emergency Power Supply System (EPSS)

with Diesel Fuel Intermediate Storage (11-12UBS)

The Standby Diesel generator Station Buildings of Emergency Power Supply System

(EPSS SDGS) with Diesel Fuel Intermediate Storages are separated by a considerable

distance from each other, and the two buildings are located on opposite sides of the

Pumping Station of the Essential Consumers (10UQB) and the Safety Building (10UKD).

Design and process solutions of buildings 11UBS and 12UBS are consistent with each other.

EPSS SDGS buildings contain standby diesel generator sets and electrical equipment of

Emergency Power Supply System, to provide power supply for safety system consumers under the unit's blackout conditions.

Both buildings are divided into two parts with reinforced concrete walls; both parts

house two fully autonomous equipment that are independent of each other’s safety

trains. The equipment of the first and second safety trains are located in building 11UBS, while equipment of the third and fourth safety trains are in building 12USB.

Building for handling and storage of solid radioactive waste (10UKT)

The Building for handling and storage of solid radioactive waste is adjacent to the

Auxiliary Building (10UKA), Safety Building (10UKD) and Pumping Station of the

Essential Consumers (10UQB) and is separated from them by aseismic joints.

All solid low-and intermediate level radioactive wastes, including solidified liquid wastes

from the Auxiliary Building, are treated and stored in the Building for handling and

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storage of solid radioactive waste. The treatment of the wastes includes e.g. actions

such as sorting, packing, compacting and characterizing of the solid wastes.

Storage Building for Fresh Nuclear Fuel and Transportation-and-Handling Equipment (10UKF)

Storage Building for Fresh Nuclear Fuel and Transportation-and-Handling Equipment is

located next to the buildings of EPSS SDGS (11UBS) and Building for handling and storage of solid radioactive waste (10UKT).

Besides being storage for the fresh fuel prior to its utilization, the 10UKF contains

storage facilities for transportation and handling equipment used in the refueling and

fuel inspections.

Nuclear Service Building with Utility Rooms in Controlled Access Area (10UKC)

The Nuclear Service Building with Utility Rooms in Controlled Access Area is adjacent to

the Control Building (10UCB) and Auxiliary Building (10UKA). The Nuclear Service Building (10UKC) is separated from the abutting buildings by aseismic joints.

The access to and exit from the unit's controlled area is arranged through the sanitary

posts (so-called shoe boundaries) of the Nuclear Service Building (10UKC). In addition

to connecting clean (uncontrolled) area and controlled area, the Nuclear Service

Building with Utility Rooms in Controlled Access Area is also an entry point to clean area

premises of the Nuclear Island (which in perspective of radioactivity or radiation levels are classified as uncontrolled areas) such as the Control Building.

Pumping Station of the Essential Consumers (10UQB)

Pumping Station of the Essential Consumers is adjacent to the Safety Building (10UKD)

and Building for handling and storage of solid radioactive waste (10UKT).

The Pumping Station of the Essential Consumers can be structurally divided into three

parts: suction chamber, the pumping building itself and discharge chamber. Suction

chamber is used as a header. From this header, sea cooling water is directed to trains

of the pumping building. The equipment of mechanical purification system and service

water system for essential consumers is placed inside the pumping building. The

building is divided into four independent safety trains, which have no other connection to each other but seawater.

Unit Diesel generator Station Building of Nuclear Island with Diesel Fuel Intermediate Storage(12UBN)

Separately standing building of the Unit Diesel Generator Station (UDGS) of Nuclear

Island (12UBN) is located opposite the Auxiliary building (10UKA).

The Nuclear Island diesel-generator sets and electrical equipment of the auxiliary

Reliable Power Supply System for the safety-related essential consumers and

consumers of systems designed for design extension conditions are to be located in the building.

The building is divided with reinforced concrete walls into two parts, which house the

equipment of two completely independent auxiliary Reliable Power Supply Systems.

Adjacent to each part of the building, there is a diesel fuel storage, where cylindrical steel tanks are located for diesel fuel storage, a separate tank for each cell.

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Turbine building (10UMA)

The Turbine Building is adjacent to the Pumping Station of the Turbine Building

Consumers (10UQA), Water Preparation Building (10UGB) and Normal Operation Power

Supply Building (10UBA). The building is separated from the Steam Cell (10UJE) by the fire-fighting access road.

The Turbine building contains the turbine and generator with their supporting systems

such as preheating and oil systems, as well as the main and auxiliary systems of the

steam-water cycle. A volumetric-planning solution of the turbine building is defined by

the structure and dimensions of the turbine generator, secondary systems and equipment layout and selection of de-airing-feedwater plant equipment.

1.3 General Plant description

1.3.1 Brief description of VVER design evolution

The design of Hanhikivi-1 NPP design is a result of an evolutionary development process

of the VVER-type Pressurized Water Reactor (PWR) technology. The operating

experience with VVER-type plants amounts to about 1 300 reactor-years (among them

are the VVER-440 power plants at the Loviisa NPP in Finland and the VVER-1000s in the

Czech Republic, Bulgaria and China). Nowadays, the advanced VVER-1000 NPPs (the

AES-91 project) are operated in China (two units), six VVER-1000/1200 units are under construction in Russia and nine in other countries.

A great deal of experience on engineering, manufacturing of equipment, construction

and operation has been gained since the development of the first commercial VVER-

1000 design V-320. In the VVER-1000/V-320 design, many primary technical and

safety features were incorporated and confirmed, and structural materials and the design philosophy for every design aspect were selected.

The AES-91 power plant design with the VVER-1000/V-428 reactor became an

evolutionary continuation from the VVER-1000/V-320. The AES-91 plant was originally

developed to be constructed in Finland. Thus, in addition to the Russian (Soviet) based

operating experiences and safety requirements, Finnish regulatory requirements and

operating experiences from the two Loviisa VVER-440 power plant units were reflected

in the design. The experience in project development and implementation in

cooperation with a wide circle of leading European and American companies became a good basis to refine the methods of international collaboration in Russian NPP projects.

In 1997, the AES-91 design was adopted for construction in China. In the course of

planning the implementation of the project, Chinese regulatory documents were taken

into account. Between 1995 and 2005, IAEA carried out twenty expert reviews of the

design. The reviews confirmed that the design meets the general worldwide safety

standards and practices. The comments and recommendations of the IAEA experts were

taken into account during project development. Since 2007, the two units of the

Tianwan NPP site have been in reliable commercial operation without any significant interruptions.

Simultaneously with the AES-91 plant, the same design institute (Atomproekt) was

developing the VVER-640 NPP design, a medium capacity passive safety NPP. The

objective of the development work was to design an economically attractive medium

capacity plant with passive safety features. However, due to changes in market trends,

the design work was temporarily frozen. Even so, primary solutions for NPP passive

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safety systems were found in the development of that design and experimentally

tested. These were later implemented in the AES-2006 design.

To participate in the bidding process for a new NPP unit in Finland, the development of

AES-91 started in the 1990s. The new design version was called the “AES-91/99

project” with the V-466 reactor unit. The AES-91/99 project was the next evolutionary

move in the VVER-1000 design development, with 60 years’ operating lifetime for the reactor. The design met the Russian and international safety regulations in effect then.

The safety concept of the design was based on primary use of active safety systems to

manage design basis accidents and to prevent their progress into more severe ones,

and a combined use of passive and active systems for severe accident management.

This design paid particular attention to, among other design features, protection against

the impact of large aircraft, use of the passive containment heat removal system,

provisions against common cause failures and realistic assessment of the probabilities

of human caused errors. These safety measures and improvements ensured compliance

against the requirements of EUR (European Utility Requirements).

In 2005, launch of a program to mass construct new NPP units in Russia and abroad

indicated the necessity to develop a further modernized design, titled "the AES 2006".

The main goal of that project was to enhance economic efficiency of the reactor unit

without substantial design changes and at the same time to improve safety. The design

modifications enabled an increase in the reactor's capacity, reaching up to 3200 MW

(thermal), and introduction of additional passive safety systems for design extension

conditions. The best operation practices from VVER-1000 NPP in Russia and abroad,

lessons from construction, commissioning and operation of China NPPs, the experience

of Finnish regulator reviews during the bidding process in 2003, as well as design and

experimental tests made for the passive safety systems in the VVER-640 development project have been taken into account in the AES-2006 design.

The Hanhikivi-1 NPP project is based on the AES-2006 design that is currently being

implemented on the Leningrad NPP-2 (LNPP-2) site. During the construction licensing

phase of the project, the design was comprehensively examined by the Russian regulatory body, Rostechnadzor.

Moreover, the design regularly participates internationally in bidding processes for NPP

projects, so its engineering solutions (especially its safety aspects) are repeatedly

reviewed by international experts. Among the most significant bidding processes was

for the construction of two new units on the Temelin NPP site in the Czech Republic in the course of which the design was reviewed against the IAEA and EUR requirements.

The evolution of the VVER design is illustrated in Figure 1.3.1.1.

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Figure 1.3.1.1 VVER technology evolution

1.3.2 Overview of the plant operation

Hanhikivi-1 NPP unit is VVER-type pressurized water reactor (PWR). The primary circuit

is composed of:

– a reactor pressure vessel containing a reactor core,

– four main circulating loops,

– four reactor coolant pump sets (RCPS),

– primary tube side of each of the four steam generators (SG), and

– a pressurizer (PRZ).

The primary circuit is designed to meet the principles of defense-in-depth pertaining to

the prevention of the dispersion of radioactive substances and related to the assurance of integrity. Primary circuit is located entirely inside the primary containment (UJA).

The secondary circuit is nonradioactive. It is composed of steam generating parts of the

SG, main steam pipelines and a turbine plant which includes turbine and generator,

condensate pumps, low-pressure regenerative heaters, main condensate systems,

deaerator, feedwater systems including feed water pumps and high pressure regenerative heaters.

The primary coolant is routed through the reactor core where it is heated by the energy

of fission reactions in the nuclear fuel. The heated water then flows to the tube side of

steam generators (SG) through four parallel circulating loops where it gives up its

energy to generate secondary circuit steam. The primary coolant from the SG is

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returned to the reactor for reheating via cold leg pipes of main circulating loops.

Circulation in the loops is effected by four reactor coolant pumps (RCP). Changes in

pressure and coolant volume, caused by temperature changes, are compensated by

means of a pressurizer. In case of excessive increase of primary circuit pressure (during

normal operating disturbances), steam from the PRZ is discharged into the pressurizer's

relief tank through safety valves. The relief tank is cooled by intermediate cooling circuit.

Steam, generated in the secondary side of the steam generators, is supplied to the

steam turbine via main steam lines and through the control valves. Steam flows

through the high pressure (HP) and the intermediate pressure (IP) turbine cylinders

(which are in a common casing) and three parallel low pressure cylinders, where its

thermal energy is converted into mechanical rotation energy. The generator coupled directly to the turbine shaft converts mechanical energy to electrical energy.

After passing through the turbine, steam flows to the condenser where it condenses

when cooled down with circulating cooling water.

The condensate is pumped from the condenser to the gland steam condenser and

condensate purification system. After purification, the condensate is supplied to

dearator through the low pressure heaters LPH1/LPH2, LPH-3 and LPH-4. The main

condensate is deaerated and heated in the deaerator by means of a counterflow of the

supplied condensate and steam.

From the deaerator, the feedwater is then supplied through high pressure heaters to

the steam generators by feedwater electric pumps. When passing through regenerative

heat exchangers, the condensate and feedwater are heated by steam, which is supplied

from the turbine extractions.

Circulating cooling water for the main turbine condensers and auxiliary cooling water

intended for heat removal from the Intermediate Cooling Circuit System for

Unimportant Consumers (PGB) is supplied from the pumping station of the Turbine

building consumers. Cooling water to the pumping station is taken from the sea via

supply tunnel. After the condensers, the heated water is returned to the Gulf of Bothnia through the discharge tunnel.

Main technical characteristics and operating parameters are given in Table 1.3.2.1, and

Figure 1.3.2.1 presents a schematic diagram of unit's main operating systems as well as safety systems.

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Table 1.3.2.1 The main characteristics of the Hanhikivi-1 NPP

№ Characteristic Value

1 Plant's design lifetime, [y] 60

2 Power, [MW]:

- electric power (net) at the cooling water temperature of: (4

ºС)

- Nuclear Steam Supply System (NSSS) thermal power:

1 175

3 212

3 Availability factor, [%] >90

4 Auxiliary power consumption (based on expenditure for water

recycling and site expenditure), [%]

7,1

5 Unplanned automatic reactor scram [1/y] <1

6 Maximum fuel assembly average burnup, [MWd/kgU] 60

7 Duration of fuel campaign, [y] 4

8 Period of refueling, [month] 12

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Figure 1.3.2.1 Unit schematic diagram

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1 – Cooling Water System for Important Consumers (PE) pump, 2 – Heat exchangers of Intermediate Cooling Circuit for Important

Consumers (KAA), 3 – Intermediate Cooling Circuit for Important Consumers (KAA) pump, 4 – Heat exchanger of the Spent Fuel Pool

Cooling System (FAK), 5 – Low-Pressure Safety Injection System pump (JNG1), 6 – High-Pressure Safety Injection (JND) System Pump,

7 – Emergency Feedwater System (LAR/LAS) pump, 8 – Demineralized water storage tank, 9 – Spent Fuel Pool Cooling Pump (FAK), 10

– Borated Water Storage Tank (JNK), 11 – Emergency Boron Injection System (JDH), 12 – Storage tank of chemical reagents, 13 –

Chemical reagents supply pump, 14 – Containment Spray system (JMN) pump, 15 – Filter, 16 – Deaerator of the Makeup and boron

control system, 17 – Pump of the Makeup and boron control system, 18 – Ventilation stack, 19 – Controlled leaks pump, 20 – Controlled

leaks tank, 21 – Secondary containment, 22 – Steam generator, 23 – Special water treatment plant, 24 – After-cooler, 25 – Spent Fuel

Pool, 26 – Bubbler tank, 27 – Regenerative heat exchanger of the Makeup and boron control system, 28 – Reactor, 29 – Reactor Coolant

Pump, 30 – Molten Core Catcher (JMR), 31 – Emergency Core Cooling System Sump and Reactor Water Storage Tank, 32 – Alkalis

emergency reserve tank, 33 – Main Steam Isolation Valve, safety and relief valves unit, 34 – Primary Containment, 35 – Pressurizer, 36

– Hydroaccumulators of the Emergency Core Cooling System, passive part (JNG2), 37 – Passive Heat Removal System tank, 38 –

Condenser of the Passive Containment Heat Removal System (JMP), 39 – Spray system (JMN), 40 – Passive hydrogen recombiner, 41 –

High-pressure heaters, 42 – Electric-driven Auxiliary Feedwater Pump, 43 – Deaerator, 44 – Electric-driven feed water pump, 45 –

Condenser, 46 – Low-pressure heaters, 47 – Condensate pumps of the first stage, 48 – Gland seal steam condenser, 49 – Main

condensate treatment, 50 – Superheater, 51 – Main Circulating Water (PA) pumps, 52 – Auxiliary Circulating Water (PC) pump, 53 –

Machine hall consumers, 54 – Standby step down transformer, 55 – Generator, 56 – Low-pressure part of the turbine, 57 –

Intermediate-pressure part of the turbine, 58 – High-pressure part of the turbine, 59 – Hydraulic hitch, 60 – Heat exchangers of ECCS, 61 - Exhaust ventilation system of Safety building, trains, 62 - Condensate pumps of the second stage

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1.3.2.1 Overview of the Nuclear Steam Supply System (NSSS)

Reactor coolant circuit

The reactor coolant circuit is designed to receive thermal power generated through

controlled nuclear reaction of fuel fission, remove heat from the core and generate steam in the secondary sides of the steam generators.

The reactor coolant circuit includes water-moderated, water-cooled power reactor and

four circulating loops consisting of SG, reactor coolant pump unit and pipelines, as well

as pressurizing system connected to one of the circulating loops. In the following

paragraphs, some key characteristics of these main components are briefly presented.

More detailed description of the functions of reactor coolant system are described in

chapter 1.5.1.

Reactor, core and reactor pressure vessel

The reactor of AES-2006 is water-moderated water-cooled power reactor with thermal

power of 3200 MW. The coolant and moderator of the reactor is water containing boric

acid solution. The concentration of boric acid in the coolant is dependent on power operation period.

The reactor core is composed of 163 fuel assemblies (FA). The fuel assemblies are

hexagonal in cross-section; the fuel is slightly enriched uranium dioxide. Control rods are included in 121 of the fuel assemblies.

The reactor vessel is a cylindrical pressure vessel made of high-strength heatproof alloyed steel; the inner surface of the pressure vessel is clad with stainless steel.

The inlet and outlet nozzles for the primary circuit connections are in two rows. The

reactor coolant inlet nozzles for the Emergency Core Cooling System enable injecting

water to the lower and upper parts of the reactor. There are no nozzles located below the top of the core level.

Reactor internals, such as core barrel and core baffle, together with a water gap between the reactor and pressure vessel limit the neutron flux impact on the vessel.

In addition, there are surveillance specimens located in the radiation area to control the

condition of reactor vessel metal. The radiation surveillance specimens are located

directly on the inside reactor vessel wall, opposite the core, as this location gives a representative approximation on the neutron flux impact to the reactor vessel.

To monitor parameters such as power distribution and temperature in the core, there are in-core instrumentation devices installed in some of the fuel assemblies.

Steam generators

The steam generators (SG) are horizontal single body heat exchangers with submerged

heat-exchange surface made of horizontal tubes. In nominal conditions, steam

generation capacity of one SG is 1600 t/h and pressure of generated steam is 7.0 MPa.

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To enhance operational reliability of SGs, the following design features are provided:

– sparse-corridor tube layout in a heat bundle;

– flushing devices (detachable nozzles on the coolant header transition rings) for

sludge removal from the SG body during preventive maintenance.

The advantages of sparse-corridor pipe layout in SG are the following:

– circulation velocity in the pipe bundle is increased;

– reduced possibility of tube space clogging by peeled-off sludge;

– tube space access for inspection is facilitated;

– water reserve in SG is enlarged;

– space under the pipe bundle is enlarged to facilitate sludge removal;

– primary coolant circuit header stress state is enhanced.

Reactor coolant pumps

Reactor coolant pumps (RCP) are vertical, single-stage pump consisting of a body and a

removable part. The pump body is spherical. The pump is supplied with welded

adapters with rust-preventing facing on the inside surface. The pumps are powered by vertical asynchronous motors.

The pump and motor shafts are connected by a rigid coupling to form a single rigid

shaft in three supports (bearings). One bearing is located in the pump (radial bearing),

two bearings (radial and radial-axial bearing) in the motor. The radial-axial bearing is

located in a motor upper crosspiece. The motor and pump bearings are water lubricated.

Pressurizer

The pressurizer is a vertical vessel with electrical heaters. It is intended for limitation of

pressure deviation during power operation and under transient conditions, protection of

the equipment and primary pipelines from overpressure, and for primary pressure

maintenance during heat-up and pressure decrease during NSSS cooldown.

The pressurizer body is made of carbon steel with inner surfaces coated with austenitic stainless facing.

The pressurizer's total volume is 79 m3, and under nominal conditions, its water volume is 55 m3.

1.3.2.2 Overview of the Engineered Safety Features

A distinctive feature of the Hanhikivi-1 AES-2006 is an optimal combination of active

and passive safety features. The primary safety systems are mainly active ones,

designed to provide high reliability to execute the safety functions in compliance with appropriate failure tolerance, redundancy, independency and diversity requirements.

The passive safety features provide diversity to execute safety functions in design extension conditions.

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The safety systems of Hanhikivi-1 unit include e.g. the following active and passive

systems:

– Low-Pressure Safety Injection System (JNG1);

– Emergency Core Cooling System (ECCS), passive part (JNG2);

– High-Pressure Safety Injection System (JND);

– Emergency Boron Injection System (JDH);

– Borated Water Storage System (JNK);

– Residual Heat Removal System (JNA);

– Containment Spray System (JMN);

– Emergency Gas Removal System (KTP);

– Emergency Feed water System (LAR/LAS);

– Containment Hydrogen Removal System (JMT);

– Containment Passive Heat Removal System (JMP);

– System of Passive Heat Removal through Steam Generators (JNB);

– Corium Localization System (Core Catcher) (JMR);

– Emergency Pressure Relief System (JEG30)

In addition, there are several safety systems, which actuate and support the

abovementioned systems. The safety systems are described in more detail in section 1.5.2.

In the following paragraphs, the containment, structure providing one of the main safety features of the NPP, is discussed.

Containment

Containment is designed to:

– limit the release of radioactive substances into the environment during normal

operation, anticipated operational occurrences and accidents,

– protect the plant against natural and human induced external events, and

– provide a protective biological shield during normal operation, anticipated

operational occurrences and accidents.

In order to achieve these targets, the containment is to have a double structure. It

consists of inner leak-tight containment (performing release confinement functions) and additional outer protective containment.

The main advantages of double containment are:

– reduction of emergency releases and level of possible radiation effect on

population and personnel;

– sharing of functions—outer containment serves as physical protection against

external impacts, inner containment ensures tightness under all NPP operation

conditions, including emergency conditions;

– availability of space (annulus) between containments, which makes it possible

to:

maintain vacuum inside the space and localize leakages;

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separation of penetrations, which improves operating conditions and

facilitates inspection;

clearly divide safety system trains, locate instrumentation sensors,

pipelines, cables;

maintain stable temperature conditions inside annulus, which leads to

reduction of thermal stresses inside the inner containment.

Outer containment is made of non-prestressed reinforced concrete. Containment

consists of a cylindrical part and hemispherical dome on which tanks of the passive

containment and steam generator heat removal systems are located. Structural

solutions and reinforcement of the outer containment is defined by external extreme loads, for example impact of aircraft crash.

Besides providing protection for reactor coolant system and other internal structures

located inside the primary containment from external hazard impact, the secondary

containment envelopes the primary containment and so creates an additional barrier

and an annular space between the two containments; potential primary containment leakages are confined in this space.

Inner containment is a prestressed reinforced concrete structure, consisting of a

cylindrical part and a dome. Inner surface of containment is lined with carbon steel to ensure tightness.

Prestressed containment and the foundation plate form together a tight containment

with strength designed to withstand maximum pressures from design basis accidents,

including large break loss of coolant accident (LOCA). Internal structures of the reactor

building are separated from the inner containment cylinder wall by a joint to provide

fast balancing of pressure between compartments in case of accident.

The containment provides also storage facilities for spent fuel. The spent fuel

assemblies are located in storage racks in the spent fuel pool adjacent to the reactor pit.

Containment diagram is presented in figure 1.3.2.2.1.

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Figure 1.3.2.2.1 Containment diagram

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1.4 General Safety Design principles

1.4.1 Fundamental safety objectives

The fundamental objectives of NPP safety are to minimize detrimental (radiation)

effects of the NPP on the personnel, population and environment and to establish and

maintain the nuclear Safety As High As Reasonably Achievable (SAHARA). These

objectives shall be achieved at each stage of the NPP life cycle, starting from selection

of the NPP location and ending with NPP’s decommissioning.

The following requirements shall be fulfilled to ensure radiation safety:

– The exposures of plant personnel and individuals of population as well as the

number of exposed people shall be kept As Low As Reasonable Achievable

(ALARA) taking into account economic and social factors.

– The personnel radiation exposure does not exceed the prescribed dose limits.

– Exposure doses to population do not exceed the limits and criteria presented

in the Finnish legislation (Government Decree 717/2013).

The safety philosophy aims at accident prevention, as well as mitigation of

consequences of accidents, which may lead to significant releases of radioactive

substances into the environment. The objective is to reduce the probability of such

events occurring, and to diminish their consequences.

The safety objectives are achieved in design by means of the following:

– application of proven techniques;

– application of Defense-in-Depth (DiD) concept;

– conservatism during design and justification of chosen design solutions;

– safety culture.

1.4.2 Safety Policy

The design solutions for safety-related systems were selected and justified using

deterministic and probabilistic safety analysis. Many years' experience of VVER-1000

NPP construction and operation was utilized in the design development. The Hanhikivi-1

design is an evolution of the VVER technology; in it systems and equipment have been

improved and safety enhanced by using up-to-date passive systems for management in

emergency conditions, including severe accidents, while retaining the advantages of

VVER technology. The design is also developed in order to take into account the higher

requirements for NPP stability against external impacts, considering events at

Fukushima NPP.

The main documents, establishing the approaches to NPP safety on the territory of

Finland, are the Government Decree on the safety of Nuclear Power Plants (717/2013),

Security in the Use of Nuclear Energy (734/2008), Emergency Response Arrangements

of a Nuclear Power Plant (716/2013) and Safety of Disposal of Nuclear Waste

(736/2008) as well as YVL guides developed by the Finnish regulatory authority STUK.

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In addition, the requirements of Russian regulatory documents in regards nuclear power

plants design are taken into consideration in the course of Hanhikivi NPP design, as well as international and other national documents, such as:

– IAEA guidelines and safety norms;

– European Utility Requirements for design of new generation nuclear power

plants with LWR type reactors (European Utility Requirements (EUR));

– WENRA requirements (West European Nuclear Regulators Association);

– Applicable US NRC requirements.

1.4.3 Defense-in-Depth concept

The Defense-in-Depth (DiD) concept is based on the application of physical barriers

preventing ionizing radiation and radioactive substances propagation into the

environment, and on the in-depth structured system of technical and organizational

measures on barriers protection, as well as of direct protection of personnel, population

and environment. Therefore, the basic safety assurance task implemented in the design

is to provide physical barriers integrity under different conditions. The physical barriers

of VVER-type NPPs include the following:

– fuel matrix;

– fuel element cladding;

– reactor coolant circuit boundary;

– containment structure.

In order to protect the integrity of physical barriers and to implement a functional

defense-in-depth principle, provisions are made in the NPP design for systems and means intended to ensure three fundamental safety functions:

– control of reactivity (to limit the reactor power, to shut the reactor down and

to maintain it in the subcritical state),

– removal of heat from the reactor core and from the spent fuel pools, and

– confinement of radioactive substances (to keep radioactive substances inside

the nuclear power plant).

Provisions are made in the NPP design for means and procedures, which ensure

fulfillment of safety functions depending on the initiating event. These means and

procedures are elaborated taking into account independence of the functional defense-

in-depth levels. The defense-in-depth levels structure (in accordance with the Government Decree on NPP safety 717/2013) has the following five defense levels:

1. Prevention in advance. At the first level, monitoring is performed to ensure

that the power plant operation is reliable and deviations from normal

operation conditions are insignificant.

2. Control under anticipated operational occurrences (DBC2). At the second level

the control is performed under anticipated operational occurrences and

accidents progression is prevented.

3. Emergency conditions management. At the third level, accidents management

is performed to prevent the release of radioactive substances into the

environment by means of physical barriers and to prevent serious damages to

fuel. At the third level of the defense-in-depth, design basis accidents and

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design extension conditions are considered separately. Thus the third level of

defense-in-depth is divided into levels 3a and 3b. At 3b level, provisions are

made for means ensuring safety functions under conditions in which safety

systems of 3a level cannot perform their functions due to multiple failures

caused by a common cause, external hazards or other complicated emergency

sequences.

4. Limitation of releases in case of severe accidents. At the fourth level, the

consequences of severe accidents are mitigated in order to limit the release of

radioactive substances into the environment;

5. Mitigation of consequences. At the fifth level, measures are taken to mitigate

the effects of radioactive radiation on the population and the environment in

cases when significant releases of radioactive substances have occurred.

In case of failures or emergency situations, the objective of the technical means

provided at different DiD levels shall be to transfer the NPP into one of the following states:

– Return to normal operation mode;

– Transfer to controlled state;

– Transfer to safe state.

Controlled state is defined as a stable NPP condition in which fission chain reaction has

been stopped and residual heat is removed from fuel.

Safe (shutdown) state refers to a state where the reactor has been shut down and is

non-pressurized, and removal of its decay heat has been secured (Government Decree 717/2013).

Final NPP state and requirements to reach this state are determined in the design,

depending on the initiating events, which define the mode and state of both the reactor

plant and the whole NPP.

The engineered safety means for certain safety functions fulfillment include all the

necessary components, systems and structures, such as process, I&C, power supply,

HVAC, cooling devices, etc. For each defense-in-depth level, the corresponding technical

and organizational measures are provided to ensure the main safety objectives.

The functional Defense-in-Depth levels, their objectives and means to achieve the

objectives are presented in the Table 1.4.3.1.

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Table 1.4.3.1 Levels, objectives and means of the defense-in-depth for NPP design

Levels Objectives Means to achieve objectives

Level 1 Prevention of deviating

from normal operating

conditions

Development of self-protection properties of the

design, quality assurance for equipment and

systems, improvement of ageing management

for NPP elements and safety culture by e.g. following means of:

- design elaboration using a conservative

approach, providing for a developed intrinsic

self-protection ability of the reactor plant;

- assurance of the required quality of NPP systems (components) and works performed;

- NPP operation in accordance with the

requirements of the regulatory documents,

design technical specifications and operating

instructions;

- maintenance of safety-related systems,

keeping them in good condition by detecting

defects in time, taking preventive measures,

replacing equipment with expired lifetime and

establishing an effective system of work and inspection records;

- selection of personnel and assurance that NPP

personnel is duly qualified to take actions during

normal operation and in case of anticipated

operational occurrences including pre-

emergency (pre-accident) situations and accidents; development of safety culture.

Level 2 Control of anticipated

operational occurrences

and accidents prevention

Limitation of impact on the physical barriers and

their integrity by means of:

- detection of anticipated operational

occurrences and compensation of such

deviations from normal operation conditions as

well as prevention of situations escalating into accidents;

- transfer to controlled state in case of

necessity.

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Levels Objectives Means to achieve objectives

Level 3 Control of accidents

Application of engineered safety features and

procedures in accidents management with the

purpose of:

- transferring and maintaining RP in the controlled state;

- transferring and maintaining RP in the safe state.

3a Under design basis accidents of classes 1 and 2

(Design Basis Conditions 3 and 4):

- reactor trip systems;

- safety systems;

- emergency procedures.

3b Under Design Extension Conditions: using

special means and procedures, which ensure recovery of critical safety functions.

Level 4 Severe accidents

management aimed at

limitation of radioactive releases

Severe accidents management using specially

designed means:

- limitation of the effect and protection against

destruction of the leak-tight enclosure and maintaining its operability;

- assurance of subcriticality and heat removal

from the destroyed core (fuel) and keeping radioactive substances within the preset limits.

Level 5 Mitigation of

radiological

consequences under

significant radioactive

releases

Emergency off-site measures aimed at

mitigation of consequences of radioactive

products released into the environment by

means of preparation and implementation (if

necessary) of emergency measures plans at NPP site and beyond its boundaries.

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1.4.4 Event categories

Postulated initiating events are identified as events that represent a challenge to

fundamental safety functions and may lead to anticipated operational occurrences or

accidents. In accordance with the level of possible negative consequences and

probability of occurrence, the list of conditions to be considered in the design is divided

into the following categories. Different analysis principles are applied and different acceptance criteria are imposed on each category:

– DBC1 – normal operation conditions;

– DBC2 – anticipated operational occurrences;

– DBC3 – design-basis accidents of class 1;

– DBC4 – design-basis accidents of class 2;

– DEC – design extension conditions;

– SA – severe accidents.

As stated in the Finnish Government Decree (717/2013), the limit for the annual dose

of an individual in the population, arising from the normal operation (DBC1) of a nuclear

power plant is 0.1 mSv.

Events of category DBC2 are anticipated operational occurrences and are initiated by

design initiating events, which can be expected to occur once in one hundred years of

operation or more often. In the worst case, they can lead to reactor shutdown, after

which the power plant operation can be resumed when the cause of the failure is

eliminated. The annual dose limit of an individual in the population arising as the result

of the anticipated operational occurrence is 0.1 mSv in compliance with Finnish Government Decree (717/2013).

Events of category DBC3 are design-basis accidents of class 1 and are initiated by

design initiating events, which can be expected to occur less than once in one hundred

years of operation, but at least once in one thousand years of operation. The dose limit

of DBC 3 events is 1 mSv for a representative person of the most highly exposed population group.

Events of category DBC4 are design-basis accidents of class 2 and are initiated by

design initiating events, which can be expected to occur less than once in one thousand

years of operation. These events are postulated because together with a number of

other consequences they result in the release of a large amount of radioactive

materials. The dose limit of DBC 4 events is 5 mSv for a representative person of the

most highly exposed population group.

In addition to the design provisions with regard to performance of deterministic

analyses of the design conditions, analyses of design extension conditions DEC and severe accidents (SA) are conducted.

The design extension conditions (DECs) are divided into three event groups:

– DEC A: anticipated operational occurrences and Class 1 postulated accidents

that involve a common cause failure in the system designed for coping with

the event concerned;

– DEC B: combinations of failures selected on the basis of a probabilistic risk

assessment; and

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– DEC C: rare external events that are unlikely to occur but nevertheless

considered possible.

The annual dose limit for the exposure received by a representative person of the

population in DECs is 20 mSv.

For severe accidents, the following criteria defined in Finnish legislation (Government Decree (717/2017) and YVL guide C.3) are implemented in the safety design:

– the frequency of reactor core damage shall not exceed 10-5 /year;

– the release of radioactive substances arising from a severe accident shall not

necessitate large scale protective measures for the population nor any long-

term restrictions on the use of extensive areas of land and water.

In order to limit the long-term effects, the limit for atmospheric releases of Cs-137 is

100 TBq. The possibility of exceeding the set limit shall be extremely small (with frequency lower than 5*10-7 /year).

The possibility of a release in the early stages of an accident requiring measures to

protect the population shall be extremely small.

1.4.5 Technical implementation and ensuring reliability of safety functions fulfillment

The high level of reliability and safety of the NPP design is ensured by the design self-

regulation and simplicity as well as using the following principles: redundancy,

independence, protection against dependent failures and common cause failures, protection against personnel errors.

Design margins, engineering safety features and self-protection properties

The following safety features and principles are implemented in Hanhikivi-1 NPP design:

– Use of inherent safety characteristics that do not rely on the control or safety

systems. Fuel and reactor core characteristics together with conditions

considered as admissible in the design provide negative feedbacks (reactivity

coefficients) in terms of fuel and coolant temperature.

– Assurance of design margins. Due to the large original coolant inventory in the

primary circuit and SG, dynamic characteristics of the reactor plant allow using

automatic parameters control within normal operation limits and creating a

margin to the activation of safety systems and corrective action ensured by

limitation functions or carried out by operator.

– Implementation of safety functions by provision of optimum combination of

passive and active engineered design features. For DEC and SA, the use of

passive safety systems and controls relying on natural phenomena (e.g.

natural circulation or gravity) or stored energy and mass (pressure and

inventory of coolant in hydro accumulators or tanks) are implemented. The

passive systems include e.g. ECCS hydro accumulator, systems of passive

heat removal from SG and from the containment, hydrogen recombiners, core

catcher, boric solution inventory tanks inside the containment, etc.

– Actuation of the safety systems either automatically by adequately validated

protection systems or passively if certain set-points or parameters are

reached. In case of failure of computer-based I&C, there is a provision for

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starting safety functions from the automation facilities based on hardwired

logic.

– Implementation of measures, which reduce or completely eliminate impact of

the operator's errors on safety. Provisions are made for measures mitigating

or preventing the impact of operator error consequences on safety, including

limitation of positive reactivity insertion rate, automatic actuation of standby

equipment, preventive protections and interlocks, emergency protection of the

reactor, automatic actuation of safety systems and temporary inhibition of

operator's intervention in the event of an accident.

Redundancy principle

Redundancy, the use of more than the minimum number of equipment or component to

accomplish a given safety function, is an important design principle in improving the

reliability of the fulfillment of safety functions and to meet the applicable failure

criterion (N+1 or N+2 criterion). The failure criterion N+2 means that a function shall

be fulfilled in case of a single failure and simultaneous repair or maintenance. The

failure criterion N+1 means that a function shall be fulfilled in case of a single failure of any system element.

Redundancy makes tolerable a failure or unavailability of at least one set of equipment

without losing the function. Redundancy can be provided by using identical or different

components.

Functional redundancy is a principle of use for fulfillment of a certain function of different systems.

The degree of and approach to redundancy shall be defined during systems design in

accordance with the regulatory requirements to safety (YVL guide B.1), as well as on

the basis of deterministic and probabilistic analyses. Different requirements in terms of

redundancy are imposed on the systems, depending on the functions to be fulfilled and on the defense-in-depth level.

Diversity principle

In order to prevent common cause failure from interfering with the fulfillment of safety

function, the reliability and independence of defense-in-depth levels is enhanced by

using different and/or similar technologies based on various operation principles in

different systems (or within one system in various trains) to perform a given safety function.

The diversity principle is applied to redundant systems and units performing one and

the same function by introducing various distinctive features in those systems or units.

In addition, the distinctive features of the hardware can be different principles of

operation, different physical variables, different operating conditions, different equipment manufacturers etc.

Independence principle

The reliability to perform a specific safety function is also enhanced by incorporating the following independence principles into the design:

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– ensuring independence of the defense-in-depth levels, so that the following

level's failure to operate is not a consequence of (or does not have the same

cause as) the failure to perform a safety function at one of the previous levels.

– ensuring independence of redundant system components;

– ensuring independence of system components from impact; e.g. an impact

shall not result in a failure of a system or a function required to mitigate the

impact;

– ensuring adequate independence between systems or components of various

safety classes.

Independence is achieved in the design of systems by providing functional and/or

physical separation, as well as application of the diversity principle.

Functional separation excludes the possibility for interaction of systems and

components, which are redundant relative to each other, as well as common cause

failures of these systems and components. For example, passive safety systems are

completely functionally separated from active safety systems and normal operation systems.

Physical separation and arrangement of plant components. To increase the

independence, especially in relation to external and internal impacts, physical

separation is applied to redundant components and systems, using the following

approaches, as far as it is feasible:

– separation using geometrical factors (distance, orientation, etc.);

– separation using barriers;

– separation using combination of the above-mentioned measures.

The design incorporates principles of physical separation of buildings, structures and NI

equipment by providing appropriate layout, distances, arrangement and physical

barriers.

Means of physical separation of safety system trains and equipment inside the

containment are designed taking into account loads of internal and external impacts

considered in the design. Protection against dependent failures in safety systems is

provided for by arranging safety system trains in different rooms. Safety system trains

are separated by physical fire resistant barriers designed to withstand loads caused by internal and external effects.

1.4.6 Autonomy

The design objectives shall include the fulfillment of the self-sufficiency criterion. The

NPP's independence on external resources, such as the ultimate heat sink and external

power supply, is ensured by the plant design. Power plant’s autonomy will be ensured

for a duration that is required to carry out actions to arrange external resources in abnormal situations.

Moreover, the NPP design is implemented to minimize the power plant susceptibility to errors or inaction by the personnel.

The following autonomy tasks are identified and implemented in the design:

– autonomy in terms of operator's actions;

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– autonomy in terms of ultimate heat sink;

– autonomy in terms of power supply systems.

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Autonomy in terms of operator's actions

To prevent transient processes, which exceed the operational limits and conditions, the

design is provided with the self-control properties (e.g. power feedback), corresponding

dynamic characteristics (e.g. thermal inertia) and other design solutions (e.g.

automation of control process) that ensure a sufficient time period for the operator for

making decisions and performing corrective actions. To exclude the operator's errors or

his/her inaction, the design is provided with automatic actuation of hardware for corrective or protective actions.

The design of NPP ensures that in case of an accident, there is no need for action by the

operator from the Main Control Room (MCR) during the first 30 minutes nor actions

outside the MCR during the first hour from the accident initiation.

Autonomy in terms of ultimate heat sink

Provisions are made in the design related to autonomy with regard to ultimate heat

sink, so that the residual heat can be removed for 72 hours. The heat shall be removed

from the reactor, from the containment and from the fuel pools in case of anticipated operational occurrences and accidents.

The 72 hours autonomy means that the systems, shall be able to perform their

functions for 72 hours. Within these 72 hours, during the first 24 hours no material

replenishments (such as filling the water or fuel tanks) is allowed. During the following

48 hours, replenishment of water and fuel inventories shall be possible to accomplish with inventories at the NPP site.

In case of rare external events (DEC C), the heat removal function shall be possible to

accomplish without material replenishments for 8 hours. After which replenishments for 72 hours shall be available at the site.

Autonomy in terms of power supply systems

The period of power plant independence from external power supply sources shall be at

least 72 hours. This requirement shall be applied to normal operation conditions, as well as to failure conditions, accidents and design extension conditions.

In case the external power supply is lost for systems requiring it, these systems shall

remain in state, which is preferable from the safety perspective. Failure in power supply

systems during normal operation conditions shall not lead to accidents. Therefore,

provisions are made in the NPP design for internal power supply systems, ensuring

functioning of safety-related systems and elements. The internal power supply system

shall provide for a 72-hour autonomy.

The capacity of internal and external power supply systems shall be sufficient for safety functions fulfillment.

Power supply systems autonomy during 72 hours entails the following:

– Storage batteries, which supply power to safety-related systems and do not

allow for interruptions in power supply, are designed to operate for 2 hours

under the highest load;

– Storage batteries of the severe accidents management systems are designed

to operate for 24 hours under the highest load;

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– The design is provided with diesel-generators that recharge the storage

batteries with the aim to supply direct current to the safety-related consumers

for 72 hours;

– Alternating current Reliable Power Supply Systems ensure fulfilment of safety

functions for 24 hours without makeup of fuel or any additional measures. The

fuel inventory required for system operation during the following 48 hours is

provided for at the NPP site;

– In case of rare external events (DEC C), fulfilment of the function of residual

heat removal and reactivity control does not depend on the external power

supply for 8 hours without application of additional means or recharging of the

direct current storage batteries. Sufficient fuel inventory and possibility to

charge the storage batteries to perform residual heat removal for 72 hours is

provided for at the industrial site.

The internal system of standby power supply is automatically actuated and connected

to ensure uninterrupted supply of electric power to the facilities performing the safety function in accordance with their operating time.

Mobile equipment is required to achieve the acceptance criteria of anticipated

operational occurrences or postulated accidents, DEC and severe accidents.

1.4.7 Internal and external hazards

Various impacts on NPP equipment and structures leading to their failure caused by

both internal and external hazards shall be considered in the design as initiating events for emergency situation occurrence and their changing for the worse.

Internal hazards

Internal hazards (such as flooding, fires, missiles) occurring in various NPP modes and

conditions are considered during the design development in order to protect safety-

related NPP systems and structures against dependent failures and common cause

failures as well as to protect operating personnel.

Protection against internal hazards is provided by compliance with requirements for

independence and physical separation as well as using additional systems and

measures. E.g. the safety systems are separated from normal operation systems inside

the safety building and subdivided into train rooms. Those rooms are designed as

independently of one another as possible, with respect to both physical separation and

process connections. Flooding or fire in one train room shall not lead to failures of equipment in neighboring train rooms.

External hazards

External natural phenomena as well as man-induced events such as rare atmospheric

phenomena, fire in the vicinity of the power plant, high and low sea levels, seismic

events, heat removal failure caused by other reasons than freezing or seismic event

consequences, airplane crash, electromagnetic phenomena, explosions or releases of

poisonous gases in the power plant area, oil spillage in the sea in the vicinity of the

power plant and unauthorized invasion of the power plant territory or of the information systems of the power plant are treated as external hazards.

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The external hazards are conditionally divided into the design external hazards and

extreme external hazards as per the degree of impact and probability of their

occurrence. The occurrence of design external hazards are considered within the

framework of design-basis conditions (DBC), while the occurrence of extreme external

hazards are considered as a rare external event categorized as DEC C events.

1.4.8 Principles of safety classification

All NPP systems, components and structures will be classified based on their function

and safety influence in accordance with the requirements of YVL B.2. The classification is done in order to determine requirements for:

– NPP systems and components design;

– manufacturing quality of NPP equipment and components;

– condition and operability checks and tests of NPP systems and equipment.

The classification will be made primarily based on a deterministic approach used to

determine a set of systems and components required to ensure safety functions and to

subdivide those systems and components into defense-in-depth levels. The NPP

structures and components are assessed according to their importance to NPP safety

and performance of safety functions. At that, requirements for their resistance, integrity

and leak tightness are taken into account. General classification approach anticipates the following:

– classification of systems based on their functions;

– classification of structures and components based on their resistance, integrity

and leak tightness required to prevent radioactive substances spread and to

perform their functions.

All NPP systems are grouped into safety class 2 or 3 or to EYT class (non-nuclear

safety) based on their significance for the reliability of safety functions from the viewpoint of management of initiating events.

The structures and components are grouped into one of three safety classes (1, 2 or 3)

or to EYT class based on the structural strength, integrity and leak tightness required of them to prevent the spreading of radioactive substances.

Besides, all NPP systems, structures and components are attributed to one of three

seismic stability categories (S1, S2A, S2B) characterizing requirements for seismic loads tolerance.

1.4.9 Radiation safety

With regard to radiation safety, the design of the Hanhikivi-1 NPP is based on:

– requirements of the Finnish Government Decrees 716/2013 and 717/2013,

Radiation Act 592/1991, Radiation Decree 1512/1991, YVL Guides and other

effective national Finnish standards on radiation safety;

– modern principles of safety assurance and NPP safety philosophy developed by

the world nuclear community, set in IAEA safety norms, publications of

International Nuclear Safety Advisory Group (INSAG), ICRP recommendations,

EUR requirements, US NRC requirements;

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– set of well-developed technical solutions proven in practice and experience in

the design of similar new generation facilities of enhanced safety;

– verified calculation methods and codes, developed methodology of safety

analysis, reliable database.

The design provides technical and organizational measures to ensure that the basic

personnel, population and environment radiation safety principles are observed as required by the Radiation Act 592/1991:

– principle of justifiability—benefit from practical use shall exceed the harm it

causes;

– the ALARA (As Low As Reasonably Achievable) optimization principle—

maintaining radiation exposure at the minimum practicable level, taking into

account economic and social factors;

– limitation principle—non-exceedance of maximum permissible exposure limits

established by Radiation Decree 1512/1991.

Under normal operation conditions and anticipated operational occurrences, limitation of

radiation exposure doses to personnel and population as well as radioactive materials

releases into environment are secured to be below the preset limits at a reasonably

achievable socially and economically justified low level, proven by operational experiences from VVER-type reactors all over the world (ALARA principle).

The following means are implemented to ensure personnel radiation safety in NPP

operation (in accordance with YVL Guide C.2):

– subdivision of NPP process building and structures into Supervised area and

Controlled area and limitation of the time of personnel presence in the

Controlled area;

– shielding of radiation sources to reduce the radiation down to acceptable

levels;

– remote control of mechanisms;

– radiation monitoring and dosimetry control; etc.

The following design features of Hanhikivi-1 NPP reduce personnel's radiation exposure

in maintenance tasks:

– selection of appropriate corrosion-resistant materials;

– limitation of cobalt content in structures which will be in contact with the

primary coolant and be exposed to significant neutron fluxes;

– optimization of primary water chemistry to reduce the activity of corrosion

product deposits;

– use of hot operational testing of primary circuit surfaces to enhance the

quality of protective oxide film on the surface of circuit equipment;

– improvement of equipment maintenance aimed at mechanization (e.g. remote

control of SG heat-exchange tubes), and reduction of maintenance man-

hours.

The releases and discharges from the NPP and radiation levels in the environment

arising from the operation of the NPP will be kept as low as reasonably achievable.

Liquid discharges and gas-aerosol releases into the atmosphere are limited in normal

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operating conditions and in case of deviations from normal operation due to the

following:

– high level of equipment leak tightness in the primary circuit ;

– effective purification of discharged water from radioactive contamination and

arrangement of non-radioactive discharge waters’ discharge via monitoring

tanks after obligatory radio-chemical monitoring which ensures specific

activity concentration meets the set criteria for permissible discharge;

– effective purification of exhaust air from radioactive gases and aerosols,

coming from the controlled access area and process vents;

– reduction of discharged water volume by means of its reuse for auxiliaries in

the power plant life cycle.

In accident conditions, doses to personnel and population in the environment are

limited in accordance with the criteria set in the Finnish Radiation Decree 1512/1991,

Government Decree 717/2013 and YVL guide C.3. The public exposure limits of each

event categories are also presented in section 1.4.4 "Event categories" above.

1.5 NPP general systems description

1.5.1 Reactor Coolant System

1.5.1.1 Reactor coolant pumps

Functions

The reactor coolant pump set is intended for creation of primary coolant circulation and

heat removal from the reactor core. RCP set has an additional function of providing

coolant circulation under coastdown in various conditions with loss of power supply, which allows for a smooth transfer to the natural circulation condition.

Design description

RCP set design is selected on the basis of the experience accumulated during

elaboration, manufacture, tests on special rigs both of separate sub-units and RP pump

sets as a whole, experience of commissioning work and operation of pumps at VVER nuclear power plants.

RCP is a vertical, single-stage pump consisting of a body and a removable part.

The pump body, made of steel, is spherical. The pump is supplied with welded steel adapters with rust-preventing facing on the inside surface.

A vertical asynchronous motor is used as a drive.

The pump and motor shafts are connected by a rigid coupling to form a single rigid

shaft in three supports (bearings). One bearing is located in the pump (radial bearing),

two bearings (radial and radial-axial bearing) in the motor. The radial-axial bearing is

located in a motor upper crosspiece. The motor and pump bearings are water lubricated.

Location of Equipment

RCPs are located inside the containment (UJA).

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System operation

Under NO and AOO, RCP set ensures coolant circulation in the NSSS reactor coolant

circuit, which includes four parallel connected circulation loops. Each loop contains one RCP set. The main mode of operation is continuous parallel operation in the circuit.

1.5.1.2 Steam generators

Functions

Steam generator is designed for generation of dry saturated steam from the heat removal from the primary coolant.

Design description

SG is a horizontal single body heat exchanger with submerged heat-exchange surface made of horizontally located pipes.

For the purpose of SGs reliability enhancement in the course of their operation the SG design with the following basic design features is used:

– sparse-corridor pipe layout in a heat bundle;

– flushing devices (detachable nozzles on the coolant header transition rings) for

sludge removal from the SG body during preventive maintenance.

The advantages of sparse-corridor pipe layout in SG are the following:

– circulation velocity in the pipe bundle is increased;

– reduced possibility of tube space clogging by peeled off sludge;

– tube space access for inspection is facilitated;

– water reserve in SG is enlarged;

– space under the pipe bundle is enlarged to facilitate the sludge removal;

– primary coolant circuit header stress state is enhanced.


Steam generators are located inside the containment (UJA).

Design principles

The following technical, design, process, and operational requirements are assumed as the basis for design and operation of steam generators:

– ensuring the required thermal-technical and separation characteristics

reliably—generation of steam of required quantity and quality;

– ensuring of primary coolant cooling to the required level of temperatures in all

design conditions reliably;

– ensuring feedwater supply to SG from the Emergency Feed Water System

(LAR/LAS) along a separate line;

– ensuring primary coolant cooling under natural circulation by selection of the

scheme and lay-out of the reactor coolant circuit together with the steam

generator;

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– ensuring operability, reliability and safety of the steam generator and its

components under loads occurring under design conditions during the whole

service life;

– utilizing the experience of operation of steam generators of this type and

consideration of the factors increasing reliability and convenience of SG

operation;

– providing convenience and processibility of assembly during manufacture of a

nuclear power plant, minimum amount of welding operations under assembly

conditions;

– providing convenience and simplicity of steam generator maintenance

(accessibility to the steam generator, secondary circuit, and primary circuit

collectors for examination and repair during preventive maintenance);

– ensuring a possibility to perform a check of welded joints and base metal

using modern diagnostic tools, including a possibility of inspection and

plugging of heat exchanging tubes during operation.

1.5.1.3 Reactor Coolant Piping

Functions

Main coolant pipeline interconnects the reactor, the steam generator and the reactor

coolant pumps forming the reactor coolant circuit, and it is intended to create

circulation of coolant from the reactor into steam generator and backwards.

Design description

The Reactor Coolant Pipeline comprises four independent circulation loops. The

circulation loops are arranged in pairs on diametrically opposite sides of the reactor.

Each loop has three sections of tubes. The section between the reactor outlet nozzle

and the steam generator inlet collector presents the hot leg. The section between the

steam generator outlet collector and the inlet nozzle of RCP set (suction) and the

section between the outlet nozzle of RCP set (pressure head) and the reactor inlet

nozzle present the cold leg. The inside diameter is chosen in order to assure acceptable

coolant velocity (of the order of 10 m/s) and Reactor Coolant Pipeline pressure loss, within the design range of coolant flow rate.

The hot leg of loop No. 4 is connected with the pressurizer (surge line).

The cold leg of loop No. 3 is connected with the pressurizer (injection line).

The parameters characterizing the normal functioning of the system are the coolant

temperature in the Reactor Coolant Pipeline hot and cold legs, as well as the difference between these temperatures.

1.5.1.4 Pressurizing and steam discharge system

Functions

The pressurizing system in the primary circuit is intended for limitation of pressure

deviation during power operation and under transient conditions, protection of the

equipment and primary pipelines from overpressure, and for primary pressure maintenance during heat-up and pressure decrease during NSSS cooldown.

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Pressurizing system in the primary circuit provides fulfilment of the following functions:

– creation and maintaining of pressure in the primary circuit of the reactor plant

during start-up;

– limitation of pressure deviations in the primary circuit, caused by change of

temperature conditions during operation of the reactor plant providing boiling

margin of primary coolant in any of hot legs of the loops;

– reactor coolant pressure boundary overpressure protection;

– primary pressure decrease at the assigned rate, in the mode of the reactor

plant cooldown, including together with Makeup and Boron Control System

(KBA);

– collecting leaks of steam through PRZ PORV and blow-offs from PRZ during

start-up;

– primary pressure decrease together with Emergency Boron Injection System

(JDH) in the mode of PRISE leaks.

Design description

Pressurizing system is composed of:

– system for creation and maintenance of pressure, including two components:

1) subsystem of creation and pressure increase, to be used when

pressure is decreasing, consisting of a set of PRZ TEH;

2) subsystem of pressure decrease, to be used when pressure is

increasing, consisting of injection lines, valves and injection device

of pressurizer;

– primary overpressure protection system.

Criterion of fulfilment of functions assigned to the primary overpressure protection

system is limitation of pressure within reactor coolant pressure boundaries not above

the permissible limit, determined by the conditions of structural integrity of the given

physical barrier on the basis of rules and regulations for DBAs and DECs.

Pressurizing and steam discharge system involves:

– pressurizer with a set of electrical heaters and fasteners;

– surge line;

– injection lines with valves;

– steam discharge line with pilot-operated relief valves;

– discharge line of steam-gas mixture with valves and a throttle device;

– relief tank.

System operation

Functioning of the system for maintaining pressure in the primary circuit under normal operating conditions is summed up in the following:

– compensation for volume variation of primary coolant due to overflow of the

primary coolant from RCS to PRZ and in reverse order;

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– pressure maintenance during the first period of heating-up process is provided

for by creation of PRZ gas blankets and further a steam one, transition to

which is provided by periodic opening and closing of valves on the line of

steam-gas mixture discharge into the relief tank;

– primary pressure control during reactor operation at power.

1.5.2 Safety Systems

1.5.2.1 Spray System (JMN)

Functions

The Spray System is designed to perform the following functions:

– reduction of pressure and temperature in the containment during primary or

secondary coolant leak accidents;

– sedimentation of fission products in the containment atmosphere during

primary coolant leak accidents;

– control of chemical composition of water in the sump tank by adding chemicals

for long time retention of iodine and prevention of corrosion during primary

coolant leak accidents;

– providing reserve for the Fuel Pool Cooling System (FAK);

– filling of the reactor internals inspection shaft during refueling operations.

Design description

The system is composed by four similar trains independent of each other. Each train

includes pump, valves, pipelines, spray collector with spray nozzles, chemicals tank.

The suction pipelines of each train are connected through the Low Pressure Safety

Injection System pipelines to their containment storage sump tank of borated water of

low concentration. The suction lines of the spray pumps are connected to the pressure

lines of the chemicals injection pumps, which deliver chemicals from the alkali supply

tanks.

The schematic diagram of the system is given in figure 1.5.2.1.


One part of the Spray System equipment, including pipelines and valves, is located

inside the containment. The other part of this equipment (pumps, tanks, valves,

pipelines) is located in separate rooms within the Safety Building, isolated from each other by fire-resistant physical barriers.

System Operation

In unit power operation, the Spray System does not operate and is in a standby mode

(in standby mode for emergency use in case of loss-of-coolant accident).

Together with the Residual Heat Removal System (JNA) and Low-Pressure Safety

Injection System (JNG1), the Spray System can fulfill the function of providing

redundancy for the Fuel Pool Cooling System (FAK) in case of failure of one of FAK system's trains.

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In the event of a primary or secondary coolant leak accident, the pumps of Spray

System are started by an emergency signal, i.e. pressure rise under containment. At

the same time, valves on the lines supplying media into the containment atmosphere open as well as a valve on the line for medium intake from sump tanks.

In the process of injection, the system reduces pressure in the containment by

removing heat from the containment atmosphere with injected solution.

In case of loss of normal offsite power supply (de-energizing) signal at any stage of

accident, the Spray System pumps are actuated according to the step-by-step start-up program of the diesel generators.

Chemicals are supplied provided that the Spray System is actuated and one of the

signals is given: activity rise inside the containment or no signal indicating that

damaged SG has been isolated. At the same time, chemicals are supplied to the suction side of the spray pumps from the chemicals storage tank.

On in-containment pressure drop signal, gate valves on the line for supply to the

header with spray nozzles close automatically, and gate valves in the recirculation line open.

The pumps remain in operation and operate in the recirculation line.

1.5.2.2 Passive Containment Heat Removal System (JMP)

Functions

Passive Containment Heat Removal System (JMP) reduces and maintains pressure

within design limits inside the containment and removes heat released into the

containment during SA and moves it to the ultimate heat sink.

Design description

The Passive Containment Heat Removal system consists of four independent natural circulation trains.

Heat is removed from the containment by means of steam condensation in heat

exchangers-condensers from which it is transferred to emergency heat removal tanks

by natural circulation of coolant. Heat is removed from the emergency heat removal

tanks to the ultimate heat sink by means of water evaporation in the tanks within, at

least for the first 72 hours after accident. Thermal capacity of the system is chosen to

ensure pressure reduction and keeping the pressure within design limits inside the

containment during SA. To ensure operability of the system after 72 hours, it is

expected that on-site water reserves will be used.

Location of equipment

Heat exchangers-condensers and part of the system's pipelines are located in the dome

part of the Reactor Building (UJA). The other part of pipelines and emergency heat

removal tanks are arranged outside the containment in separate rooms of the outer structure of containment (UJB), isolated from each other by fire resistant barriers.

System operation

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During operation of the unit under normal operating conditions and AOOs, the Passive

Containment Heat Removal system is in a standby condition (ready for a loss of coolant accident).

System will start up without any signals from I&C, in case of any pressure and

temperature increase in the containment. The system serves as a diverse passive

system to the Spray System (JMN).

During DEC and SA, the Passive Containment Heat Removal system is used to maintain

in-containment pressure below the design level without operator intervention for at

least 72 hours. After that it is to use on-site water reserves (for makeup of emergency heat removal tanks with cooling water).

1.5.2.3 Corium Localization System (JMR)

Functions

The Corium localization System (JMR) is designed to retain corium and solid fragments

of degraded core, reactor vessel and reactor internals as well as concrete protections in

the event of a severe accident accompanied by core degradation and reactor pressure

vessel failure. Corium is localized and cooled within the under-reactor space in the concrete cavity for an unlimited period of time.

Design description

The system JMR consists of the core catcher, active and passive channels for filling core

catcher cavity with water, passive channel (water supply valves) for supplying water

onto the melt from the core catcher cavity, instrumentation system and core catcher

cavity drainage channel. The system also includes outer and inners door of the reserve corridor, which provides access to the core catcher.


All equipment of the Corium Localization System is arranged inside the containment.

System operation

Under normal operating conditions and AOOs of the reactor plant, the core catcher is in

cold standby. It does not require special measures to put it on hot standby or make it

technically available. Under normal operating conditions, no condition monitoring of the

core catcher is provided. The same applies even in the event of DBA and SAs during in-vessel stage.

During SA with RPV failure, accompanied by the release of corium melt, functioning of

the core catcher is determined by the concept of corium localization. After the melting-

through of the reactor vessel, corium melt enters the space confined on the side and from below by the water-cooled steel walls of the core catcher vessel.

Part of the water-cooled space inside the core catcher is filled with sacrificial material,

which consists of a special composition of steel and relatively light and low-melting iron and aluminum oxides.

Сorium melt entering the core catcher from the reactor interacts with the sacrificial

material. This optimizes heat removal, levels out the differences between various

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scenarios of SAs, and ensures inversion of the metallic and oxidic components of the

melt before water is supplied onto its surface.

Inversion of the metallic and oxidic components prior to water flooding of the melt

surface provides a guarantee against steam explosions and minimizes hydrogen

release. Experimental studies have proven that water can be safely supplied onto the

surface of oxidic melt.

1.5.2.4 System of Emergency Use of Water from Reactor Internals Inspection shaft (JNB90)

Functions

System of Emergency Use of Water from Reactor Internals Inspection (JNB90) shaft is a

facility designed for severe accidents management.

The System of Emergency Use of Water from Reactor Internals Inspection Shaft is intended for:

– borated water supply from the reactor internals inspection shaft to the core

catcher during SAs, in case ofmelting of the reactor core and corium leaks

outside the reactor pressure vessel;

– alkali solution supply to sump tanks and to level below the RPV in order to

decrease the generation rate of volatile iodine forms inside the containment.

In addition, the system performs the following functions under normal operation conditions:

– filling and draining of the reactor internals inspection shaft during the

operations associated with refueling and inspection of reactor internals;

– removal of all possible leaks from the reactor cavity (core catcher room);

– maintaining the chemistry of borated water inside the reactor internals

inspection shafts.

Design description

The system is composed of four similar trains being independent of each other.

The system consists of two identical trains with common passive elements. Each train

consists of pump, valves and pipelines.

System JNB90 consists of two parts functionally:

1. Lines of water supply to the core catcher from the reactor internals inspection shaft.

2. Lines of alkali supply to the sump tanks and to level below the RPV.


One part of the system equipment, including pipelines and valves, is located inside the

containment, while another part of the system equipment—concentrated alkali storage tank, pumps, valves and pipelines—is located outside the containment.

System Operation

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The system is in standby under power operation of the unit (in ready state in case of an

accident leading of reactor core meltdown and melt releasing beyond the reactor vessel).

Under normal operation conditions, the system pipelines are used for makeup, drainage

and purification of water in the reactor internals inspection shaft. The reactor internals

inspection shaft is filled with low concentration boron acid solution, supplied from the sump tanks using the pump of Fuel Pool Cooling System (FAK) or the sprinkler pump.

In the event of emergency signals indicating the initiation of reactor core disruption, the

system pumps are activated providing alkali supply to sump tanks and to level below

the RPV. In 30 minutes after occurrence of the signal indicating the core melt releasing

beyond the reactor vessel into the core catcher, the valves open in the water supply line

from the reactor internals inspection shaft. Water from the reactor internals inspection shaft is supplied by gravity to the melt surface in the core catcher.

1.5.2.5 Containment Isolation System (JMK)

Functions

Containment Isolation System (JMK) is designed for operation under normal operating conditions, AOOs, DBAs, DECs and SAs.

Under all conditions and states of the NPP, the JMK system shall ensure the following functions:

– prevention or limitation of radioactive substances propagation;

– limitation of ionizing radiation release (biological protection);

– protection against natural and man-induced events.

Design description

At process pipelines, where containment isolation is required, in most cases two motor-

operated containment isolation valves are installed at each pipeline: one is installed

inside the containment and another outside as well as a check valve at the bypass of

one of the valves (for the pipelines containing media subject to thermal expansion). In this case, manual isolation valves are installed at the lines used only during repairs.

For the process pipelines belonging to safety systems that supply medium to the

containment during LOCA, a single motor-operated containment isolation valve is installed outside the containment and a check-valve inside the containment.

Motor-operated containment isolation valves outside the containment are also installed

at steam pipelines and feed water pipelines. As during accidents with steam generator

header rupture or SG heat exchanging tube rupture, primary and secondary circuits are

combined, and accident confinement boundary is located along the steam pipeline up to

MSIV.

The pipeline at the medium intake, from the containment emergency sump up to the

containment isolation valves inclusive, has a specific structure. It is built as a part of

containment and located inside the metal enclosure (capsule) designed for the

parameters inside the containment during accidents (static pressure to be taken into

account) and is an accident confinement boundary.

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Location of JMK system valves is described in item Design description above.

System operation

During accidents taking place after the containment isolation valves have been closed

(except for the containment isolation valves belonging to the safety systems that supply

medium into the containment during LOCA), automatic isolation of the containment is

performed for two groups of containment isolation valves that are actuated upon the

corresponding signals from the protection system. Groups are actuated based on

specific limit values of containment pressure.

The target for the period of containment cutoff (period when the quick-acting isolation

valves are closed, taking into account time for signal generation), when the pressure in

containment reaches 0.105 MPa (abs.) and 0.13 MPa (abs.), is limitation of radioactive

gases and aerosols release through the leak-tight penetrations up until full closing of

the containment isolation valves.

1.5.2.6 High Pressure Safety Injection System (JND)

Functions

The High Pressure Safety Injection System (JND) is intended for supplying boric acid

solution into the reactor coolant system during a loss-of-coolant accident at coolant

system pressure below the maximum operating pressure of JND system (7.9 MPa), if the loss of coolant exceeds the compensation capacity of the normal makeup system.

Design description

The system is composed of four similar trains being independent of each other. Each

train includes a pump, valves and pipelines.

Suction pipelines of every train are connected to the containment sump of low

concentration borated water through pipelines of the Low Pressure Safety Injection

system (JNG-1). Two of the trains are connected to one Borated Water Storage System

(JNK) tank and two others to a similar redundant tank. Pressure pipelines of the system

are connected via pipelines of ECCS hydrotanks (JNG50-80) to the upper and lower chambers of the reactor.



One part of JND system equipment, including pipelines and valves, is located inside the

containment. The other part of this equipment (pump, valves, pipelines) is located in

separate rooms within the Safety Building, isolated from each other by fire-resistant

physical barriers.

System Operation

In unit power operation, the High Pressure Safety Injection System is in standby mode

(available for loss of coolant accident occurrence). Under AOOs, the system operation is not required either.

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In case of accident, the pumps of the system start up in response to emergency process

signals indicating loss of coolant accident: primary circuit approaching saturation condition or containment pressure rising above a given limit.

In case of loss of normal offsite power supply (de-energizing signal) at any stage of

accident, the pumps of the system are actuated according to the step-by-step start-up

program of the diesel generators.

When the system pumps are started by emergency signals and the primary circuit

pressure exceeds the working pressure value of High Pressure Safety Injection System, the pumps operate through recirculation line.

When primary circuit pressure reaches the value of JND pump working pressure and

required flow occurs in pressure pipeline, valves installed on the recirculation line close

and the pumps begin to supply boron solution from tank-sump to the reactor.

In case of an accident, the system operates following a closed circuit: sump tank of JNK

system – safety systems water intake (pipeline of Low Pressure Safety Injection System

(JNG1)) – suction pipeline of JND system – JND pump – pressure main line – reactor pressure vessel (core) – leak – sump tank.

When water temperature in sump tanks exceeds 45 ºC, the heat exchangers of JNG-1

system, ensuring water cooling prior to its supply to suction of pumps of JND system, are connected.

When the JNG-1 system connection parameters in primary circuit are reached and the

emergency functions are being performed by this system, the operator may interrupt JND system operation.

1.5.2.7 Emergency Core Cooling System, passive part (JNG2)

Functions

The Emergency Core Cooling System passive part (JNG50-80) is intended for supply of

boric acid into the reactor when the primary pressure is below 5.9 MPa and in the

amount sufficient for cooling the reactor core before start-up of the low-pressure safety injection system pumps under design basis loss of coolant accident.

Design description

ECCS passive part comprises four identical and independent channels. Each channel

includes an accumulator, stop valves, safety and check valves, and pipelines. Two

channels are connected to the reactor collection chamber, and two others to the reactor

pressure chamber.

ECCS passive part pipelines are intended for connection of ECCS accumulators to the

reactor and provision of boron solution into the reactor under the conditions requiring actuation of ECCS passive part.

The binding lines of ECCS PORV are intended for connection of the ECCS accumulators with the corresponding pilot-operated relief valves of accumulators.


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ECCS passive part is located inside the containment (UJA)

System operation

Under NSSS operation at power or under AOOs, no system actuation is required. The

system is in standby mode, the quick-acting isolation gate valves are open, and each accumulator is separated from the reactor by two check valves arranged in sequence.

To prevent actuation of the system as a result of pressure decrease under planned cooling down of NSSS, the system is disconnected from the reactor by the operator.

Under accidents caused by Reactor Coolant Pipeline break, when the reactor pressure

drops below 5.9 MPa, the check valves open passively (due to pressure differential) and

the boric acid solution enters the reactor from accumulators. With level decreasing in

the accumulator to the minimum level, the quick-acting isolation valves close automatically to prevent nitrogen getting into the reactor.

1.5.2.8 Low Pressure Safety Injection System (JNG1)

Functions

The Low Pressure Safety Injection System (JNG1) is intended for boric acid solution

supply into the reactor coolant system during loss of coolant accident, including break

of main circulation circuit, when the coolant system pressure drops below operating parameters of the system.

The system performs the following functions:

– cooldown of the reactor plant at normal shutdown, AOO and DBA, provided

that the primary circuit integrity is maintained (together with Residual Heat

Removal System (JNA));

– removal of residual heat from the core in case of accidents;

– removal of residual heat from the core and maintenance of the primary circuit

temperature at no more than 60 ºС during transport and handling operations

for refueling;

– removal of residual heat from the core and maintenance of the primary circuit

temperature at no more than 70 ºС under cooldown conditions for repair;

– removal of residual heat from the fuel pool in case of failure of one train of

system Fuel Pool Cooling System (FAK), at full and emergency unloading of

the reactor core (together with one train of system FAK).

Design description

The system consists of four identical and independent trains. Each train includes a pump, heat exchangers, valves and pipelines.

When choosing the system performance characteristics, the determining factor is the

reactor plant emergency cooldown from 130 ºС down to 60 ºС at the tight primary circuit, as it requires the highest flow rate in pump sets and system heat exchangers.

Suction pipelines of each train have connections to:

– sump tanks of the containment for low concentration borated water storage.

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– supply lines from the Residual Heat Removal System (JNA).

Delivery pipeline of each train of the JNG1 system has connection to both cold and hot legs of one of the four loops within the Reactor Coolant Piping System (JEC).

At design-basis accident with primary circuit coolant leakage, water supply from each

train of the system is performed simultaneously to both cold and hot legs.

Under the reactor plant cooldown conditions and residual heat removal from the reactor

core, water supply from JNG1 system is performed to cold legs of the primary circuit system.


Equipment location

One part of the JNG system equipment, including pipelines and valves, is located inside

the containment. The other part of this equipment (pump, heat exchangers, valves,

pipelines) is located in separate rooms within the Safety Building, isolated from each

other by fire-resistant physical barriers.

System Operation

In unit power operation, the Low Pressure Safety Injection System does not function and is in standby mode (state of availability for loss of coolant accident occurrence).

Under the conditions of the unit scheduled and emergency cooldown (provided the

primary circuit integrity is maintained), the system ensures RP cooldown within the

range of primary circuit temperatures from 130 оС to 60 оС with a design cooldown rate

together with the JNA system.

At AOOs, the system functions only together with the JNA system in cases when the

reactor plant cooling down and residual heat removal from the core at a temperature below 130ºС in the primary circuit system is required.

In case of accident, the pumps of JNG1 system start up in response to emergency

process signals indicating loss of coolant accident: primary circuit approaching

saturation condition or containment pressure rising above a given limit.

In case of loss of normal offsite power supply (de-energizing signal) at any stage of

accident, the pumps of JNG1 system are actuated according to the step-by-step start-up program of the diesel generators.

In case of accident, the system operates over a closed circuit: sump tank of JNK system

- safety systems water intake (pipeline of JNG-1 system) – JNG1 pump - pressure main

line – primary circuit system pipelines - reactor pressure vessel (core) - leak - sump tank.

When water temperature in sump tanks exceeds 45 ºC, the heat exchangers of JNG-1

system, ensuring water cooling prior to its supply to suction of pumps of JNG1 system, are connected.

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1.5.2.9 Borated Water Storage System (JNK)

Functions

The Borated Water Storage System (JNK) is intended for storage of borated water with

low and high concentration required for NPP operation under all operating conditions.

The system is intended for low concentration borated water storage for the following purposes:

– emergency core cooling during loss of coolant accidents;

– injection into containment during loss of coolant accidents or break of steam

line inside the containment;

– supply of makeup water into coolant system during reactor power operation

and during reactor cooling down at unit shutdown;

– supply of borated water for the initial filling of core catcher heat exchanger at

SAs;

– supply of borated water for filling of fuel pool, reactor cavity, RI inspection

shafts, at refueling.

System is designed for high concentration borated water storage for the following purposes:

– primary circuit boron concentration control under normal operating conditions

and AOOs;

– borated water injection into the pressurizer at primary-to-secondary leakage;

– borated water injection into the reactor during anticipated transients

(operational occurrences) without scram (ATWS).


Design description

The system consists of two identical and independent trains. Each train consists of a

sump tank of low concentration borated water reserve and one tank of high

concentration borated water reserve.

The sump tanks are connected to each other by pipelines to create a common volume

for borated water storage. The sump tanks are equipped with strainers to prevent

ingress of thermal insulation fragments and other mechanical foreign matter into emergency pumps.

Four independent water intakes for safety systems (two in each sump tank) connected

to Low Pressure Safety Injection System (JNG1) trains are installed in the sump tanks.

In each high concentration borated water storage tank, two boric acid solution supply lines are installed to the suction of the Emergency Boron Injection System (JDH).


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One part of equipment of the Borated Water Storage System, including sump tanks,

pipelines and valves, is located inside the containment, while the other part of system

equipment (tanks, valves, and pipelines) is located in separate rooms within the Safety Building, with fire resistant physical barriers between them.

System Operation

In power operation, the system ensures:

– boron solution storage in case of accident and for refueling;

– makeup water supply to coolant system;

– water supply for boron control in primary circuit.

Under unit cooldown and refueling, the system ensures:

– boric solution supply to maintain shutdown concentration in the primary

circuit;

– supply of boric solution for filling of fuel pond, reactor cavity, RI inspection

shafts.

Under AOOs, the system ensures:

– in the case of signal on de-energizing, the system ensures supply of boric

solution to safety systems pumps which are switched on as per step-by-step

start-up program of the diesel generators for its operation in recirculation

lines;

– makeup water supply to the coolant system;

– water supply for boron concentration control in the primary circuit.

At design basis accidents, the system assures boric solution supply to safety systems pumps for:

– emergency core cooling at LOCA;

– injection under the containment at LOCA or steam line rupture within the

containment;

– borated water injection into pressurizer at primary-to-secondary leakages.

In case of emergency, isolating valves are closed at the pipeline supplying boric solution to the normal operation consumers.

At DECs, the system assures boric solution supply for:

– high concentration borated water injection into the reactor at anticipated

transients without scram (ATWS);

– primary circuit system makeup from sump tanks at small loss of coolant

accident and common-cause failure of HP and LP safety injection systems;

– spent fuel cooling pool makeup at common-cause failure of the cooling system

for the spent fuel cooling pool.

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At SAs, the system assures boric solution supply for initial filling of the core catcher.

1.5.2.10 Emergency Boron Injection System (JDH)

Functions

The Emergency Boron Injection System (JDH) is intended for performance of the

following functions:

– injection of boric acid solution to the pressurizer at primary-to-secondary leak

accidents;

– supply of high concentrated boric acid solution into the primary circuit system

for quick transfer of the reactor plant into subcritical condition under AOOs,

followed by a failure of the emergency reactor protection (ATWS).

Design description

The system consists of four identical and independent trains.

Each train includes a pump, valves and pipelines.

The suction pipelines of every train are connected to high concentration borated water

storage tanks. The pressure pipelines of each train are connected to the cold pipelines

of main circulating loops and to the pressurizer steam volume. Normally, closed isolation valves are installed on the pressure pipelines.


One part of the emergency boron injection equipment, pipelines and valves, are located

inside the containment, the other part of the system equipment (pumps, pipelines,

valves) in separate rooms of the Safety Building, isolated from each other by fire-resistant physical barriers.

System Operation

In unit power operation and under AOOs, the Emergency Boron Injection System is in standby mode (available for accident occurrence).

In the case of accident associated with primary-to-secondary leak, the Emergency

Boron Injection System ensures injection of boric acid solution into the PRZ steam

volume in order to reduce primary circuit pressure, and as the result, restrict the

volume of radioactivity release. This is started in response to signals of primary-to-secondary leak accident.

In the case signal of NPP de-energizing appears at any moment of accident, the pumps

of the JDH system are actuated as per the step-by-step start-up program of the diesel

generators.

Under DECs with failure of emergency protection actuation (ATWS), the system supplies

the highly concentrated boric acid solution into the primary circuit for quick transfer of the reactor plant into subcritical condition.

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In the event of blackout signal appearing at any moment of accident, the pumps of the

JDH system are put into operation as per the step-by-step start-up program of the diesel generators.

1.5.2.11 Emergency Gas Removal System (KTP)

Functions

The Emergency Gas Removal System (KTP) is intended for removal of steam-gas

mixture from the NSSS primary circuit (reactor, PRZ and SG collectors) and decrease in

the primary pressure together with PRZ PORV to reduce the consequences of DBAs and

DECs.

Use of the system under normal operation and incident conditions is allowed.

Design description

The system comprises the pipelines and the valves installed on them for stem-gas mixture removal into the relief tank or into the containment:

The removed steam-gas mixture is directed to the relief tank through the pipeline between PRZ PORV and the relief tank by opening the valves.

Removal of steam-gas mixture into the containment is made by opening the valves.

The system provides a possibility of steam-gas mixture discharge separately from the

reactor, SG collectors or PRZ, or of simultaneous discharge from the mentioned equipment in any combinations.


The Emergency Gas Removal System is located inside the containment (UJA)

System operation

Under normal operation, the pipeline routes of the system and points of its connection

to SG collectors and the reactor upper part can be actuated for primary circuit

blowdown with nitrogen to remove hydrogen that can be generated in the circuit after cooling down.

Under NSSS operation at power or under AOO, the system is in standby mode. Then all the system valves are in closed position.

During accident, the system is to remove steam-gas mixture from the NSSS primary

circuit to avoid hydrogen fire and explosion and to exclude loss of natural coolant circulation in the primary circuit.

1.5.2.12 Emergency Feedwater System (LAR/LAS)

Functions

The LAR/LAS system main function is emergency feedwater supply of steam generators under AOOs, DBAs and DECs.

Design description

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The system consists of four identical and independent trains. Each train provides

emergency feed pump, valves and pipelines.

Demineralized water storage tanks are used for demineralized water storage. If

necessary, there is a possibility of connecting demineralized water storage tanks of

Makeup Water System (LCU) to suction headers of emergency feed pumps. It is also

possible to connect tanks to suction headers of emergency makeup pumps to makeup PHRS tanks using LCU system under DEC and SA.

For prevention of CO2 entering demineralized water of LAR tanks, filter absorbers are provided at air pipelines of demineralized water storage tanks.

Pressure pipelines of each train supply water to steam generators through specially provided emergency supply branch pipe.

Water supply from each train to one steam generator is provided.


One part of the system's equipment, including pipelines and check valves, is located

inside the containment and the other part (pumps, valves, pipelines and tanks) in

separate rooms of Steam Cell UJE, separated from each other by fire-resistant physical barriers.

System operation

In NPP power operation, Emergency Feedwater System is in a standby mode (in state of readiness for accidents).

At AOOs related to stop of normal heat removal via the secondary circuit (vacuum loss

in turbine condenser, loss of normal feedwater flow etc.), emergency feed water pumps

are started, and water is supplied to the steam generator.

At NPP de-energizing, emergency feed pumps are switched on by the step-by-step

start-up program of diesel generators, and they operate at recirculation line. If the

steam generator liquid level decreases further, below the nominal level, isolation valves

are opened at the head of the emergency feed pump and water is supplied to the steam

generator.

At DBAs, emergency feed pumps are started in response to a signal of level decrease in

the steam generator below the nominal level, and water is supplied to the steam generator from the demineralized water storage tanks of system.

If a signal of de-energizing is received at any moment of accident, the system pumps

are changed over to the Emergency Power Supply System under step-by-step start-up

program of diesel generator.

In case of failure of normal feed water supply, the system ensures removal of residual heat and cooldown of the reactor plant at the steam stage of cooling down.

In accidents with break of steam and feedwater pipelines, which cannot be isolated

from the steam generators part, the emergency feed pump is stopped. Simultaneously,

isolation valves are closed at the line of feed water supply into this steam generator and

supply of feed water can be performed only to undamaged steam generators.

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At DECs, Emergency Feedwater System (LAR) is used for its actual purpose, if there is

power supply to the system equipment.

The system provides demineralized water reserve storage in tanks to provide makeup of emergency heat removal tanks PHRS using LCU system under DEC and SA.

1.5.2.13 System of Passive Heat Removal Through Steam Generators (JNB)

Functions

The System of Passive Heat Removal Through Steam Generators (JNB) is designed for

long-term residual heat removal of the core to the ultimate heat sink through the secondary circuit during DEC.

Under DEC, the system shall perform the following:

– residual heat removal and cooldown of the reactor plant under conditions of

NPP blackout;

– residual heat removal and cooldown of the reactor plant under the conditions

of total feedwater loss;

– ensure a reserve for the active safety systems in case of their failure during

primary coolant leak accidents.

Design description

The system consists of four independent natural circulation trains connected to steam and water spaces of corresponding SGs.

Each system train includes an emergency heat removal tank, sixteen sections of

emergency cooldown heat -exchangers, "large" and "small" start valves, steam and condensate pipelines, as well as containment isolation valves.

Emergency cooldown heat exchangers are designed for heat transfer from steam

generators to cooling water stored in the emergency heat removal tank. The emergency

cooldown heat exchangers are located below the water level in the lower part of the emergency heat removal tank.

Heat is removed from emergency heat removal tanks to the ultimate heat sink by

means of water evaporation in the tanks within, at least, 72 hours after accident. After

72 hours it is supposed that emergency heat removal tanks will be made up using on-site reserves.

Installed in parallel on the downcomer line of each train of the System Of Passive Heat

Removal Through SGs, upstream of SG, are the "large" and "small" start valves. The

start valves provide automatic connection to appropriate cooldown mode. The starting

valves are closed in standby condition.

Containment isolation valves are designed for automatic removal of safety train of the

System Of Passive Heat Removal Through Steam-Generators from service in case of loss of circulation circuit integrity.


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Emergency heat removal tanks are installed in separate rooms of the outer structure of

the containment. The emergency heat removal heat exchangers are installed on special support structures inside the tanks.

Some part of pipelines, containment isolating valves and starting valves of the system

are arranged inside the Reactor Building (UJA). The other part of the containment

isolation valves of the system are outside the containment in separate rooms in the outer structure of containment, isolated from each other by fire-resistant barriers.

System operation

See items "Functions" and "Design description" above.

1.5.2.14 Residual Heat Removal System (JNA)

Functions

The Residual Heat Removal System (JNА) is intended to remove residual heat and to

cooldown the Reactor plant during NPP normal shutdown, in the event of AOOs as well

as DBAs, keeping the primary circuit integrity together with the Low Pressure Safety Injection System (JNG1).

The Residual Heat Removal System (JNА) is also intended to protect the primary circuit

against overpressure at cooldown modes and remove residual heat when temperatures in the primary circuit have decreased to the operation range of the system.

Design description

The system consists of four identical and independent trains. Each train includes a control valve, safety valve, valves and pipelines.

To organize coolant circulation in the system and heat removal from the Reactor core,

the pumps and heat exchangers of Low-Pressure Safety Injection System (JNG1) are

used. The JNA system provides coolant intake from the primary circuit system and

supplies it to JNG1 system to be cooled and returned to primary circuit system. The

pipeline of each system train is connected both to hot and cold loops of the primary circuit system.



One part of the JNA system equipment, including pipelines, safety valves and valves,

are located inside the containment, the other part (control valves, other valves,

pipelines) are located in separate rooms of Safety Building, isolated from each other by fire-resistant physical barriers.

System Operation

In the Reactor power operation, JNA system operation is not required.

The JNA system is connected when the primary circuit temperature drops below 150 С

and primary circuit pressure below 2.0 MPa in the primary system under conditions of RP cooldown.

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The residual heat removal and RP cooldown are carried out through the following closed

circuit: reactor (core) – Reactor Coolant Pipeline – JNA pipeline – emergency and

planned cooldown heat exchanger and bypass of this heat exchanger – emergency low

pressure injection pump – discharge pipeline of JNG-1 pipeline – Reactor Coolant

Pipeline – reactor (core).

Under cooldown conditions, the cooling rate of the systems is controlled by a regulator.

1.5.2.15 Intermediate Cooling Circuit For Important Consumers (KAA)

Functions

The Intermediate Cooling Circuit System For Important Consumers (KAA) is designed to

supply cooling water and remove heat from the reactor plant equipment, reactor plant

auxiliary systems and NPP safety systems under normal operating conditions, AOOs and

DBAs, as well as to provide a barrier between auxiliary systems containing radioactivity and the Cooling Water System for Important Consumers (PE).

Design description

The system consists of four trains, being independent in their process and electrical

connections. Each system train consists of a pump, heat exchangers, breathing tank, valves and pipelines.

“Cold” cooling water is pumped to the consumers; from the consumers the water is

drained into the return main pipeline, from which it is routed to the intermediate circuit

heat exchanger where it is cooled with service water of Pipelines System of Cooling

Water for Important Consumers (PEB). Then “cold” water re-enters the suction of the intermediate circuit pumps.


Equipment of the intermediate circuit (pumps, heat exchangers and tanks, as well as

valves and pipelines in the piping of pumps and heat exchangers) are located in rooms

of Safety Building (UKD). Each train is located in a separate room, isolated from the

other ones by fire-resistant physical barriers. One part of the system (pipelines and

valves) is located in the Auxiliary Building (UKA) and within the containment (water

supply and removal to/from the consumers of the Auxiliary Building/Reactor Building).

Mechanical filter is located in the Auxiliary Building.

System operation

Power operation: two trains of system KAA are in the operating and two others in

stand-by condition.

AOOs: When the temperature in intermediate circuit downstream heat exchangers

(KAA) rises above 30 ºС in unit power operation mode, decision is made to transfer the reactor plant into safe shutdown mode.

Under DBA conditions accompanied by coolant leakage (signal from the protection system), the following consumers are provided with cooling water of KAA:

– Heat exchangers of ECCS of low pressure (JNG);

– HP/LP safety injection pumps (JND, JNG);

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– emergency boron injection pumps (JDH);

– air coolers of the Safety Building rooms and those of inter-containment space;

– heat exchangers of the Fuel Pool Cooling system (FAK);

– heat exchangers of Intermediate Cooling Circuit for High-Pressure Important

Consumers (KAB);

– compressors of Radiation Monitoring Systems (KUK).

The consumers of the Reactor Building and Auxiliary Building are disconnected in

response to a signal from the plant protection system by closing cut-off valves. Service water pumps of each train automatically start operation at the same time.

1.5.2.16 Containment Hydrogen Removal System (JMT)

Functions

The Containment Hydrogen Removal System (JMT) is intended to avoid hydrogen

explosions in the containment in DBA, DEC and SA conditions. The Containment

Hydrogen Removal System prevents generation of explosive mixtures in containment by

means of maintaining the volume concentration of hydrogen in mixtures on a safe level.

Design description

The Containment Hydrogen Removal System’s (JMT) its functional elements are

independent from each other and other systems of the NPP. The passive autocatalytic

hydrogen recombiners, which are the basic functional elements of the system, start

operating when hydrogen concentration in the room increases and continue operating

until hydrogen concentration is reduced to the safe level. Containment Hydrogen Removal System is an arrangement of recombiners in containment rooms.

To ensure maximum efficiency of the system, the recombiners are installed in places

where hydrogen concentration during an accident may reach maximum levels, as well as in the routes of steam-and-gas medium.


All functional elements of the system are uniformly distributed over the containment.

Steam generator rooms are an exception.

Most recombiners are installed in the steam generator rooms, due to maximum

hydrogen concentrations reached in the steam generator rooms during loss of coolant

accidents. To protect the recombiners against jets and missiles, they are located in areas enclosed by engineering structures and large-sized equipment.

System operation

See items above.

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Figure 1.5.2.1 The schematic diagram of the safety systems.

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1.5.3 Instrumentation and control systems (I&C)

1.5.3.1 General

The NPP is controlled and supervised with control systems that form an I&C systems

architecture with a structure fulfilling requirements set in applicable legislation, YVL

guidance as well as codes and standards. The I&C systems architecture consists of I&C

systems and control room systems for safety, safety-related and non-safety functions,

with related necessary field equipment and instrumentation. There are also separate

individual special systems for certain control and monitoring functions apart from the main NPP processes.

The NPP control systems utilize mainly computerized I&C technology. There are

different platforms for safety and non-safety systems. Both of them use mainly

software-based programmable computer technology.

I&C systems are implemented using specific control technology with high reliability. The

technology is designed to be used in the control and monitoring applications of nuclear

or large power plants. The technologies used have been proven in similar applications.

Project-specific implementation of the systems will be designed, implemented and

tested during the design and realization phases of the plant. For non-safety systems,

I&C systems technology is comparable with high-quality standards of corresponding

nuclear or large power plants. For the safety and safety-related I&C systems,

technology is based also on nuclear industry standards.

I&C systems architecture is implemented to meet the requirements of defense-in-depth

(DiD) principle. This is presented in separate I&C systems architecture description.

Generally, I&C systems belonging to different DiD levels shall be adequately

independent of each other and this adequacy shall be justified during the design. See

chapter 1.4, particularly sub-chapters 1.4.3 "Defense-in-depth concept" and 1.4.4 "Event categories".

Human System Interface (HSI) consists of equipment located in control rooms and

other control posts/places to enable the operation personnel to monitor and control the

processes. The HSI equipment is a part of the I&C systems and follows the same design

and quality requirements. Control rooms will be designed to meet high ergonomic

requirements as well as to provide protection against any events that could risk their

habitability. Human factors engineering (HFE) principles will be followed in the design,

verification, validation and implementation of control room environment and HSIs.

The I&C systems and equipment are classified according to the safety classification

principles of YVL B.2. Graded approach will be applied to the requirements of the

systems and equipment and to their quality activities, design processes and qualification requirements.

(Nuclear) redundancy is required in the I&C systems, which are designed to satisfy N+1

or N+2 criteria, depending on the safety significance of the system. Defense-in-Depth

level 3a satisfies N+2 criteria (four independent trains) for the functions required for

achieving a controlled state. From controlled state to safe state, N+1 criteria is fulfilled.

Other levels are designed based on either N+1 or N+0 criteria as required in YVL

guidance. Other related design principles regarding redundancy, separation and

diversity are presented in chapter 1.4, especially sub-chapter 1.4.5 "Technical implementation and ensuring reliability of safety functions fulfilment".

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Diversity will be used as an essential means to increase the reliability of the I&C

systems. There will be a separate I&C system, serving as a backup for the functions

required in DBC 2/3. This separate system will use different technology than the

programmable main safety systems, so the risk of common-cause failure can be

minimized. Diversity is also applied in the measurements required for actuation of

safety systems. In principle, each protection function in a reactor protection system

shall have at least two different measurement parameters for starting the function. If

this is not possible, at least two different measurement principles for a parameter shall

be used. Exceptions to these principles will be justified separately.

In general, diversity falls into the following types:

– hardware diversity for software based functions

– parameter diversity

– equipment diversity

More detailed description of the I&C systems is presented in a dedicated PSAR chapter.

1.5.3.2 Basic tasks of I&C systems

I&C tasks are divided into:

– information tasks

– control tasks

– system tasks

The implementation of these tasks differs in different systems. Generally, non-safety

systems provide a wider scope of functionality than safety systems, which only provide functionality necessary for fulfilling safety functions reliably.

Information tasks are as follows:

– monitoring of operating conditions;

– monitoring and diagnostics of process equipment and their condition;

– monitoring of unit safety parameters;

– support to operating personnel in emergency situations with specific

information;

– monitoring of personnel habitability, safety and post-accident measures;

– providing necessary information for higher level information systems (plant

level and corporate levels).

Complex calculations (e.g. process equipment condition monitoring and diagnostics) are

not implemented in safety I&C systems, due to the requirement for simplicity to ensure highest reliability.

Control tasks are the following:

– remote manual control of actuators (also called "automated control")

– automatic control of equipment;

– interlocking of controls with certain criteria due to safety and other reasons

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Control tasks are mainly similar in safety, safety-related and non-safety systems.

System tasks are the following:

– universal time of system;

– communications between systems (where feasible);

– continuous failure tolerant operation of subsystems and control system as a

whole;

– acquisition, presentation and recording of data both on operation of the NPP

processes and on operation of the I&C systems

– monitoring and diagnostics of the current state of control systems

Safety I&C systems are real-time systems, which operate continuously. Therefore their

functionality regarding system tasks is more limited than non-safety systems. This concerns e.g. the handling of time information in the systems.

1.5.3.3 Design requirements

The I&C systems design process will follow the highest standards regarding design

control and quality. Particular attention will be paid on safety class 2 systems and

equipment. Design processes will be guided by specific I&C system related quality plans. Particularly, IEC 61513 standard will be used to guide the design processes.

The design process will have clear phases; development of the I&C systems design will

proceed from the high level requirements towards detail level plans. This process

involves requirement management activities to ensure that the detail level

implementation is in accordance with the requirements addressed in all phases of the

design. Configuration management will be applied to ensure the consistency of different

plans and implementations. Verification and validation activities will be implemented as

part of all design activities.

The architectural level design of I&C systems will be considered a design phase of its

own. I&C systems architecture addresses the overall structure of the I&C systems,

allocation of I&C functions, interfaces between systems and e.g. interfaces to the HSIs.

Architectural solutions will prescribe essential requirements and constraints to the design of individual I&C systems.

The resulting documentation will be presented in dedicated documentation, subject to regulatory proceedings approval if applicable.

Individual I&C and control room systems will be designed taking into account the

requirements of I&C systems architecture design, regulatory guides, relevant standards

and other sources. A graded approach will be applied in considering required quality

activities control and quality assurance. This will be based mainly on the safety

classification of the systems. Nuclear standards will be applied to systems of safety class 2, (e.g. reactor protection system).

I&C systems and the equipment of safety classes 2 and 3 will be qualified for their

intended use. Qualification programs will be established to justify the use of each

component considering the intended use of equipment and the environmental

conditions of the intended place of installation. The most critical equipment will be

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subjected to a type approval process involving third party assessments. Qualification

and type approval activities will be carried out according to YVL E.7.

1.5.3.4 Security

Information security requirements will be applied to the design and implementation of

the I&C systems. Particular attention is paid to ensure the sufficient physical,

technological and administrative security arrangement over the lifecycle of the I&C

systems' design and implementation. In case of software-based systems, cyber security

of software-based systems as well as software development processes will be considered. Efficient means to prevent cyber-attacks will be established.

1.5.3.5 I&C functions

I&C functions are defined based on plant safety functions. This design process involves

defining the postulated initiating events and specific means to respond to each of them.

This process defines plant level functional architecture, which gives requirements and

constraints to the I&C systems design, and gives guidance for the definition of I&C

functions. I&C functions are allocated to individual I&C systems. Details about the

safety functions of Hanhikivi-1 are described in dedicated PSAR chapters. As a basis for

the safety functions, there are defined particular plant conditions, which are referred e.g. as normal operation, controlled state and safe state.

In the normal operation conditions, the plant is controlled by normal operation I&C systems.

When a postulated initiating event takes place, the plant moves to an abnormal

condition. According to DiD principle, relevant safety functions by predefined I&C

systems are actuated. Safety systems have protecting functions, which automatically

initiate the responses to the initiating events, whenever it is needed, and maintain the

control of mitigation actions until the reactor is brought to the controlled state. In these

situations, the division of tasks between the I&C system and human operator is

designed based on task analyses. I&C and process safety systems, in combination, shall

be able to function 30 minutes from the moment of their actuation in relevant

Postulated Initiating Event, keeping the situation under control without human

intervention.

Transition from the controlled state to the safe state may involve manual or automatic

functions, since it involves no actions with quick response requirements.

1.5.3.6 I&C systems architecture and systems

I&C systems architecture will be developed to ensure that the overall structure of the

I&C systems complies with safety principles and with the defense-in-depth concept of

the plant. The I&C systems architecture also addresses the allocation of I&C functions (see above) to I&C systems.

Levels of defense-in-depth (DiD) are adequately independent of each other.

Independence of the levels will be required considering all relevant Postulated Initiating Events.

Based on the DiD concept, I&C systems will be divided to:

– Normal operation I&C (NO I&C), DBC 1 functions

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– Safety-related normal operation I&C (SR NO I&C), DBC 2 functions

– Safety I&C, DBC 3–4 functions

– Design extension condition I&C (DEC I&C), DEC functions

– Severe accident I&C (SA I&C), SA functions

When same equipment can be controlled by different systems (e.g. normal operation

control and safety system actuation), there will be requirements and technical

implementation, which describe the prioritization of any control actuations, even if there

are contrary commands in force.

In addition to the main I&C systems, there is a group of separated individual special

systems for certain specific control or monitoring functions apart from the main NPP

processes. These systems will be safety classified based on their significance to safety, and they follow the design requirements based on the classification.

1.5.3.7 Control rooms and human system interfaces

Monitoring and control of the plant will be centralized into the main control room (MCR).

It provides HSIs of all relevant systems (I&C systems as well as other types of systems) needed to control the NPP.

MCR provides an ergonomic working environment for the plant operators. MCR is

continuously manned by operators working in shifts. In addition to NPP process control,

MCR plays a key role during any maintenance work carried out in the plant. Necessary

provisions for supporting maintenance work control will be provided.

The HSIs of the I&C systems provide operators with relevant real-time information of

the process parameters. Important features are e.g. trend curve displays presenting the

history of any parameter value, and alarm functions focusing operators' attention to any

abnormal parameter value requiring attention. The HSIs of digital I&C systems consist

mainly of digital display interfaces, but also hard-wired conventional monitoring and control equipment is present.

Emergency control room (ECR) provides functions for shutting down the plant in case of

loss of the MCR, e.g. in case of fire. ECR and MCR are physically, functionally, and

electrically separated from each other, so that no event could disable both of them

simultaneously, and any event in the MCR could not affect the usability of the systems from ECR.

In addition to MCR and ECR, an Emergency response command post (ERCP) is provided,

to be used in emergency situations (or emergency exercise). The ERCP shall contain an

applicable monitoring system with accessories indicating current plant situation by

relevant plant and environmental parameters. These enable a comprehensive snapshot

for situation awareness of the NPP and its environment. ERCP also provides space and

tools for the emergency support organization personnel. From ERCP it is necessary be

able to handle the necessary communications (e.g. with authorities), as well as to

advise the control room personnel on how to handle the emergency situation of the plant.

A plant specific training simulator will be built for training the operators, maintenance

personnel and other personnel of the plant. The training simulator will be a full-scale

replica of the main control room with HSI's, connected to necessary process simulators

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and to particular simulator controlling systems. The training simulator will be built well

in advance to enable necessary amount of training for the personnel.

1.5.3.8 HFE

Human factors engineering (HFE) program will be established in the design phase of the

project. Its objective is to ensure that risk for human errors is minimized with the right

design solutions. HFE ensures that the plant facilities and tools, such as HSIs and

procedures, support safe operation of the plant in all plant conditions and operation

modes. Nureg-0711 standard will be used in HFE activities.

1.5.4 Power supply system

1.5.4.1 General

The NPP power supply system consists of a transmission system, power generation

system and auxiliary power supply systems. This sub-chapter of PSAR chapter 1 gives a

generic overview of the power supply main structure and its functions. More detail

system descriptions are provided in dedicated chapter 8 of PSAR.

The power distribution system is designed according to plant safety principles,

described in chapter 1.4. The following principles are applied to the NPP power systems

in order to guarantee reliable power source to NPP consumers in all operational and design accident conditions:

– redundancy principle

– diversity principle

– independence principle

– protection against common course failures

– controllability and maintainability.

The power supply system is designed to comply with Finnish laws and regulatory

guides, as well as appropriate design standards. It is subject to defense-in-depth

concept. Safety classification and graded approach is applied to electric systems and equipment.

Main diagram of electrical connections of unit is given on figure 1.5.4.1. Power supply

for unit auxiliaries is shown in figure 1.5.4.2.

Plant has 400 kV and 110 kV connections to the national grid. Schematic diagram is in

figure 1.5.4.1.

1.5.4.1.1 400 kV system

The 400 kV switchgear is constructed on the NPP site. Two overhead power

transmission lines of 400 kV are constructed to connect to the national grid system at

Hanhela substation, located 21 km from the NPP site. Capacity of each 400kV air

overhead power transmission line is designed for full power of the unit.

The 400 kV switchgear is designed for connection of:

– two generator transformers of the unit ;

– two auxiliary normal transformers;

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– two transmission lines of NPP Hanhikivi – Hanhela substation;

– reserve cell for connection of unit 2 is provided.

The 400 kV switchgear is a double busbar gas insulated switchgear.

1.5.4.1.2 110 kV system

The 110 kV switchgear is constructed on the NPP site. Two overhead power

transmission lines of 110 kV are constructed to connect to the national grid system at Valkeus substation, located 21 km away from the NPP site.

The 110 kV switchgear has the following inputs:

– two auxiliary stand-by transformers;

– two transmission lines NPP Hanhikivi – Valkeus substation;

– one transformer for site 20 kV distribution

– reserved cell for connection of gas turbine plant.

The 110 kV switchgear is a double busbar gas insulated switchgear.

1.5.4.1.3 Power generation system

Turbogenerator power of about 1200 MW with rotation speed of 1500 rpm, 24 kV, 50

Hz is installed on the unit.

Generator and two three-phase transformers are connected via air-cooled isolated phase busbars (IPBB).

Power of each generator transformer is selected based on 70% of unit capacity.

In both generator transformer lines, there are full capacity generator circuit breakers.

1.5.4.1.4 Auxiliary transformers

Two three-core auxiliary normal transformers (ANT) of capacity 80/40-40 MVA each,

voltage rating 400/10.5-10.5 kV, are connected to high voltage 400 kV NPP switchgear.

ANTs of the unit are equipped with on-load tap changers for voltage regulating.

Auxiliary 10 kV main switchgear (four pcs) for the unit is backed up by 110 kV stand-by

power supply network via two auxiliary standby transformers (AST) of 110/10.5-10.5, each connected to 110 kV switchgear of NPP Hanhikivi with cable.

Two three-core auxiliary standby transformers (AST) of capacity 80/40-40 MVA each,

voltage rating 110/10.5-10.5 kV, are connected to the 110 kV NPP switchgear. ASTs of

the unit are equipped with on-load tap changers for voltage regulating.

Secondary side of ASTs are provided with a 10 kV switchgear, permanently energized.

Standby power supplies are routed to the unit auxiliary 10 kV main switchgear.

Two power infeeds are provided for each 10 kV switchgear of the unit normal power supply system: from the ANT and from the AST.

High-speed automatic switchover device is provided for each 10 kV switchgear.

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Figure 1.5.4.1 Main diagram of electrical connections of unit (simplified).

Plant 400

kV GIS

`

GT 1 GT 2

Generator

ANT 1 ANT 2

10kV switchgear

Reservation for

connection Unit 2Reservation for

connection Gas turbine

AST 1 AST 2

110 kV GIS

To grid To gridTo grid To grid

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1.5.4.2 Operation of power distribution system

Under normal operation conditions, the turbine plant is operating. Unit switching

equipment—24 kV generator circuit breakers and 400 kV switchgear circuit breakers of unit—is closed. The 400 kV switchgear lines circuit breakers are closed.

Under normal operation conditions, unit auxiliary power supply systems are supplied from ANTs of 400/10.5 -10.5 kV.

Network of the 110 kV standby power supply system is energized. Standby supply incoming circuit breakers on unit 10 kV sections are open.

Main anticipated operational occurrences related to unit output are:

– loss of external load 400 kV;

– deviation of frequency or voltage in power system to an inadmissible level;

– unit disconnection by protection system of the generator transformer unit;

– disconnection of one of the unit generator transformers by relay protection.

Consequences of anticipated operation occurrences are described in dedicated parts of

PSAR Chapter 8.

The 400 kV switchgear and 400 kV transmission lines are protected with a set of main and

standby quick response protecting devices. This measure provides stability and safety of

power system operation, as well as accident mitigation and minimum consumers blackout.

Protection devices for the generator transformer unit and auxiliary normal and standby transformers are implemented as a set of two independent systems.

The NPP automated electrical instrumentation and control system (Electrical I&C) is

designed to improve power output availability of the power plant. It is realized through

increasing equipment reliability and preventing erroneous actions of operating personnel

as well as speeding up and improving interactions with other control systems. The

electrical I&C system performs the functions related to measuring, recording, diagnosing

and controlling of electrical equipment, and it provides information to operators. The plant

is also provided with a diverse monitoring system.

1.5.4.3 Plant auxiliary power supply system

1.5.4.3.1 Functions

Auxiliary power supply system is intended to supply power to NPP consumers under

different operating conditions:

– normal operating conditions;

– anticipated operational occurrences;

– design basis accidents;

– design extension conditions;

– severe accident.

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1.5.4.3.2 Design bases

According to their purpose, the auxiliary power consumers of unit are divided in relation to

their safety significance as follows:

– normal operation consumers,

– normal operation safety-related consumers,

– safety systems consumers,

– consumers of systems designed for design extension conditions,

– consumers required for operation under severe accident conditions.

In accordance with plant safety concept, the following auxiliary power supply systems are

provided in the NPP design:

– Normal Operation Power Supply System for the auxiliary consumers (NO PSS),

which ensures functioning of the NPP systems and equipment at DiD level 1. The

system is referred to safety class EYT. The reference designation of NOPSS

power supply trains is accepted in the design as "9" and "10" (see figure 1.5.4.2

below).

– Normal Operation Reliable Power Supply System for the safety-related auxiliary

consumers (safety-related NO RPSS), which ensures functioning of the NPP

systems and equipment at DiD level 2. The system is referred to safety class 3.

The reference designation of safety-related NO RPSS power supply trains is

accepted in the design as "5" and "6".

– Emergency Power Supply System (EPSS), which ensures functioning of the NPP

systems and equipment at DiD level 3a. The system is referred to safety class 2.

The reference designation of EPSS power supply trains is accepted in the design

as "1", "2", "3" and "4"; the number of trains is equal to the number of safety

system trains, which are the EPSS consumers.

– Reliable Power Supply System for DEC (DEC RPSS), which ensures power supply

to the NPP systems and equipment at 3b level. The system is referred to safety

class 3. The reference designation of DEC RPSS power supply trains is accepted

in the design as "5b" and "6b".

– Reliable Power Supply System for SA (SA RPSS), which ensures power supply to

the NPP systems and equipment at level 4. The system is referred to safety class

3. The reference designation of SA RPSS power supply trains is accepted in the

design as "7" and "8". Technical solution for total loss of AC network is described

in dedicated PSAR chapter 8.

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Figure 1.5.4.2 Schematic diagram of power supply for unit auxiliaries (simplified)

Based on the requirements specified for uninterrupted power supply, the NPP auxiliary

consumers are divided into three groups in order to ensure the continuity of the

appropriate processes (considering their importance with respect to ensuring of NPP main

processes, nuclear safety and personnel safety). Grouping criteria is based on whether it is acceptable that power supply to the consumer is interrupted, and for how long.

– The first group: AC and DC consumers that due to safety and equipment integrity

requirements do not allow power interruption for more than a split second under

all conditions, including complete loss of voltage from operating and standby

external power supply sources;

– The second group: AC consumers requiring power supply under conditions of

complete loss of voltage from operating and standby external power supply

source. These consumers allow interruptions in power supply for the time

determined by conditions of the main equipment safety and integrity, but not

less than the time required for starting up and connecting standby diesel-

generators;

– The third group: AC consumers allowing interruptions of power supply for the

time of automatic switchover to backup device actuation and not requiring

GT 1 GT 2

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Control building

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Standby diesel-

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Control building

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obligatory availability of power supply under conditions of complete loss of

voltage from operating and standby external power supply sources;

Auxiliary power supply system consists of an alternating current system and direct current

system. A schematic diagram of the power supply of the unit auxiliaries is given in figure 1.5.4.2 above.

1.5.4.4 Alternating current electrical systems

1.5.4.4.1 System design

According to voltage level, the following AC power supply networks are accepted for

auxiliary power supply systems:

– Auxiliary AC systems of 10 kV,

– Auxiliary AC systems of 0.4 kV/0.23 kV;

– Backup diesel generators for various purposes;

– Uninterrupted auxiliary AC systems of 0.4 kV/0.23 kV.

1.5.4.4.2 Backup Diesel Generators

In case of complete loss of voltage from operating and standby external power supply

sources (loss of off-site power), transition to the power supply from 10 kV and 0.4 kV diesel generator sets is performed at unit.

Installation of the following diesel generator sets is provided:

– Turbine Island diesel generator;

– Nuclear Island diesel generators;

– diesel generators of Emergency Power Supply System;

– diesel generators of severe accidents.

In accordance with IAEA recommendations and YVL B.1, a diesel generator station is

provided as an external power supply source to ensure additional independent redundancy

of Emergency Power Supply Systems (EPSS). Additional independent diesel generator

source can be connected manually, when needed. This is called a Site Backup Diesel. The Diesel generator can be manually connected to EPSS 10 kV switchgear.

Transferring of power supply of the Nuclear Island and safety system essential consumers

from the relevant diesel generators is performed automatically at a signal generated in

case of unacceptable decrease of voltage or frequency in 10 kV reliable power supply

switchgear for a period exceeding the time of automatic switchover to backup.

The total load required in case of loss of off-site power will be supplied with power from diesel generators by the automatic step-by-step start-up program.

The power of each diesel generator set is chosen in such a way that the supply power

required for consumers is provided without overloading. During normal operation, diesel generators are in standby mode and ready for automatic actuation.

1.5.4.4.3 Severe accident power supply

Power supply for loads of severe accident system is performed from the separate power

network consisting of two trains. In case of loss of voltage at 0.4 kV switchgear, the

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consumers are supplied with power from batteries forming a part of uninterrupted power

supply unit (UPS unit), designed to provide power for 24 hours. In order to ensure the

power supply of equipment control and management system during severe accidents for

72 hours (and more if necessary), two diesel generator sets are provided—one set for

each train. The diesel generator is able to ensure power supply of equipment for severe

accidents management during the required time period. The possibility of replenishment fuel and oil storages is provided for, when diesel generators need to be operated longer.

1.5.4.5 Direct current electrical systems

The following backup batteries are installed at unit

– batteries of 220 V for power supply of the turbine building loads (Normal

Operation Reliable Power Supply System),

– batteries of 110 V for the reactor control and protection system (control rods),

– batteries of 220 V for power supply of the control and reactor buildings, Normal

Operation Reliable Power Supply System for safety related auxiliary consumers

and Reliable Power Supply Systems for consumers under design extension

conditions,

– batteries of 220 V for power supply of loads of Emergency Power Supply System,

– batteries forming a part of uninterrupted power supply unit (UPS unit) for power

supply of loads of the instrumentation and control system during severe

accidents,

– batteries of 220 V for power supply of 400 kV, 110 kV switchgear load and load

of auxiliary buildings and structures.

In case of loss of alternating current, direct current consumers are supplied with power

directly from the DC switchgear, to which batteries and charger rectifiers are connected.

These batteries and charger rectifiers are fed with power from the relevant 0.4 kV/0.23 kV switchgear.

After restoration of AC power supply for rectifiers, the power supply of the direct current

consumers is automatically transferred to rectifiers. Rectifiers provide charging of the relevant batteries and feeding of the direct current consumers with power.

Divisional separation is performed in the design of AC and DC power supply systems.

1.5.4.6 Location of equipment

All electrical equipment is installed in electrical rooms for safety and non-safety

equipment.

All electrical equipment for different voltage levels is also installed in separated rooms:

– Medium voltage switchgear rooms,

– Low voltage switchgear rooms,

– Battery rooms,

– Direct current system equipment rooms,

– 10/0.4 kV transformers rooms,

– I&C system equipment rooms.

All switchgear and I&C rooms have cable room below.

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The main equipment of power generation and transmission system is located in the

following buildings and structures:

– Turbine Building 10UMA;

– 400 kV Switchgear Building 10UAB;

– 110 kV Switchgear Building 10UAE;

– structure of unit transformers 10UBF.

The main electrical auxiliary equipment of the unit is located in unit buildings 10UCB,

10UBA, 10UBN, 10UBS, 10UGB.

The EPSS electrical equipment and electrical equipment ensuring power supply of the main

unit I&C systems and electrified processing equipment located in buildings 10UKA, 10UJA, 10UJE, 10UKC, 10UKD, are located in Control Building (10UCB).

The 10 kV and 0.4 kV switchgears of unit (NO PSS), 10/0.4 kV transformers and inverters,

relay protection and unit automatics equipment as well as batteries and direct current

boards of 220 V for power supply of the consumers of Turbine Island building are located in Normal Operation Power Supply Building (10UBA).

The Turbine Island diesel generator and its auxiliary equipment, 0.4 kV switchgears of

Normal Operation Reliable Power Supply System (NO RPSS) auxiliaries are located in the

Unit Diesel generator Station Building of Turbine Island with Diesel Fuel Intermediate

Storage 11UBN.

Two Nuclear Island diesel generators and their auxiliary equipment, 10 kV switchgears of

auxiliaries of Normal Operation Reliable Power Supply System for safety-related

consumers (safety-related NO RPSS), 10/0.4 kV transformers and 0.4 kV switchgear for

DGS auxiliaries are located in the Unit Diesel generator Station Building of Nuclear Island

with Diesel Fuel Intermediate Storage 12UBN.

Two diesel-generators of Emergency Power Supply System (EPSS) and their auxiliary

equipment, 10kV and 0.4 kV switchgear of auxiliaries, 10/0.4 kV transformers, inverters,

batteries and direct current boards of 220 V for power supply of safety system consumers,

0.4kV switchgear for DGS auxiliaries are located inside the building of each Standby

Diesel-generator Station Buildings of Emergency Power Supply System (EPSS) with Diesel Fuel Intermediate Storage 11(12) UBS.

Diesel generators of the first and second safety trains are located in the standby diesel

generator station building 11UBS; diesel generators of the third and fourth safety trains

are in building 12UBS.

Secondary switchgears of 0.4/0.23 kV are provided to power supply the consumers of

remote buildings with low total load as well as to decrease the number of cable connections in the main buildings and unit structures.

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1.5.5 Auxiliary Systems

1.5.5.1 Fuel Storage and Handling

1.5.5.1.1 Spent Fuel Pool Cooling and Cleanup Systems

1.5.5.1.1.1 Fuel Pool Cooling System (FAK)

Functions

The Fuel Pool Cooling System (FAK) is designed for:

– removal of decay heat from the spent fuel assemblies stored in the fuel pool

under all design basis conditions;

– formation of a radiation protection layer above the fuel assemblies in the reactor

cavity, fuel pool and refueling well;

– filling the reactor cavity and fuel pool during refueling;

– drainage of the reactor cavity and reactor internals inspection well;

– drainage of the fuel pool and refueling well for the liner repairing purposes;

– FAK system pipelines are used for water supply from containment spray pumps

(JMN) for filling the reactor internals inspection well during refueling.

System description

The system consists of two identical and completely independent trains; each includes a

pump, heat exchanger, intake and distribution collectors, valves and pipelines. Both trains

take warmed water from the upper elevations of the fuel pool, supply it to suction of FAK

pumps and then to heat exchangers, where it is cooled by Intermediate Cooling Circuit for

Important Consumers (KAA). After cooling, the water is returned to the fuel pool via the header installed approximately at the elevation of lower ends of fuel assemblies.

During normal operation and partial refueling of reactor core, one train of the system

ensures heat removal from the fuel pool. Simultaneous operation of both trains is required

in case of full or emergency core unloading into the fuel pool. The system design provides

redundancy of each train by systems JNG/JNA/JMN. Under design extension conditions

(DEC) related to simultaneous failure of main and backup systems or Loss of Ultimate

Heat Sink, cooling of the fuel pool is performed by means of boiling the water in the pools.

For these conditions, makeup of the fuel pool is provided by an emergency makeup system, which is connected to the pipelines of the FAK system inside the containment.

In case of fuel pool leakages, makeup is performed from the Borated Water Storage Tanks

(JNK) with the help of Purification System of Water in Fuel Pool and Borated Water Storage

Tanks (FAL) or Spray (JMN) system pumps, depending on the size of the leak.

1.5.5.1.1.2 Purification System of Water in Fuel Pool and Borated Water Storage Tanks (FAL)

Functions

The Purification System of Water in Fuel Pool and Borated Water Storage Tanks (FAL) is

intended to perform the following functions:

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– purification of fuel pool water (FAK) from mechanical and dissolved impurities to

lower its activity and assure transparency;

– purification of borated water storage tank water (JNK) from mechanical and

dissolved impurities.

System description

The system consists of a heat exchanger, pumps, powdered resin ion-exchange units,

mixed bed filter, filter trap, layer maintenance pumps, ionite powder pump, ionite powder

deposition tank, pipelines and valves. The system operates continuously during reactor

refueling and periodically under other operating conditions, depending on the quality of the fuel pool and the borated water tanks water.

The water to be purified is first cooled in a heat exchanger by Intermediate Cooling Circuit

for Important Consumers (KAA) and then pumped through either of the parallel powdered-

resin ion-exchange units. One of the filters is in operation, the other in standby. Then

water is directed to mixed bed filter. After purification, water is returned back either to the fuel pool or borated water storage tank.

1.5.5.1.2 Fuel Storage System

Functions

The fuel storage system is designed for reception, storage and incoming inspection of

fresh nuclear fuel, as well as for the storing of spent fuel in the fuel pool inside the reactor

building containment.

System description

Fresh fuel to be loaded into the reactor is first brought to the fresh fuel storage in

transportation package sets. The storage is provided with equipment for storage,

transportation and handling of fuel assemblies. The fresh fuel storage is equipped with the following auxiliary systems:

– ventilation system;

– fire alarm system;

– working and emergency lighting;

– security alarm;

– radiation monitoring system.

Spent fuel is stored on cooling pool racks in the reactor building. The pool has space for

complete core unloading in case of emergency. For the fuel pool, the design provides

water supply, purification and cooling systems as well as radioactivity, temperature and chemical composition monitoring facilities.

1.5.5.1.3 Refueling System

Functions

The refueling system is designed to replace spent fuel and rods cluster control assemblies (RCCA) with new ones. The refueling system provides:

– unloading of fuel assemblies and RCCAs from the core into the cooling pool;

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– rearranging fuel assemblies and RCCAs inside the reactor core;

– loading of fresh fuel assemblies and RCCAs into the reactor;

– loading of spent fuel assemblies from the cooling pool into transportation

package set.

System description

Reactor refueling is carried out using the refueling machine under water. The refueling

machine can perform operations only with one fuel assembly or one RCCA, or one fuel

assembly together with RCCA at a time. Transport operations with fresh fuel jackets and spent fuel transport containers in the reactor building are performed with a polar crane.

1.5.5.1.4 Fuel Rod Cladding Leak Check System

Functions

The system for leak check of fuel assembly (FA) fuel rod claddings is intended to perform inspection at the shutdown reactor and includes the following branches:

– FA check in fuel handling machine (FHM) mast with the help of leak check of

claddings system fuel handling machine (LCCS FHM);

– FA check with the help of defective assemblies detection system (DADS);

– check and repair of leaky fuel assemblies with the help of fuel repair and

inspection equipment (FRIE).

System description

If there is a need for leak check of claddings at the shutdown reactor, all FAs of a fuel

cycle are subject to it by means of online checking in the fuel handling machine mast

(LCCS FHM) in order to obtain data on FA leak tightness. Leak tightness is checked by

analyzing gas samples taken through the internal space of the FHM mast from the FA.

The FAs found to leak using the online method are checked with the help of the sampling

part of the defective assemblies detection system (DADS) to determine the degree of FA leakage and a possibility for its further operation.

The fuel repair and inspection equipment (FRIE) serves to repair leaking fuel assemblies to return them into fuel cycle for reduction of operating costs of the NPP.

The FRIE is intended to:

– inspect fuel assemblies (visual examination, check of geometric dimension, etc.);

– perform rod-by-rod leak check of claddings;

– repair fuel assemblies.

1.5.5.1.5 Leakage Detection System for Fuel Pool Liner (FAB)

Functions

The Leakage Detection System for Fuel Pool Liner (FAB) is designed to monitor integrity of

the fuel pool liner and reactor internals inspection shaft to detect water in the gap between the liner and the wall in every liner section.

System description

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The system consists of discharge tubes penetrating the walls of the fuel pool, refueling

pool and reactor internals inspection shaft, removing possible leaks from the gap between

the liner and the walls. Each tube is fitted with a three-way valve. The discharge branch

pipes of this valve are connected either to the gathering header with electric-driven

valves, whose outlet is directed to the collecting tank, or to the control funnel. Outlets of

the control funnel also lead to the header and then to the drainage line downstream of the

cut-off electric-driven valves. The leaking section can be determined by switching over the three-way valves and monitoring the filling speed of the collecting tank.

1.5.5.2 Water Systems

1.5.5.2.1 Station Service Water System

All the cooling water systems of Hanhikivi-1 NPP are based on a direct-flow principle with a

single seawater circulation through heat exchanging equipment. The Baltic Seawater is

used as a source of service water and as an ultimate heat sink (UHS). The structures of ultimate heat sink includes:

– seawater inlet channel 10UPF

– intake structure 10UPC with coarse screens

– outlet channel 10UQF back to the sea

1.5.5.2.1.1 Main Circulating Water System (PA)

Functions

The PA system is designed to supply cooling water and remove heat from the turbine condensers to the ultimate heat sink under all conditions of normal operation.

System description

The system comprises the elements ensuring cooling water supply and discharge, as follows:

– supply tunnel 10UPN from intake structure to the inlet surge pool adjacent to the

Pumping Station of Turbine Building Consumers 10UQA

– pumping station with fine screens, rotating screens and main cooling water

system pumps

– pressure pipelines between the pumps and outlet surge pool through condensers

– discharge tunnel 10UQN between outlet surge pool and seawater outlet structure

10UQQ

During NPP power operation, the seawater is purified with coarse screens within the intake

structure 10UPC and supplied to pumping station 10UQA water intake through supply

tunnel 10UPN. Within the pumping station, the seawater is filtered with fine and rotating

screens and then pumped by main cooling water pumps through turbine condensers via

pressure pipelines. After turbine condensers, the warmed cooling water is gathered into

the outlet surge pool and led through discharge tunnel 10UQN and outlet channel 10UQF

back to the sea.

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1.5.5.2.1.2 Auxiliary Circulating Water System (PC)

Functions

The PC system provides cooling water supply and residual heat removal from the

Intermediate Cooling Circuit System for Unimportant Consumers (PGB) and the chiller

condensers of the chilled water supply systems for HVAC systems.

System description

The PC system comprises the elements providing cooling water supply and discharge, as follows:

– pumps;

– filters;

– inlet and discharge conduits, pipelines and valves.

The PC system takes filtered cooling water from the same suction basin as the main

cooling water system (PA) pumps in the Pumping Station of the Turbine Building

Consumers 10UQA. Cooling water is supplied to the consumers, and warmed water is discharged to the common outlet surge pool with PA system.

1.5.5.2.1.3 Cooling Water System for Important Consumers (PE)

Functions

The PE system provides cooling water supply and heat removal from the Intermediate Cooling Circuit for Important Consumers (KAA).

System description

The system consists of four identical and completely independent trains, each including a

fine screen, rotating screen, two pumps, valves and pipelines. Cooling water from the sea

is led to the PE pumping station 10UQB via dedicated supply tunnel for essential

consumers 10UPP, branching off the PA system supply tunnel 10UPN after intake

structures 10UPC and coarse screens. Within the pumping station, the water is filtered and

then pumped through heat exchangers of Intermediate Cooling Circuit for Important

Consumers (KAA). Warmed water is led back to the sea via PE system dedicated Discharge

Tunnel of Essential Consumers 11UQP, which joins to the PA system discharge tunnel 10UQN before outlet structure 10UQQ.

If the normal seawater intake side is unavailable, there is a possibility to take cooling

water for the PE system from the discharge side. Cooling water from the outlet channel

10UQF is supplied to the Pumping Station of the Essential Consumers 10UQB through the

outlet structure 10UQQ and Discharge Tunnel of Essential Consumers 11UQP. Heated

water is discharged back into the sea through the Standby Discharge Tunnel of the

Essential Consumers 12UQP and Standby Water Outlet Structure 10UQU.

1.5.5.2.2 Makeup Water System (LCU)

Functions

The LCU system’s main function is demineralized water storage and supply to primary and secondary circuit consumers under different operating conditions.

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System description

The system comprises the elements ensuring makeup water supply to primary and

secondary circuit consumers, as follows:

– demineralized water storage tanks;

– makeup water pumps;

– emergency makeup water pumps;

– filters-absorbers;

– pipelines and valves.

For normal makeup of the unit, two makeup water pumps are installed (one operating, the

other standby). The system ensures makeup water supply for filling pipelines before start-

up, flushing of condensate purification filters, makeup of Intermediate Cooling Circuit for

Important Consumers (KAA) and Intermediate Cooling Circuit System for Unimportant

Consumers (PGB).

For emergency makeup of the unit, two emergency makeup water pumps are installed

(one operating, the other standby). If the water inventory at the unit decreases

considerably, emergency makeup is performed into the feedwater deaerator (LAA). The

emergency makeup water pumps also provide demineralized water supply to Emergency

Feedwater System (LAR) and makeup of Passive Heat Removal Systems (PHRS) tanks via Emergency Makeup System for Passive Heat Removal System Tanks (JNC).

1.5.5.3 Process Auxiliary Systems

1.5.5.3.1 Process and Post-accident Sampling Systems

To enable monitoring of the primary circuit and NI auxiliary systems water chemistry

parameters, the following systems are provided:

– Automated Chemistry Monitoring System of Primary Circuit (KUB)

provides real-time operative monitoring of the primary water chemistry.

– Sampling System of Special Water Treatment Plants and Reactor Plant Auxiliary

Systems (KUA)

enables periodical manual sampling to carry out laboratory analyses on

quality indices of the primary coolant and water media of the NI auxiliary

systems.

– System of Post-Emergency Sampling (KUL)

provides manual sampling from the sump tanks of the emergency low

concentration boron solution storage system (JNK) for conducting

laboratory analysis of emergency core cooling water.

– Sampling System in Severe Accident Conditions (KUL 50)

provides water sampling of the storage pool, sump tanks with low

concentration borated water and the containment steam-gas medium in

accident modes, including severe accidents.

The sampling systems do not perform safety functions, but are important in terms of risk

because the possibility of initiating events resulting from the damage to the systems.

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1.5.5.3.2 Intermediate Cooling Circuit for High-Pressure Important Consumers (KAB)

Functions

The Intermediate Cooling Circuit for High-Pressure Important Consumers (KAB) is designed to:

– supply cooling water and remove heat from the reactor plant auxiliary systems;

– provide a barrier between high-pressure primary coolant and the Intermediate

Cooling Circuit for Important Consumers (KAA).

System description

The KAB system consists of four trains, independent in their process and electrical

connections. Each train contains a pump, heat exchanger, valves and pipelines. Cooling

water is pumped to the consumers, from where it is sent to the intermediate circuit heat

exchangers cooled by Intermediate Cooling Circuit for Important Consumers (KAA). Each KAB train is connected with corresponding KAA train.

1.5.5.3.3 Demineralized Water Supply System (KBC-2)

Functions

The Demineralized Water Supply System is designed to supply demineralized water e.g.

for makeup and flushing purposes to consumers located in the reactor building, safety building, Nuclear Service Building (10UKC) and auxiliary building.

System description

The system consists of piping and valves. Demineralized water to the system is supplied

from the demineralized water storage tanks by the pumps of the Makeup Water System (LCU).

1.5.5.3.4 Steam Supply System of Auxiliary Building (LBG30)

Functions

The Steam Supply System of Auxiliary Building (LBG30) is designed to supply heating

steam from the auxiliary steam header of the turbine building (LBG50) to the consumers of the primary circuit, special water treatment and liquid radioactive waste solidification.

System description

Heating steam is supplied e.g. to the following consumers:

– Makeup and Boron Control System (KBA) deaerator

– Pure Condensate Supply System heat exchangers

– Liquid Radioactive Waste Solidification System (KPC)

– Primary Coolant Treatment System (KBF) evaporators

– Drain Water Treatment System (KPF) after evaporator

Heating steam condensate is cooled by the Intermediate Cooling Circuit for Important

Consumers (KAA) before sending to the Drainage Collection System of Turbine Building

(LCM).

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1.5.5.3.5 Controlled Area Equipment and Floor Drainage Systems

The controlled area equipment and floor drainage systems are designed for collection of

contaminated leakages and drains from the controlled area systems and rooms. Collected

drains are either returned back to the system in question or supplied to the proper storage or treatment systems for further processing.

The following systems are provided:

– Reactor Coolant Leakage Collection System (JET)

– Reactor Building Equipment Drainage System (KTA)

– Boron-Containing Drains Collection System (KTC)

– Reactor Building Special Sewerage System (KTF)

– Auxiliary Building Special Sewerage System (KTH)

– Safety Building Special Sewerage System (KTL)

– Nuclear Service Building Special Sewerage System (KTT)

1.5.5.3.6 RCP Sealing Water System (JEW)

Functions

The JEW system together with the Makeup and Boron Control System (KBA) and

Intermediate Cooling Circuit for High-Pressure Important Consumers (KAB) is designed to

perform the following functions:

– ensure tightness of the primary circuit pressure boundary at the reactor coolant

pump (RCP) set;

– cooldown carbon graphite seal of the RCP shaft.

System description

Under normal operating conditions, sealing water for the reactor coolant pumps is taken

after the heat exchanger of the coolant emergency outlet line (KBA). The heat exchanger

is cooled by the KAB system. After the reactor coolant pumps, the sealing water is discharged into the KBA deaerator via the controlled leakage lines.

In case of failure of heat removal from RCP independent circuit (temperature increase of

independent circuit, or failure in КАА system intermediate circuit), emergency injection of

cooling water is provided into independent circuit from the head of pumps of Makeup and

Boron Control System (KВА).

1.5.5.3.7 Discharge and Monitoring System for Leakages from Borated Water Storage Tank

(JMM)

Functions

The Discharge and Monitoring System for Leakages from Borated Water Storage Tank

(JMM) is designed to monitor the integrity of the containment sump tank (JNK) liner and

tanks of high concentrated borated water (JNK) to detect water in the gap between the

liner and the wall in every liner section.

System description

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The system consists of discharge tubes penetrating the walls of the containment sump and

high concentrated borated water tanks, removing possible leaks from the gap between the

liner and the walls. Each tube is fitted with a three-way valve. The discharge branch pipes

of this valve are connected either to the gathering header with electric-driven valves,

whose outlet is directed to the collecting tank, or to the control funnel. Outlets of the

control funnel also lead to the header and then to the drainage line downstream of the

cut-off electric-driven valves. The leaking section can be determined by switching over the three-way valves and monitoring the filling speed of the collecting tank.

1.5.5.3.8 PHRS Tanks Makeup System (JNC)

Functions

The Passive Heat Removal System (PHRS) Tanks Makeup System (JNC) is designed for emergency makeup of PHRS tanks.

System description

The JNC system consists of two identical and independent trains, each train containing:

– emergency makeup pump;

– isolating valves;

– pipelines.

The pumps are connected to bypass of Makeup Water System (LCU) tanks and used for

pumping water from LCU and Emergency Feedwater System (LAR) tanks to the PHRS tanks (of JNB system).

1.5.5.3.9 Chemical and Volume Control Systems

1.5.5.3.9.1 Makeup and Boron Control System (KBA)

Functions

The main functions of the KBA system are to keep the required physical and chemical

parameters of the primary circuit in different operating modes and to test the primary circuit for strength and tightness.

System description

The KВА system consists of the primary coolant blowdown unit with outlet lines for normal

operation and DBC2, the makeup unit with an inlet line, its equipment and isolation

valves. The KBA system also includes strength and leak tightness hydro-testing lines of the primary circuit and lines for supplying medium to the ECCS accumulator tanks.

During normal operation, the coolant removed from the primary circuit is first cooled down

in regenerative heat exchangers. The cooling is done by makeup water fed back to the

primary circuit via inlet lines. Then it is taken to aftercooler by the Intermediate Cooling

Circuit for High-Pressure Important Consumers (KAB). Temperature of the coolant is

controlled by the flowrate of the KAB cooling water and shall be lower than a permissible

temperature of the Primary Coolant Purification System (KBE) filters. After filtration,

depending on the operation mode of the KBA system, the coolant is either returned into

the inlet line upstream of the regenerative heat exchanger or is sent to be discharged into the deaerator.

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The boron control during different operating modes of the unit is provided by means of

changing boric acid concentration in the primary coolant. Boric acid concentration is

changed by discharging coolant from the primary circuit and replacing it with the pure

condensate from Pure Condensate Supply System (KBC1) or boric acid solution of proper

concentration. Boric acid solution or pure condensate is supplied directly to the suction

manifold of the KBA makeup pumps.

1.5.5.3.9.2 Primary Coolant Purification System (KBE)

Functions

The Primary Coolant Purification System (КВE) is designed to perform the following

functions to maintain primary circuit water chemistry parameters in compliance with standards:

– removal of dissolved anionic and cationic forms of impurities from the primary

coolant;

– removal of radioactive corrosion products from the primary coolant;

– variable control of ammonia-potassium water chemistry conditions.

System description

The KBE system is connected to the Makeup and Boron Control System (KBA) downstream from the after-cooler. The system consists of two parallel trains, each train including:

– mixed-bed filter;

– filter trap;

– pipelines and valves.

Depending on the operating mode of the unit, either one or two trains are used to purify

primary coolant.

1.5.5.3.9.3 Chemicals Preparation and Supply System for Keeping Primary Coolant Chemistry

(KВD-1)

Functions

The Chemicals Preparation and Supply System for Keeping Primary Coolant Chemistry (KBD-1) is designed to:

– reception, preparation and temporary storage of chemicals in required working

concentration;

– supply of the chemicals into the primary circuit.

System description

The KBD-1 consists of tanks, pumps, pipelines and valves for ammonia, hydrazine-

hydrate, potassium hydroxide and zink acetate solution preparation and dosing into the

primary circuit. Dosing pumps supply the chemicals either continuously or periodically to the suction header of the Makeup And Boron Control System (KBA) pumps.

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1.5.5.3.9.4 Coolant Storage System (KBB)

Functions

The Coolant Storage System KBB is designed to perform the following functions:

– receive and store the coolant after primary circuit water exchange;

– maintain the makeup deaerator (KBA) water level;

– initial filling of the primary circuit.

System description

The KBB system consists of coolant storage tanks, coolant removal pumps, heat

exchanger, sumps, ion exchangers, resin trap, valves and pipelines. The system is

designed to receive and store the reactor coolant, which is discharged during normal

operation of the plant as a result of the need to compensate the burn-up during the cycle,

load variations and start-up and shutdown transients.

1.5.5.3.9.5 Pure Condensate Supply System (KBC-1)

Functions

The Pure Condensate Supply System KBC-1 is designed to perform the following functions:

– supply of pure condensate into the primary circuit;

– maintain the Makeup And Boron Control System (KBA) deaerator water level.

System description

The KBC-1 system consists of condensate storage tanks, pumps, heat exchanger, pipelines

and valves. During unit start-up mode, pure condensate is supplied from tanks the

through the heater into the suction line of KBA pumps to decrease boron concentration in

the primary circuit coolant. In boron control mode, pure condensate is supplied either to

the suction line of KBA pumps or to the KBA deaerator, according to the boron control

programs. During cooldown mode, pure condensate pumps transfer boron solution from JNK tanks to the KBA deaerator to compensate primary coolant shrinkage.

1.5.5.3.10 Primary Coolant Treatment System (KBF)

Functions

The Primary Coolant Treatment System KBF is designed to process boron-containing water for reuse either as a condensate or boron concentrate to make up the primary circuit.

System description

The KBF system consists of several equipment groups, which are dedicated to the

following tasks:

– supply of boron-containing water for its treatment;

– evaporation of the boron-containing water;

– collection and purification of the boron concentrate;

– air discharge from filters.

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Boron-containing water from the coolant storage tanks (KBB) is pumped to the evaporator

through the regenerative heat exchanger, where it is preheated. In evaporator, the boric

acid is concentrated and secondary steam from the evaporator is led to the condenser.

The condensate is pumped through the cooler to makeup water storage tanks (KBC-1) or,

if necessary, it can be discharged to the environment through the ion exchange filters and

control tanks of the drain water treatment system (KPF). Non-condensed steam and gas

mixtures from the condenser are directed to vent cooler and after that to the Tank Vent Treatment System (KPL-3).

The boron concentrate from the evaporator is discharged into boron concentrate tank

through the cooler. The boron concentrate is purified before supply to the high

concentration borated water storage sump tanks of system JNK.

1.5.5.3.11 Decontamination System for Equipment and Rooms (FK)

Functions

The Decontamination System For Equipment And Rooms (FK) is designed for

decontamination of equipment and rooms prior to scheduled preventive maintenance and examinations to reduce the operating personnel’s exposure dose during the NPP operation.

System description

The Decontamination System For Equipment And Rooms (FK) consists of movable and

stationary systems. The stationary system is designed for removable equipment

decontamination, and the movable system is designed for decontamination of rooms and stationary equipment; it consists of a complex of movable modular units.

1.5.5.3.12 Process Compressed Air Supply System (QEB)

Functions

The process compressed air supply system is designed to perform the following functions:

– supply compressed air to the consumers of the whole unit;

– supply compressed air for containment testing.

System description

The process compressed air supply system includes distribution pipelines and isolating, control and safety valves. Air supply is provided from the compressor station (QEA).

1.5.5.3.13 Group of Equipment for The Special Laundry Water Collection (as a part of system

SRP)

Functions

The group of equipment is designed for collection and monitoring of the spent special laundry water.

System description

Spent washing water is collected in the water collection tanks of the special laundry. When

a tank has been filled, a water sample is taken and analyzed in the laboratory to

determine its radionuclide composition. If water quality is found to be satisfactory, the

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water is sent to the sanitary wastewater treatment works of the controlled access area. If

the allowable values are exceeded, the water is sent for re-purification with the ion-selective filters of the Drain Water Treatment System (KPF).

1.5.5.4 Heating, Ventilation and Air Conditioning (HVAC) Systems

Functions

The HVAC systems are designed to perform the following functions:

– maintain optimal operating conditions for mechanical and electrical equipment

and instrumentation and control systems;

– ensure allowable working conditions for operation and maintenance personnel

during normal and post-accident NPP operation modes.

In addition to that, the controlled area HVAC systems are designed to:

– reduce radioactive substances content in the air

– prevent spreading of radioactive substances from more contaminated areas to

less contaminated or clean areas of the plant

– limit radioactive substances release into the environment.

System description

NPP buildings and facilities are divided into two groups according to the external dose rate, surface contamination and airborne radionuclide concentration:

– controlled area where possibility of radiation effect on operation and maintenance

personnel is not excluded;

– uncontrolled area where possibility of radiation effect on operation and

maintenance personnel is in practice excluded.

The design approach of HVAC system arrangement is based on separate systems for

controlled and uncontrolled areas. The rooms in the controlled area are divided into three

zones based on external dose rate, surface contamination and airborne radionuclide

concentration. Predetermined pressure differences between these zones are maintained by

HVAC systems in order to ensure that air always flows from the clean areas towards the

less clean areas to maintain radiation safety. Controlled area exhaust air is filtered before

release to the environment in a 100-meter high ventilation stack.

The HVAC systems are organized into five groups as follows:

1. Systems designed to perform normal operation functions:

– Containment recirculation air cooling systems (KLA10, 20, 30, 50, 60, 80)

– Containment recirculation air cleaning system (KLA13)

– Auxiliary Building recirculation air cooling system (KLE40)

– Electrical and I&C rooms HVAC systems (SAC50, 51, 52, 70, 71)

– NI unit diesel-generator station HVAC systems (SAD60, 61, 64, 65, 66, 67, 70,

71, 73, 74, 75, 76, 77)

– TI unit diesel-generator station HVAC systems (SAD80, 81, 83, 84, 85, 86, 87)

– Normal operation power supply building HVAC systems (SAL02, 12, 14, 16)

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– Makeup water system (LCU) pump room HVAC system (SAS02)

– Refueling machine control room HVAC system (SAT02)

2. Systems designed to ensure safety system equipment operating conditions:

– Reactor building annulus space HVAC systems (KLC10, 20, 30, 40; KLC13, 23,

33, 43)

– Safety building HVAC systems (KLG10, 20, 30, 40; KLG11, 21, 31, 41)

– Control Building HVAC systems (SAA10, 20, 30, 40; SAA11, 25, 35, 45)

– Passive containment heat removal system (JMP) safety train rooms cooling

system (SAA11, 21, 31, 41)

– Main Control Room HVAC system (SAC12, 22, 32, 42)

– Emergency Control Room HVAC system (SAC17, 27, 37, 47)

– Pumping Station of the Essential Consumers HVAC systems (SAG10, 20, 30, 40;

SAG11, 21, 31, 41)

– Standby diesel generator station building HVAC systems (SAD10, 20, 30, 40;

SAD11, 21, 31, 41; SAD12, 22, 32, 42; SAD13, 23, 33, 43; SAD14, 24, 34, 44;

SAD15, 25, 35, 45; SAD16, 26, 36, 46)

– Steam cell safety system rooms HVAC systems (SAS10, 20, 30, 40; SAS11, 12,

13, 14)

3. Systems designed to provide chilling medium for HVAC systems:

– Control Building chilling medium supply systems (QKC10, 20, 30, 40)

– Standby Diesel Generator Station Building chilling medium supply system

(QKD10, 20, 30, 40

– Steam Cell chilling medium supply system (QKS10, 20, 30, 40)

– Main Control Room chilling medium supply system (QKC12, 22, 32, 42)

– Emergency Control Room chilling medium supply system (QKC17, 27, 37, 47)

– NI unit diesel generator station chilling medium supply system (QKD60, 70)

– TI unit diesel generator station chilling medium supply system (QKD80)

4. Systems designed to limit radioactive substances release into the environment:

– Controlled area exhaust air systems (KLE20, 30)

– Containment exhaust air system (KLD11)

– Containment annulus area exhaust air system (KLC12, 22, 32, 42)

– Safety Building exhaust air system (KLG12, 22, 32, 42)

5. Systems ensuring personnel radiation safety:

– Main Control Room filtered air intake system (SAC11, 21, 41)

– Emergency Control Room filtered air intake system (SAC16, 26, 46)

– Containment Purging system (KLD20, 21)

1.5.5.5 Radiation Monitoring Systems

This description is focused on the on-site radiation monitoring systems. In addition to the

on-site monitoring, the environment of the unit will be monitored through radiation measurements.

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Functions

The radiation monitoring system provides verification for the conformance of the unit

operating radiological conditions with the operating technical specifications concerning

integrity of the physical barriers, radiation levels and radioactive releases as well as surveillance of and protection against the personnel radiation exposures.

Radiation monitoring is performed under all Design Basis Conditions (DBC), Design

Extension Conditions (DEC) and Severe Accidents (SA) as well as during NPP’s

decommissioning.

The key targets to establish radiation monitoring system are as follows:

– radiation safety assurance for the operating personnel and the public inhabiting

the NPP effective areas as well as improving the reliability of radiation safety

assurance through early detection of deviation from normal mode of the process

equipment operation, which affects radiation safety and identification and

elimination of the reasons for deviation;

– warning and control signals generation as soon as radiation parameter values

have exceeded the preset limits, in violation of normal operating conditions upon

loss of integrity in protective barriers, as well as providing recommendations to

the operating personnel for eliminating the reasons that caused this exceedance;

– provision of operational evaluation and dose load prediction and well-timed

guaranteed provision of recommendations to the public on protective measures

needed at any phase of an accident;

– monitoring of compliance with the radiation safety regulations and rules at all

NPP life cycle phases: commissioning, operation and decommissioning.

System description

The on-site radiation monitoring system (RMS) is comprised of automated radiation

monitoring system (ARMS) and separate auxiliary equipment and means supporting the

ARMS measures (mobile monitoring means, laboratory complexes, sampling systems, independent stationary instruments etc.). ARMS is composed of the following subsystems:

– Automated process radiation monitoring system (APRMS);

– Automated system of radiation situation monitoring in rooms and on site

(ASRMS-R&S);

– Automated radiation contamination monitoring system (ARCMS);

– Automated individual radiation dose monitoring system (AIRDMS).

The ARMS is a hierarchical structure of two-level information measuring system.

At the lower level, there are intelligent detection devices and actuator control units. The

following actions are performed at the lower level: information is collected from different

types of measuring channels and is primarily processed, actuators are controlled, data is

generated and transmitted to the upper level of the system, and information is submitted and stored locally.

At the upper level, the data from lower level devices is collected, ARMS equipment is

controlled, information on NPP radiation state is processed, displayed and presented to different users under different NPP operation modes.

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ARMS operates exchanges information continuously with the plant's process

instrumentation and control systems and plant environment radiation monitoring system.

1.5.5.6 Other Auxiliary Systems

1.5.5.6.1 Communication Systems

Hanhikivi-1 NPP will have various communication and warning facilities, which support

both daily operations management and emergency situation activities at the NPP.

The external communication facilities are designed to ensure communication with relevant external parties, e.g. authorities.

The external communication facilities incorporate the following communication systems:

– direct radio communication system with the authorities (Virve);

– satellite communication system;

– grid communication system.

The internal communication facilities are designed to provide on-line operating and office

administrative communication for the NPP staff during their day-to-day activities and emergency situations.

The internal communication facilities may incorporate systems for the following types of communication functions:

– operative two-way communications, including portable equipment, both in field

and in control room

– personnel alarm and search

– operating talks recording system

– various telephone systems

– clock system

– industrial television system (CCTV)

Each internal communication system is built, as a rule, based on its own technical

requirements and facilities, which provides independence and reliability as a whole.

As a rule, networks of personnel alarm and search systems cover all places where

personnel is likely to be. Networks of wireless telephone communication system cover all

necessary premises of the NPP and are only limited by EMC requirements and technical

and economic feasibility. Networks of other communication systems cover basic work places of the personnel and basic places where personnel is likely to be.

1.5.5.6.2 Lighting Systems

Lighting systems are intended to provide efficient lighting in the buildings, rooms, working

areas and NPP site. The following types of lighting for the NPP are planned:

– normal lighting is provided in all the rooms and spaces to enable work execution,

service personnel attendance and passage as well as traffic;

– emergency lighting is provided in the areas requiring maintenance operations

during an emergency outage of the regular lighting and to facilitate safe

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movement of the personnel when regular lighting fails; it is also intended for

firefighting measures;

– escape route lighting is intended for maintenance personnel evacuation from the

premises during an emergency outage of the regular lighting; it is provided along

the escape routes, illuminating the main passages and stairs;

– in case of a severe accident, lighting is provided in MCR, ECR and local control

centers;

– obstruction lighting of high structures assure air traffic safety over the NPP area;

– outdoor lighting assures free movement of people and transport over the

territory, operations needed to maintain process equipment at the sites outside

the buildings.

Power supply sources for the lighting system are 0.4 kV switchgears:

– Normal Operation Power Supply Systems;

– Reliable Power Supply Systems for Nuclear Island and Turbine Island;

– Emergency Power Supply Systems (EPSS).

Under normal operation conditions at power operation of the unit, power is supplied to the

lighting system from switchgear of the normal operation system. At anticipated operational

occurrences (blackout), the lighting system is supplied with power from switchgear 0.4 kV,

groups 2 and 1, Reliable Power Supply System and EPSS having independent power

sources: diesel generators and accumulator batteries.

1.5.5.6.3 Diesel Generator Auxiliary Systems

The following diesel generator units are provided:

– Four Emergency Power Supply System (EPSS) diesel generator units;

– Two NI diesel generator units for Reliable Power Supply System (RPSS of NI);

– Two severe accidents management diesel generator units;

– One TI diesel generator unit for Reliable Power Supply System (RPSS of TI).

In order to ensure the operation of diesel generator systems, each diesel generator unit is equipped with the following auxiliary systems:

– diesel fuel storage and supply system XJN;

– lubricating system XJV;

– starting air system XJP;

– cooling system XJG;

– air intake system XJQ and exhaust gas system XJR;

– system of collection and discharge of fuel and oil leakages GMF;

– power supply system;

– I&C system;

– heating, ventilation and air conditioning systems (HVAC).

1.5.5.6.4 Fire Protection Systems

The NPP unit fire protection systems consist of outdoor and indoor fire hydrants as well as

automatic water and gas firefighting units. Water for these systems is supplied from fire-

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fighting pumping station (10USG) by Fire-Fighting Water Supply System (SGA). The SGA

system consists of water storage tanks, firefighting pumps, ring networks with partitioning and isolation valves and firefighting equipment, such as fire hydrants, sprinklers etc.

In normal operation conditions, when there is no fire, the system is filled with water and is

pressurized by the holding pumps of the SGA system. In the event of fire, the firefighting

pump is actuated automatically and the holding pumps are switched off. In case of failure

of the operating pump, the standby pump is actuated automatically. The water supply into

the ring networks can also be arranged by mobile firefighting equipment connected to the coupling heads of the water supply pipelines.

The automatic water fire-fighting systems (SGB and SGD) are designed to protect oil-filled

equipment and main feedwater pumps, diesel generator station equipment and rooms,

cable rooms, transformer units and storage rooms. Rooms containing electronic equipment are protected by the Automatic Gas Fire-Fighting System (SGE).

1.5.6 Steam and Power Conversion System

The steam and power conversion system removes energy from the reactor coolant in the

steam generators and converts it into electric power in the turbine generator. Main parameters of the Steam and Power Conversion System are presented below:

Gross electric power of the unit, MWe 1 265.05

Main steam rated flow, kg/s 1 783.05

Rated parameters of main steam upstream of the turbine:

Pressure (abs), MPa 6.8

Temperature, °C 283.9

Humidity, % 0.5

Design temperature of cooling water, °C 4

Feedwater temperature, °C 225±5

1.5.6.1 Turbine Generator

Functions

The function of the Turbine Generator (TG) is to convert the thermal energy supplied by the Main Steam System into electrical energy.

System description

The TG package consists of a 1500 rpm, single-flow high pressure (HP) turbine and a

single-flow intermediate pressure (IP) turbine in a common casing and three double-flow

low pressure (LP) turbines in tandem. Moisture separation and reheating of the steam is

provided between the HP turbine and IP turbine by two combined moisture separator reheater (MSR) assemblies. The MSRs have two stages of reheating.

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The generator is a four-pole machine, directly driven by the ,turbine and supplies the step-

up transformer with high voltage electrical output. The rotor winding is cooled by hydrogen gas and the stator winding by an internal circulation of stator cooling water.

Steam generated in the four SGs is supplied by the Main Steam System (LBA/LBU) to the

high pressure (HP) turbine through four units of stop and control valves, which regulate

steam flow. After expanding across the HP turbine blading, exhaust steam is reheated in

two moisture separator reheaters (MSR). The MSRs supply the intermediate pressure (IP)

turbine through stop and intercept valves. After expanding across the IP turbine blading,

exhaust steam flows to the three low pressure (LP) turbines. The main condenser

condenses the LP turbine exhaust and transfers the heat rejected in the cycle to the Main

Circulating Water System (PA).

Extraction steam from the HP turbine supplies the 6th and 7th stages of feedwater heating

as well as the heating steam to the first stage reheaters (MSR). Most of the HP turbine

exhaust steam is routed to the MSR inlet, but part of it is diverted and supplies the 5th

stage of feedwater heating (deaerator).

Moisture removed in moisture separators is drained to the moisture separator drain tanks

and then pumped to the deaerator/feedwater storage tank. The dry steam passes across

two stages of reheaters, which are supplied with HP turbine extraction steam (first

reheating stage) and main steam (second reheating stage). The first stage reheater drains

are discharged to the 6th stage feedwater heaters (LAD), and the second stage reheaters

drains to the 7th stage feedwater heaters (LAD). Emergency drainage to the main condenser (MAG) is also possible.

After removal of the water content and reheating in the MSRs, the steam is directed to the

IP part of the HP/IP turbine module. Extraction steam from the IP turbine supplies the 4th

stage of feedwater heating. Most of the IP turbine exhaust steam is routed to the LP turbine inlet, but part of it is diverted and supplies the 3rd stage of feedwater heating.

Extraction steam from the LP turbine supplies the 1st and 2nd stages of feedwater heating.

1.5.6.2 Main Steam System (LBA/LBU)

Functions

The Main Steam System (LBA) is intended for main steam supply from steam generators

to turbine generator, main condensers (MAG) via Turbine Bypass System (MAN), MSRs

second stage reheaters (LBB) and Auxiliary Steam System (LBG). The system also

includes main steam line isolation and secondary circuit overpressure protection valves (LBU).

System description

Main steam from steam generators (SG) is supplied to turbine stop and control valves via

four pipelines. Main steam header (MSH) is arranged for pressure equalization and steam

supply to Turbine Bypass System (MAN), MSRs second stage reheaters (LBB) and Auxiliary

Steam System (LBG) upstream of turbine valves.

In order to ensure overpressure protection of the Main Steam System (LBA) and the

Steam Generators (SG) as well as residual heat removal to atmosphere and SG secondary

side isolation in transient and accident conditions, each main steam pipeline from SG is

provided with main steam valve unit, which consists of two main steam safety valves

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(MSSV), fast-acting atmospheric steam dump valve (BRU-A) with isolation valve and fast-

acting main steam isolation valve (MSIV).

1.5.6.3 Main Condenser (MAG)

Functions

The Main Condenser (MAG) functions as the steam cycle heat sink, condensing steam from

the main turbine or from the Turbine Bypass System (MAN), It also collects drains and

non-condensable gases from various sources and deaerates primary condensate and normal makeup water.

System description

The main condenser is divided into three separate shells. Each shell is located beneath its

respective LP turbine. Each condenser shell contains duplex 1st and 2nd stage LP feedwater

heaters and Turbine Bypass System steam receivers installed on the condenser neck, tube bundles, condensate hotwells and cooling water chambers.

During normal operation, the main condenser is operated under a vacuum, and steam

from the exhaust of the low pressure turbines is expanded down into the main condenser

shells across the main condenser tubes and is condensed and collected in the hotwells.

The main condenser also serves as a collection point for steam, demineralized water, equipment drains, extracted water and vented air from other systems.

Main condenser vacuum is maintained by the Condenser Vacuumizing System (MAJ). Air

in-leakage and non-condensable gases contained in the turbine exhaust steam are

collected in the main condenser and removed by the vacuum pumps and steam jet air

ejectors (three pumps running and one stand-by) of the MAJ system. Non-condensable

gases are removed from the system into atmosphere. The exhaust line is provided with radiological activity monitoring.

The condenser is cooled by Main Circulating Water System (PA). Even distribution of the

cooling water between condenser tubes is ensured by the shape of the water chambers.

The condenser pipes are expanded and welded on both sides to the tube plates, which are

welded to the condenser housing and have bolt connection with water chambers. Water

chambers are designed to ensure filling top tubes with water during all starting operations

and under normal operating modes. The Condenser Ball Cleaning System (PAH) provides cleaning of condenser pipe inner surfaces with rubber balls during unit operation.

1.5.6.4 Turbine Bypass System (MAN)

Functions

The Turbine Bypass System (MAN) is designed for steam discharge from steam generators into the turbine condensers through steam dump valves (BRU-K) in case of:

– start-ups/shutdowns of the unit and initial loading of the turbine;

– reactor trips to prevent actuation of main steam Atmospheric Steam Dump (BRU-

A) or Safety Valves (MSSV);

– reactor cooldown at the specified rate;

– pressure maintenance in main steam header.

System description

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The system consists of bypass valves of the fast acting pressure reducing station (BRU-K)

equipped with desuperheaters (condensate injection), steam receivers inside the

condenser shells, pipelines and valves. At the unit start-up, the Turbine Bypass System

(MAN) provides heating of main steam pipelines at a certain rate. The BRU-K valves open as steam capacity of SG increases to maintain nominal pressure in MSH.

In case of the turbine load shedding or other transients causing a pressure increase in

MSH, BRU-Ks open, while steam is partially discharged into condenser MAG. Furthermore,

BRU-Ks are closed upon reaching a certain setpoint with the pressure decreasing. Loss of

vacuum in condenser primarily leads to a turbine trip and secondarily to closing of the Turbine Bypass System (MAN) valves.

Under reactor cooldown conditions, the BRU-K controller is switched by operator to

cooldown mode with a certain rate, thus maintaining the specified condition of pressure and temperature reduction.

1.5.6.5 Main Circulating Water System (PA)

Description of the Main Circulating Water System (PA) is presented in chapter 1.5.5.2.1.1.

1.5.6.6 Condensate and Feedwater systems

The Condensate and Feedwater Systems provide feedwater to the steam generators (SG)

at the required temperature, pressure and flow rate. Condensate is pumped from the main

condenser (MAG) hotwells by the condensate pumps (LCB); it passes through the low

pressure (LP) feedwater heaters (LCC) and the deaerator feedwater storage tank (LAA) to

the main feedwater pumps (LAC), and is then pumped through the high pressure (HP)

feedwater heaters (LAD) to the SGs. The Condensate and Feedwater Systems include a

number of stages of regenerative feedwater heating and provisions for maintaining

feedwater quality. The Condensate and Feedwater Systems are connected with extraction

piping to the steam turbines. The Condensate and Feedwater Systems ends at the

feedwater heater vents and drains as well as drains from the moisture separator reheaters

(MSR).

1.5.6.6.1 Main Condensate Pipeline System (LCA, LCB)

Functions

The Main Condensate Pipeline System (LCA, LCB) is designed to:

– supply condensate from the main condenser hotwells (MAG) to deaerator-

feedwater storage tank (LAA) via gland seal steam condenser (GSSC), Unit

Demineralization Plant (LD) and Low Pressure Heaters System (LCC);

– control condensate flow to maintain level in deaerator feedwater storage tank

(LAA) within preset limits;

– ensure condensate flow through the recirculation line back to the main condenser

required for condensate pumps normal operation, while the turbine is in idle run

or at low loads;

– supply spray water to the steam dump valves to turbine condenser (BRU-K).

– supply condensate for the LP turbines exhaust components cooling.

System description

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The system consists of condensate pumps (LCB), pipelines and valves (LCA). The main

condensate is discharged from condenser hotwells by booster condensate extraction

pumps (two pumps running and one standby). The booster condensate extraction pumps

supply the condensate through gland seal steam condenser (GSSC) and Unit

Demineralization Plant (LD) to main condensate extraction pump suction. The main

condensate extraction pumps (two pumps running and one standby) supply the main

condensate through regulating unit and low pressure heaters LPH-1/LPH-2, LPH-3, LPH-4

into the deaerator feedwater storage tank (LAA). Condensate recirculation line back into

the condenser downstream of the main condensate extraction pumps is installed to

provide stable operation of the condensate pumps (LCB) at start-up and low load modes. Low pressure heaters LPH-3 and LPH-4 are equipped with bypass lines.

The regulating unit is installed on the main condensate (LCA) pipeline to maintain the level

in deaerator feedwater storage tank (LAA). Regulating unit comprises three control valves,

one of which starts up with approximately 30% capacity and two others with 70% capacity each. The unit full load is provided by opening two main valves.

1.5.6.6.2 Feedwater System (LAA, LAB, LAC)

Functions

The feedwater system is intended to perform the following functions:

– initial filling of steam generators with feedwater;

– feedwater storage and deaeration within the range of dissolved oxygen content

required for SG;

– steam generator level control by providing feedwater during normal operation

and anticipated operational occurrences;

– regenerative contact heating of feedwater;

– cooling down of the reactor plant at the steam stage;

– feedwater line isolation in case of feedwater line rupture or steam generator

overfilling;

– isolation of feedwater supply to the failed steam generator, under the conditions

of uncontrolled steam removal from steam generators and in the case of

primary-to-secondary circuit leaks.

System description

The Feedwater System consists of a deaerator feedwater storage tank (LAA), booster and

main feedwater pumps (LAC) (three pumps running and one stand-by), pipelines and

valves. During power operation of the unit, the system provides degassing and supply of

feedwater to SGs at the required temperature, pressure and flowrate. The main feedwater

pumps take water from the feedwater storage and deaeration tank (LAA) and supply it

through the High Pressure Heaters System (LAD) to the SGs. The main control valves

control the required level in the SGs. Steam is utilized in the deaerator to remove non-

condensable gases and to heat feedwater. Primarily steam is supplied from the HP turbine

exhaust during normal operation or from the main steam header through the steam dump valves to deaerator (BRU-D) during transient or low load conditions.

The Feedwater System provides safety-related isolation functions in case of a feedwater

line rupture, steam generator overfilling, uncontrolled steam removal from steam

generators and primary-to-secondary circuit leaks.

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1.5.6.6.3 High Pressure Heaters System (LAD, LBQ, LCH)

Functions

The High Pressure Heater System (LAD) is designed for regenerative heating of feedwater in the high pressure heaters.

System description

The High Pressure Heater System (LAD) consists of two parallel strings of 6th and 7th stage

feedwater heaters, pipelines and valves. Heating steam to the heaters is supplied from HP

turbine extractions (LBQ). The extraction lines are provided with pneumatic assisted check valves as well as gate valves with a bypass line before each high pressure heater.

On the water side, each string of 6th and 7th stage feedwater heaters is designed to

operate at 50% of its delivery capacity. They are equipped with a common bypass line.

When the other heater string is isolated under full load, the remaining group receives 50%

and the bypass line 50% of the nominal feedwater flow. If necessary, both strings can also be bypassed at the same time.

High pressure heater condensates are cascaded from the higher pressure heaters through

lower pressure heaters into the feedwater storage tank (LAA) or via emergency extraction

line into the main condenser (MAG).

1.5.6.6.4 Auxiliary Feedwater System (LAH, LAJ)

Functions

The Auxiliary Feedwater System (LAH, LAJ) is designed to provide steam generators with

feedwater during unit start-up, shutdown, cooldown and anticipated operational occurrences.

System description

The Auxiliary Feedwater System (LAH, LAJ) consists of two auxiliary feedwater pumps

(LAJ), pipelines and valves (LAH). The auxiliary feedwater pumps feed steam generators in

plant heating-up and cooling down modes. The auxiliary feedwater pumps are connected

to the deaerator feedwater storage tank (LAA) and supply feedwater to the header

downstream from the high pressure heaters during cooldown, shutdown or SG level

control operations. During unit start-up (for the heating of pipelines) or when the unit is

maintained in hot standby condition, the auxiliary feedwater pumps supply feedwater to the pressure header of the main feedwater pumps of system (LAB).

1.5.6.7 Auxiliary Steam System (LBG)

Functions

The auxiliary steam system (LBG) is designed to:

– supply steam to the auxiliary header under all operating conditions of the unit;

– provide redundancy for turbine extractions during operation under load

variations;

– supply steam to the consumers located at the Nuclear Island (NI).

System description

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The Auxiliary Steam System (LBG) consists of auxiliary steam headers, steam dump

valves to deaerator and auxiliaries (BRU-d and BRU-SN), pilot operated safety valves, isolation and control valves and pipelines.

During operation at nominal power, the deaerator feedwater storage tank (LAA) and other

consumers are fed with steam from uncontrolled turbine extraction (LBQ). Under operation

with decreased loads, start-up, shutdown and cooldown conditions, the feeding of the

system is changed over to BRU-SN and BRU-D from main steam header (LBA).

For feeding consumers during unit start-up, the design provides steam supply to the

auxiliary steam system from the plant header, where the steam shall be supplied from the start-up and standby boiler house.

1.5.6.8 High Pressure Steam Line Drains System (MAL30)

Functions

The High Pressure Steam Line Drains System (MAL30) is intended for:

– reception of high pressure steam line drains under start-up, shutdown and

normal operation conditions;

– pressure maintenance in main steam header (LBA) under low load conditions

with closed steam dump valves to turbine condenser (BRU-K);

System description

The High Pressure Steam Line Drains System (MAL30) consists of an expansion tank,

pipelines, isolation and control valves. Drainage expansion tank is intended for steam

condensate removal from high pressure steam pipelines under start-up and shutdown

conditions, for reception of permanent drains from the dead-end sections of pipelines

under power operation, as well as for discharge of excessive steam from main steam

pipelines under start-up and shutdown conditions with steam dump valves to turbine condenser (BRU-K) being closed.

Steam and condensate are removed from the expansion tank to the turbine building

drainage expansion tank at the initial stage of heating high pressure steam pipelines under

unit start-up conditions. With increasing pressure in steam pipelines and availability of

vacuum in the condenser, the condensate removal is changed over to the main condenser

(MAG). Under nominal operation conditions of the unit, the condensate of high pressure

steam pipelines is removed to deaerator feedwater storage tank (LAA).

In all the condensate removal lines, which maintain the heating rate, the pressure in main

steam pipelines or the specified condensate level in the expansion tank depending on unit operation conditions, the provision for control valves is made.

1.5.7 Radioactive Waste Management

1.5.7.1 Liquid Waste Management Systems

1.5.7.1.1 Drain Water Treatment System (KPF)

Functions

The Drain Water Treatment System (KPF) is intended to collect and treat liquid radioactive

media generated in the NPP operation to produce residuum for subsequent disposal and

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condensate for subsequent release to the environment. A possibility to recycle condensate

back to the NPP cycle is provided.

System description

The drain water treatment system consists of a few functionally interconnected equipment

groups as well as pipelines and valves.

The group of equipment, intended for drain water collection and supply for processing:

– drain water tank sump, sump pumps and pump for leak removal from the drain

water sump room;

– drain water tanks, drain water tanks pump and mixing pump, pump for leak

removal from the drain water tanks room.

A group of equipment intended for cleaning of drain water from mechanical impurities:

– hydrocyclones, strainers, pipelines and valves.

The group of equipment, included in the evaporator unit:

– evaporator, after-evaporator, condensers, condensate tank, blow-off cooler,

condensate pump, pump for non-concentrated waste from the after-evaporator

back into the drain water tanks.

The group of equipment, intended for evaporator condensate after-purification:

– condensate cooler, condensate filters, trap filter, control tanks, control tank

pumps, pump for leak removal from the control tanks room.

The group of equipment intended for purification of water with low specific activity and its

removal from the controlled access area:

– low specific activity medium receiving tanks, control tank, strainers, ion-selective

filters, trap filter, low specific activity medium receiving tank pumps, control tank

pump, pump for leak removal from the low specific activity medium receipt tank

and control tank room.

The contaminated drain water from the Auxiliary Building, Reactor Building, Nuclear

Service Building (10UKC) and Safety Building special sewerage systems is collected in the

drain water tank sump. The tank is divided into “clean” and “contaminated” cells by a

separating wall. Water with mechanical solid particles is supplied to the “contaminated”

cell, where particles are preseparated by gravity. The sediment from the "contaminated"

cell is periodically pumped out with the KPK system pump to the Liquid Radioactive Waste

Solidification System (KPC). The purified drain water flows to the “clean” cell; from there it

is pumped through the hydrocyclone to separate mechanical solid particles from the water.

Solid particles are periodically discharged as sludge from the hydrocyclone into the KРС

solidification system tanks. After mechanical purification, drain water flows to drain water

tanks (KРF20).

By means of drain water tank pump, drain water is supplied for processing into main and

after evaporators through the strainer. Solution, evaporated in after-evaporator, is

discharged into the Liquid Radioactive Waste Storage System (KРK) tanks. Secondary

steam from the evaporator is supplied into the condenser, from which condensate flows by

gravity into the condensate tank. From the tank, condensate is pumped through cooler

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and filters into one of the control tanks КPF40. After monitoring in the control tanks, the

condensate is pumped to neutralization tanks of Flushing Water Removal and

Neutralization System (GCR) to be then discharged into the environment. If necessary,

provision is made for condensate supply to Pure Condensate Supply System (KBC-1)

storage tanks for reuse in the NPP cycle, to preoperational cleanup water tanks of the

System for Preoperational Rinsing of Condensate Feeding Line (LDT) or back to the drain

water storage tanks KРF20 for secondary processing, if the condensate quality is insufficient.

Low specific activity water of the Nuclear Service Building (10UKC), Auxiliary Building and

discharge water of the turbine Condensate Purification System filters (if active) is collected

in low specific activity medium receiving tanks KРF60. After monitoring, water is supplied

through ion-selective filters to the control tank KPF60. In case the water activity in control

tank is lower than the criteria set for discharges into environment, the solution is removed

beyond the controlled area.

1.5.7.1.2 Liquid Radioactive Waste Storage System (KPK)

Functions

The Liquid Radioactive Waste Storage System (KPK) is intended for storing liquid

radioactive waste.

System description

The system consists of four subsystems:

– KPK10 – reception of evaporator bottoms from Drain Water Treatment System

(KPF) and transfer to Liquid Radioactive Waste Solidification System (KРС);

two evaporator bottoms reception tanks, two hose-type pumps, pipelines

and valves.

– KPK20,30 – spent ion-exchange resins and ion-selective sorbents reception and

transfer to Liquid Radioactive Waste Solidification System (KРС);

two reception tanks for medium-active sorbents, one reception tank for

low active sorbents, two hose-type pumps, pipelines and valves.

– KPK60 – reception and monitoring of spent low active resins;

intermediate vessel for ion-exchange resins, pipelines and valves.

– KPK70 – dehydration and removal of nonradioactive spent ion-exchange resins.

distribution funnel, decanter, pipelines and valves.

Provision is made for one standby tank, which is designed for evaporator bottoms or

sorbent slurry reception, when the system tanks have been completely filled or when one of the said tanks needs repair.

Evaporator bottoms from the Drain Water Treatment System (KPF) evaporators flow by

gravity to evaporator bottom tanks. Supply pipeline is blown down by heating steam from

the Auxiliary Building Steam Supply System (LBG30) through Drain Water Treatment System (KPF).

Ion-exchange resins from the Purification System of Water in Fuel Pool and Borated Water

Storage Tanks (FAL), Coolant Storage System (KBB), Primary Coolant Purification System

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(KBE) and Primary Coolant Treatment System (KВF) enter the medium-active sorbents

tanks, where they remain for three months for short-life radionuclides decay.

To decrease the amount of liquid radioactive waste supplied for solidification, provision is

made for a line of spent sorbents supply from the Steam Generator Blow-Down System

(LCQ-2), Drain Water Treatment System (KPF) and Intermediate Cooling Circuit for

Important Consumers (KAA) filters to intermediate ion-exchange resin vessel. After being

subject to radiation monitoring and dehydrating in decanter under laboratory conditions,

the sorbents are removed from the controlled access area for further utilization outside the NPP site or directed to low active sorbents tank.

Evaporator bottoms and ion-exchange resins are supplied for solidification by transfer to

the Liquid Radioactive Waste Solidification System (KPC) reception tank. If the salinity of

evaporator bottoms is below the design value, a return to drain water tanks (KPF20) is provided.

1.5.7.1.3 Liquid Radioactive Waste Solidification System (KPC)

Functions

The Liquid Radioactive Waste Solidification System (cementation plant) is intended to treat

and package NPP-generated liquid radioactive waste in normal and anticipated operational

occurrences, to assure operation and maintenance personnel radiation safety and to

prevent radioactive contamination of the environment in radioactive waste handling.

System description

The LRW solidification plant includes the following units:

– evaporator bottoms concentration and portion metering in the volume mixing

tank;

– reception and concentration of ion-exchange slurry and sludge, as well as portion

metering in the volume mixing tank;

– reception of cement and dry aids (bentonite clay), their mixing and metering of

the mixture into the volume mixing tank;

– ion-exchange sorbents dehydration;

– cement compound generation;

– cement compound and sorbents packing into containers, container sealing and

storage of filled containers;

– manual (remote) and automatic control system.

The initial evaporator bottoms from the Liquid Radioactive Waste Storage System (KPK)

are supplied via the monte-jus through the flowmeter into the evaporator, where it is

further concentrated. Separation of concentrate from the steam-gas phase occurs in the

cyclone. The evaporator bottom concentrate is collected in monte-jus, while the other monte-juses supply portions of evaporator bottom concentrate to the measuring tank.

When using the unit for ion-exchange sorbents dehydration, slurry from monte-jus is

supplied to a centrifuge. The produced residue is supplied to the mixing tank.

A portion of the cement-aid mixture is supplied by the feeder into the measuring tank. The

process is monitored based on indication of the measuring tank weighing device as well as

on the level indicator. The mass portion is defined according to the amount and

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composition of the liquid radioactive waste supplied to the mixer. Mixing of liquid

radioactive waste and the mixture of cement with the processing aid is carried out inside the mixer.

After preparation of a homogenous mixture in the mixer, the hose gate and the mixture

valve are opened to fill the container. The design of the solidification unit provides the final product, packed into containers ready for continuous storage.

1.5.7.2 Gaseous Waste Handling Systems

To clean radioactive gas vents from the tanks containing liquid radioactive media and

bring them down to the permissible values before release into the environment via ventilation stack, the following systems are provided:

– Hydrogen Burning System (KPL1)

– Radioactive Gas Treatment System (KPL2)

– Tank Vent Treatment System (KPL3)

The gas clean-up is carried out with charcoal absorbers as well as aerosol and iodine

filters. To ensure explosion and fire safety, the vented gas mixtures containing hydrogen

are supplied to the Hydrogen Burning System (KPL1), where hydrogen is burned by catalytic recombiners before sending on to gas treatment systems.

1.5.7.3 Solid Radioactive Waste Handling System (KPA)

Functions

The Solid Radioactive Waste Handling System (KPA) is intended for collection, sorting,

treatment, packaging and disposal of solid and solidified liquid radioactive waste produced

during normal operation of the NPP, while performing repairs, as a result of an accident

and during decommissioning of the NPP.

System description

Two types of solid radioactive waste are formed during normal operation and outages of a nuclear power plant:

– solid radioactive waste:

parts and equipment extracted from the reactor;

contaminated dismantled equipment, pipelines and valves;

contaminated tools and repair appliances;

contaminated spent aerosol filters of the ventilation and gas purification

systems;

contaminated special clothing, shoes, individual protection gear which

cannot be decontaminated;

contaminated construction and heat insulation materials;

contaminated wiping material.

– solidified radioactive waste.

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Solid radioactive waste is collected and sorted according to level of activity and processing

methods by loading it into containers or non-returnable packages. For preprocessing of

low and medium activity solid radioactive waste, the design provides pressing technology,

a burning process as well as equipment for fragmentation of long-length pieces and for handling ventilation system filters.

Pressed, vacuum-packed and categorized low and medium activity solid radioactive waste

is stored at the nuclear power plant for ten years. Drums (containers) with low and

medium activity solid radioactive waste are placed in a solid radioactive waste storage.

After the expiry of storage time, the drums are removed from the plant to the final repository or cleared from the regulatory control, based on the activity level.

High activity solid radioactive waste extracted from the reactor internals or the concrete

reactor cavity is placed in a special capsule by means of dedicated devices which are part

of the reactor plant. Such waste is transported in a protective container to the solid radioactive waste storage, dismantled and disposed to the final repository.

preliminary safety analysis report (psar) chapter 1

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