imagination_APR1400

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Beyond your imagination
APR1400
Advanced Power Reactor 1400

CONTENTS
Brighter than your Expectation
Summary
What is APR1400 02
CHAPTER_2
Design 08
CHAPTER_3
Safety 34
Nuclear Power Plants
in Korea
CHAPTER_1
Introduction 04
Introduction of
Nuclear Power
Construction of
Kori #1(‘71-‘78)
PWR(W/H) 1
Promotion of
Localization
Establishment of
Localization Plan (‘84)
PWR(W/H) 5
PWR(Framatome) 2
PHWR(AECL) 1
Technology
Self-reliance
OPR 1000
Development (‘95)
PWR(OPR1000) 4
PHWR(AECL) 3
Development of
Advanced Reactor
APR 1400
Development (‘02)
PWR(OPR1000) 8*
PWR(APR1400) 4**
Building and Structure 10
Reactor Containment Building 11
Auxiliary Building 12
Compound Building 13
Turbine Building 13
Primary System 14
Reactor Coolant System 14
Reactor Vessel and Internals 16
Reactor Core 17
Fuel Assembly 18
Control Element Assembly 20
Integrated Head Assembly 20
Pressurizer 21
Steam Generator 22
Reactor Coolant Pump 24
Secondary System 26
MMIS and Electrical System 28
I & C System 28
Main Control Room 30
Electrical System 33
Safety Goal and Design Philosophy 36
Safety System and Feature 36
Seismic Design 39
Reactor Containment Building Design
against Severe Accident 39
Severe Accident Mitigation Design 40
Reactor Containment Building Design
against Aircraft Impact 42
NRC Design Certification Process of APR1400 43

Clearer than your Anticipation
CHAPTER_4
Proven & Evolutionary
Technology 44
CHAPTER_5
Radioactive Waste
Management 50
CHAPTER_7
Plant Operaion &
Maintenance 62
Safety Injection Performance Test for
Direct Vessel Injection 46
Performance Test for Fluidic Device in
Safety Injection Tank 46
Performance Test for IRWST Sparger 47
Advanced Thermal Hydraulic Test Loop
for Accident Simulation 48
Vitrification of Low and Intermediate
Level Waste 52
High Density Spent Fuel Storage Rack 53
CHAPTER_6
Construction 54
Improvement for In-Service Inspection 64
Enhanced Refueling Work 64
Design Feature for Reducing Unplanned Trip 65
Excellent Operation Performance of Korean
Nuclear Power Plant 66
Korean Nuclear Group Synergy 68
Reactor Containment Building Work 56
Modularization 57
Design Verification by 3D CAD System 58
APR1400 Construction Schedule 61
Conclusion 70

Summary
What is APR1400 ?
It stands for Advanced Power Reactor with a 1,400 MW electrical power and pressurized water reactor developed
in Korea. Based on the self-reliant technologies and experiences from the design, construction, operation and
maintenance of OPR1000, the APR1400 adopts advanced design features to dramatically enhance plant safety,
economical efficiency, and convenience of operation and maintenance.
Advanced Design
4 Train Direct Vessel Injection Safety System and Fluidic Device in Safety Injection Tank
In-containment Refueling Water Storage Tank
Digital I&C and Operator-Friendly Man-Machine Interface
Enhanced Safety
Adoption of Proven and Evolutionary Technologies
Reduced Core Damage & Containment Failure Frequency
Reinforced Seismic Design Basis
Improved Severe Accident Mitigation System
Improved Cost Effectiveness
Extended Plant Design Lifetime
Reduced Operation & Construction Cost
Minimum Site Boundary through Plant General Arrangement Optimization
State-of-the-art Construction Technologies
Convenient Operation & Maintenance
Extended Operator Response Time
Reduced Occupational Exposure
Convenient Facilities for Improved Maintenance & Inspection
This document provides a brief description of the APR1400
major design features including general plant design,
characteristics of major systems, and performance of nuclear
power plants in Korea.

APR1400 Design Features
General Plant Data
Reactor Core
Reactor Coolant System
Turbine
Electrical Power Output Gross/Net
1,455 / 1,400 MWe
Active Core Length
150 in
Number of Coolant Loops 2
Number
1 High Pressure and 3 Low pressure
Thermal Power
3,983 MWth
Equivalent Core Diameter
143.6 in
Operating Pressure
2,250 psia
Type
6 Flow, Tandem Compound
Design Lifetime
60 years
Average Linear Heat Rate
5.6 kW/ft
Coolant Inlet Temperature
555
Speed
1,800 rpm
Seismic Design Basis SSE 0.3g
Number of Fuel Assemblies 241, 16 16
Coolant Outlet Temperature
615
Number of Control Element Assemblies 93
Fuel Cycle Length
18 months

Advanced Power Reactor 1400 (APR1400) and Advanced Power Reactor 1000 (APR1000), we are conducting overseas
nuclear power projects as a source for future profit.
Since nuclear power plants were first introduced into Korea, KEPCO has played a leading role in the nuclear power industry
as well with its subsidiaries such as KHNP for nuclear power generation, KOPEC for engineering, KNF for nuclear fuel,
KPS for maintenance, and KEPRI for R&D.
Based on abundant experience accumulated in overseas projects and credibility in global markets, KEPCO leads the Korean
nuclear industry for joint development of overseas nuclear power plant markets.
Optimum Energy Mix
Nuclear Coal Gas Oil Hydro
Installed Capacity
(As of the end of 2009)
In Operation : 20 Units (17,716 MWe)
Nuclear
17,716MW
(24.1%)
5,515MW
(7.5%)
6,541MW
(8.9%)
18,357MW
(25.0%)
24,205MW
(32.9%)
*Others : 1,136 MW(1.5%)
Total : 73,470MW
Nuclear Power Plants in Korea
5,544GWh
(1.3%)
25,018GWh
(5.8%)
60,349GWh
(13.9%)
Electrictity
Nuclear
147,771GWh
(34.1%)
193,218GWh
(44.6%)
*Others : 1,410GWh(0.3%)
Total : 433,310GWh
Under Construction : 8 Units (9,600 MWe)
Geographic Overview
In Operation
Under Construction
A Main Player of the Nuclear Renaissance
As the single largest state-owned company in Korea, KEPCO has lighted the way to modernization and prosperity. For the
last 120 years, KEPCO has been in charge of generation, transmission and distribution, and construction and management
of electric facilities. KEPCO's mission is to supply affordable and quality electricity in a stable manner.
Nuclear power is the most economical source of energy for green growth. And the world is seeing the advent of a nuclear
renaissance where some 300 nuclear reactors are expected to be built throughout the world by 2030. KEPCO is the world' s
6th nuclear powerhouse with an installed capacity of 17,716MW as of the end of 2009. KEPCO commercially operates 20
In operation
20 units
(17,716MW)
Under
construction
8 units
(9,600MW)
Radioactive Waste
Disposal Facility
(Under construction)
Ulchin
8units
Wolseong
6units
Kori
8units
units as of 2009, with 8 more units currently under construction and additional 10 units planned to be built by 2030.
We retain world-class operating capabilities with unplanned shutdown of 0.35 times/unit-year and a capacity factor of
93.4%.
Under Planning
10 units
Yonggwang
6units
We also boast first rate competitiveness in building nuclear power plants. Based on our accumulated know-how and
technological expertise in constructing and operating nuclear power plants, such as the export-type nuclear reactors of the

KEPCO and APR 1400 chosen by the UAE
On December 27, 2009, KEPCO was awarded the UAE Civilian Nuclear Power Program (CNPP), which is the
single largest nuclear power plant project. The UAE CNPP is hosted by Emirates Nuclear Energy Corporation
(ENEC) and an international competitive bid was held in February 2009 for the construction of four nuclear reactor
units with the first unit to be delivered by May 1, 2017.
KEPCO will supply the full scope of works and services including engineering, procurement, construction, nuclear
fuel and operations and maintenance support for the four units of APR 1400, the latest reactor model of Korean
Standard Nuclear Power Plant, for the UAE s peaceful nuclear power program.
KEPCO is the Prime Contractor of this project, with the assistance of other Korean members, including Hyundai,
Samsung, Doosan Heavy Industries and KEPCO subsidiaries, also with non-Korean companies such as
Westinghouse of the US and Toshiba of Japan.
The plants will be built in the region of Ruwais, about 270km west of Abu Dhabi, the capital city of the UAE.
The APR1400 model is currently under construction in Korea for Shin-Kori Units 3&4 and Shin-Ulchin Units 1&2,
each of these units to be completed consecutively from 2013 to 2016. The first unit to be built in the UAE, therefore,
will be the fifth reactor of its kind in the world.
Why APR1400 of KEPCO?
Enhanced safety features and proven
technology adoption
Designed to meet the requirements of the
EPRI URD, US NRC, and IAEA
World-class operational performance
On-time deliverability enabled by Kepco's
long years of experience
Cost effectiveness proven by the
construction and operation cost per unit
Owner : UAE ENEC(Emirates Nuclear Energy
Corporation)
Capacity : 5,600MW (1,400 X 4)
First Unit complete : 2017. 5. 1
Contract Type : Turnkey

CHAPTER 1

History of Korean Nuclear Technology Development
Flow Diagram of APR1400 Development

06 Beyond your imagination_ APR1400 CHAPTER_1 INTRODUCTION
Introduction
Since the late 1980s, nuclear energy has proven as one of the cleanest energy sources. According to global
environmental preservation movements such as United Nations Framework Convention on Climate
Change (UNFCCC), many countries have promoted the development of evolutionary Advanced Light
Water Reactors (ALWRs) as a vital step for preserving the global environment and ensuring public safety
in order to not be threatened again by accidents like the TMI-2 and Chernobyl mishaps. Korean
government launched the development of a new concept reactor in 1992, which effectively respond to
global environmental preservation programs and dramatically enhance the plant safety and economical
efficiency.
The APR1400 design evolved from four decades of nuclear power plant constructions and development in
Korea. This could be divided into 4 eras.
The 1970s was a period of introducing the first nuclear power plant in Korea. This period was started with
the construction of Kori # 1 with a turn-key contract. One unit was built during this period.
The 1980s was a period of starting national companies to participate in the construction with an
individualized separate contract for each company. During this period, eight units were constructed with
partially acquired foreign nuclear technology.
Introduction of
Nuclear Power
Promotion of
Localization
Technology
Self-reliance
Development of
Advanced Reactor
Construction of
Kori #1(‘71-‘78)
Establishment of
Localization Plan (‘84)
OPR 1000
Development (‘95)
APR 1400
Development (‘02)
PWR(W/H) 1
PWR(W/H) 5
PWR(Framatome) 2
PHWR(AECL) 1
PWR(OPR1000) 4
PHWR(AECL) 3
PWR(OPR1000) 8*
PWR(APR1400) 4**
* 4 units in operation, 4 units under construction
** 4 units under construction
History of Korean Nuclear Technology Development

Beyond your imagination_ APR1400
07
The 1990s was a period of technological self-reliance. Self-reliance of nuclear technology was the main
theme of this period by designing the OPR1000 without relying on foreign companies. Seven units were
constructed during this period.
The 2000s has been the period of developing an advanced nuclear reactor. The repeated constructions and
the operation experiences of OPR1000 brought forth our internationally competitive construction
technology and outstanding operation and maintenance capabilities. By adopting advanced design features
based on the self-reliant technologies, Korea developed the APR1400 to meet Korean Utility Requirement
Document (KURD) that reflects the ALWR design requirements developed in Electric Power Research
Institute (EPRI), European Utilities Organization and others.
The APR1400 obtained its Design Certification (DC) from the Korean nuclear regulatory body in May
2002. Construction of Shin-Kori # 3 & 4 was commenced as the first APR1400 plants in 2007 and they are
planned to start commercial operation in 2013 and 2014, respectively. Two units of Shin-Ulchin # 1 & 2
are under construction as the second construction project of the APR1400.
Flow Diagram of APR1400 Development

CHAPTER 2

Building and Structure
Primary System
Secondary System
MMIS and Electrical System

10 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
Building and Structure
AB
RCB
RCB
CB
TGB
AB
TGB
APR1400 Two-unit Plant Layout of Major Structures
The general arrangement of the APR1400 has been developed based on the twin-unit and slide-along
concepts. The layout of the APR1400 can be divided into Nuclear Island (NI) and Turbine Island (TI). The
NI consists of Reactor Containment Building (RCB), Auxiliary Building (AB) and Compound Building
(CB). The CB includes access control area, radwaste treatment area, and hot machine shop as common
facilities for both units. The TI is composed of Turbine Building (TB) and Switchgear Building (SB).
The RCB is located at the center of the NI and is placed on a common basemat with the AB, which houses
Emergency Diesel Generators (EDGs) and Fuel Handling Area (FHA). The layout of NI improves the
structural safety margin against external events such as a seismic event.
The layout of AB, particularly for the physically separated arrangement of safety equipment, is designed to
enhance plant safety. As examples, four train Safety Injection Systems (SISs) and two set of EDGs are
arranged that each one is placed in the physically separated division of AB. This configuration prevents the
propagation of system damage by internal and external events such as flood, fire, security and sabotage.
Other internal structures are also arranged to improve maintainability, accessibility, and convenience of
equipment replacement.

Beyond your imagination_ APR1400 11
1. Reactor Containment Building
The Reactor Containment Building (RCB) is a pre-stressed concrete structure in the shape of a cylinder
with a hemispherical dome as seismic category I and is founded on a common basemat with the Auxiliary
Building (AB). The inner surface of RCB is steel-lined for leak-tightness and a protective layer of concrete
covers the portion of the liner over the foundation slab.
The In-containment Refueling Water Storage Tank (IRWST) is located in the RCB, in an annular-shape
configuration between the secondary shield wall and the containment wall. Therefore, the Safety Injection
Pumps (SIPs) always take water from the IRWST without switching its suction from the IRWST to the
containment sump for the long term cooling following a Loss of Coolant Accident (LOCA).
As measures to mitigate the consequences of severe accidents, the reactor vessel cavity is designed for heat
transfer area of corium to be not less than 0.02MW such that it is cooled and solidified on the cavity
floor. Also, the convoluted vent path of the reactor vessel cavity prevents the molten core debris from being
released into the containment atmosphere.
In order to improve the convenience of maintenance, the equipment hatch, the arrangement of the
structures, and the polar bridge crane are designed for a steam generator to be replaced in one piece. Also,
work platforms are installed to enhance the convenience of In-Service Inspection (ISI) for steam generators
and maintenance of Reactor Coolant Pumps (RCPs).
IRWST System Structure
Core Debris Chamber and Convoluted Flow Path
Major Design Characteristics
In-containment Refueling Water Storage Tank for improving the safety function of the Safety
Injection System
Spacious reactor vessel cavity and
convoluted vent path for mitigating the
consequences of severe accidents.
Work platforms for In-Service Inspection
of steam generators and maintenance
of RCPs

12 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
2. Auxiliary Building
The Auxiliary Building (AB) is designed with reinforced concrete structure as seismic category I and
placed on a common basemat with the Reactor Containment Building (RCB). It wraps around the RCB
with a quadrant arrangement.
The AB houses the Main Control Room (MCR), Emergency Diesel Generators (EDGs) room, Fuel
Handling Area (FHA), and the safety related components such as Safety Injection System (SIS). The
systems and internal structures in the AB are arranged to provide a physical separation for minimizing the
hazard from internal and external events such as flood, fire, security problem, and sabotage without
adversely affecting accessibility. The safety equipment is spatially separated to enhance its actuation
reliability. Each train of SIS, which consists of four trains, is located in each separate division. The EDGs
are also spatially separated on opposite sides.
In order to improve the convenience of operation and maintenance, the internal layout of the AB is
designed to provide the sufficient space and the lifting rig for replacing heat exchangers and to replace a
generator of EDG without removing the outer wall. Technical Support Center (TSC) is located adjacent to
the MCR to improve communication between operators and technical crews during abnormal plant
situations. The passages are designed for visitors and plant crews not to interfere with each other in the
FHA, the MCR and turbine operating floor. And the internal arrangement of components is divided into the
radiation area and clean area to reduce the occupational exposure dose.
Common Basemat of Containment Building and Auxiliary Building
Quadrant Arrangement of Auxiliary Building
Major Design Characteristics
Designed with reinforced concrete structure as seismic category I
Quadrant arrangement with wrapping around the RCB on the common basemat to enhance
the structure safety against external events
Physically separated arrangement of safety systems to enhance actuation reliability
Optimized arrangement of components, structures, passages, and rooms to enhance the
convenience of operation and maintenance
Distinctly classified installation of components between the radiation zone and the clean zone
to reduce the occupational exposure dose

Beyond your imagination_ APR1400 13
3. Compound Building
As a common facility for both units, the Compound Building (CB) is designed with reinforced concrete
structure as seismic category II. This building houses an access control area, a radwaste treatment area,
primary and secondary sampling laboratories, and a hot machine shop. This arrangement makes access
from each unit more convenient and reduces the size of the power block by making it more compact.
Major Design Characteristics
Designed with reinforced concrete structure as seismic category II
Integrated arrangement of the common facilities for both units to improve the accessibility and
to reduce the site of the power block
4. Turbine Building
The Turbine Island (TI) consists of the Turbine Building (TB) and Switchgear Building (SB), arranged in
the radial direction to the Nuclear Island (NI). Both buildings are situated on a common basemat and
designed with steel structure and reinforced concrete turbine pedestal as seismic category II. The TB
encloses the components that constitute heat cycle and produce the electricity. The SB houses the electrical
distribution equipment.
The conventional construction method for outdoor underground facilities delays the construction because
of repeated digging and filling for installation of various underground equipment. To reduce the
construction schedule, the underground common tunnel is designed to accommodate all underground
facilities in the base floor of the TB. In addition, for the effective maintenance, the whole demineralizers in
the plant are located at the same level.
Major Design Characteristics
Designed with steel structure and reinforced concrete turbine pedestal as seismic category II
Underground common tunnel for accommodating all underground facilities to reduce the
construction schedule
Installation of the whole demineralizers in the plant at the same floor for effective maintenance

14 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
Primary System
1. Reactor Coolant System
Reactor Coolant System
The APR1400 is a two-loop Pressurized Water Reactor (PWR) developed according to the evolutionary
approach that relies on the proven design concepts and components of the OPR1000. The major
component dimensions of Reactor Coolant System (RCS) are increased to deal with the upgraded core
power of 3,983 MWth under the same RCS geometrical configuration as that of OPR1000.
RCS of APR1400 consists of two hot legs, four cold legs, two Steam Generators (SGs), four Reactor
Coolant Pumps (RCPs), and one Pressurizer connected to a hot leg. The major components are designed to
have a lifetime of 60 years and the seismic design basis of 0.3 g SSE is applied to strengthen the resistance
to earthquake.
Advanced design features and design improvements are incorporated in the APR1400 design to enhance
the plant safety and the economical efficiency as well as to extend the plant lifetime and to protect the
utilities’ asset.
These advanced design features are derived from Korean Next Generation Reactor (KNGR) development
program launched in 1992. These features also reflect the operation and maintenance experiences gained
from the conventional nuclear power plants.

Beyond your imagination_ APR1400 15
The core outlet temperature is lowered to increase the core thermal margin. This contributes to the
decreased unplanned reactor trips during normal operation and to the enhancement of the operation
flexibility. In addition, the reduction of the core outlet temperature relieves the degradation of the steam
generator tube due to the stress corrosion by adopting Inconel 690 as the steam generator tube material,
which is known to be more resistant to the stress corrosion cracking than Inconel 600 of the conventional
plants.
The pressurizer volume is designed to be larger than those of the conventional plants. This design makes
the APR1400 accommodate transients without Power Operated Relief Valves (PORVs) and minimizes
unplanned reactor trips during the transients.
The Pilot Operated Safety Relief Valves (POSRVs) are used to perform simultaneously the functions of
Pressurizer Safety Valves (PSVs) and Safety Depressurization Valves (SDVs). This change ensures
reliable valve operation without chattering and leakage for any type of discharge flow condition and allows
remote manual operation for valves under post-accident conditions. The POSRVs perform the function of
pressure relief for RCS overpressure transients and RCS depressurization for the Feed and Bleed
operation. Moreover, during Severe Accidents (SAs), these valves relieve the RCS pressure rapidly to
prevent the high pressure ejection of molten core, which induces direct containment heating.
Major Design Characteristics
Upgraded power: 3,983MWth (1,400MWe)
Extended plant design life time: 60 years
Reinforced seismic design basis: 0.3g SSE
Delayed degradation of the steam
generator tube due to stress corrosion by
reducing core outlet temperature and
adopting Inconel 690 as a steam generator
tube material
Increased pressurizer volume to improve
the capability of coping with transients
Enhanced reliability of RCS pressure
control, the leak-tightness, and the
capability of mitigating Severe Accidents
by adopting POSRVs

16 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
2. Reactor Vessel and Internals
The reactor vessel consists of a vertically mounted
cylindrical vessel welded with a hemispherical lower
head and a removable hemispherical upper closure
head. The internal surfaces that are in contact with
the RCS coolant are clad with austenitic stainless
steel to prevent corrosion. The reactor vessel is
manufactured with three shell sections, a vessel
flange, and a hemispherical bottom head. The three
shell sections, bottom head forging and vessel flange
forging are welded together, along with four inlet
nozzle forgings, two outlet nozzle forgings, four
Direct Vessel Injection (DVI) nozzle forgings, and
sixty-one In-Core Instrument (ICI) nozzles. The
upper closure head is fabricated separately and is
bolted to the reactor vessel. The dome and flange are
welded together to form the upper closure head, on
which the Control Element Drive Mechanism
(CEDM) nozzles are welded.
The RTNDT of the reactor vessel is lowered from
10 of the conventional reactor vessel to -10 with
using low carbon steel, which has lower contents of
Cu, Ni, P, S than those of the conventional reactor.
This material improvement extends the lifetime of
reactor vessel to 60 years. In addition, the Cobalt
(Co) content in the reactor vessel material is lowered
to reduce the occupational exposure dose.
The reactor vessel internals of the conventional plant
are fabricated into three parts of upper guide
structure, core support barrel, and lower support
structure. This compares with the reactor vessel
internals of the APR1400, which are manufactured
into two parts by integrating the core support barrel
and lower support structure. This improvement
contributes to the shortening of the construction
schedule.
Reactor Vessel
Major Design Characteristics
Extended lifetime of the vessel beyond 60 years by reducing the RTNDT to -10 with using
low carbon steel as the base metal of the vessel.
Reduced occupational exposure dose by decreasing the content of Co in the base metal of
the vessel
Shortened construction schedule by the integrated manufacture of core support barrel and
lower support structure

Beyond your imagination_ APR1400 17
3. Reactor Core
The reactor core generates heat with a controlled nuclear reaction and transfers the heat generated to the
reactor coolant. It consists of 241 fuel assemblies, 93 Control Element Assemblies (CEAs), and 61 In-Core
Instrument (ICI) assemblies. The core is designed for the refueling cycle to be greater than 18 months with
a maximum discharge rod burn-up of 60,000 MWD/MTU and for the thermal margin to be increased by
more than 10 %. This core design leads to the improvement of the economical efficiency and safety by
increasing the plant availability factor with a longer refueling cycle and the reduction of unplanned reactor
trips.
A B C D E F G H J K L M N P R S T
180
90
N Full Core Box Nmber
1 2 3 4 5 6 7
XXXX Fuel Batch ID
C0 B0 C0 B0 C0 B0 C0
8 9 10 11 12 13 14 15 16 17 18
C0 B0 C1 A0 B2 C2 B2 A0 C1 B0 C0
19 20 21 22 23 24 25 26 27 28 29 30 31
C0 C3 B1 A0 C2 A0 B2 A0 C2 A0 B1 C3 C0
32 33 34 35 36 37 38 39 40 41 42 43 44 45 46
C0 C3 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 C3 C0
47 48 49 50 51 52 53 54 55 56 57 58 59 60 61
B0 B1 A0 B2 A0 B2 A0 B2 A0 C1 A0 C2 A0 B1 B0
62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78
C0 C1 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 C1 C0
79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95
B0 A0 C2 A0 C1 A0 C2 A0 C2 A0 C2 A0 C1 A0 C2 A0 B0
96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112
C0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 C0
113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 128 129
B0 C2 B2 A0 B1 A0 C2 A0 A0 A0 C2 A0 B1 A0 B2 C2 B0
130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 145 146
C0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 C0
147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163
B0A0 C2 A0 C1 A0 C2 A0 C2 A0 C2 A0 C1 A0 C2 A0 B0
164 165 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180
C0C1 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 C1 C0
181 182 183 184 185 186 187 188 189 190 191 192 193 194 195
B0 B1 A0 C2 A0 C1 A0 B1 A0 C1 A0 C2 A0 B1 B0
196 197 198 199 200 201 202 203 204 205 206 207 208 209 210
C0 C3 B2 A0 B2 A0 B2 A0 B2 A0 B2 A0 B2 C3 C0
211 212 213 214 215 216 217 218 219 220 221 222 223
C0 C3 B1 A0 C2 A0 B2 A0 C2 A0 B1 C3 C0
224 225 226 227 228 229 230 231 232 233 234
C0 B0 C1 A0 B2 C2 B2 A0 C1 B0 C0
235 236 237 238 239 240 241
C0 B0 C0 B0 C0 B0 C0
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
0
Core Loading Pattern
Major Design Characteristics
Improved economical efficiency with a longer refueling cycle greater than 18 months
Increased thermal margin by more than 10 % to reduce unplanned reactor trips

18 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
4. Fuel Assembly
The fuel assembly consists of fuel rods, spacer grids, guide tubes, and the upper and lower end fittings. 236
locations of each fuel assembly are occupied by the fuel rods containing UO2 pellets or the burnable
absorber rods containing Gd2O3-UO2 in a 1616 array. The remaining locations are the 4 CEA guide tubes
and 1 in-core instrumentation guide tube for monitoring the neutron flux shape in the core. Each guide tube
is attached to fuel assembly spacer grids and to upper and lower end fittings to provide a structural frame to
position the fuel rods.
The advanced fuel assembly, named PLUS7, is developed by Korea Nuclear Fuel (KNF) in accordance
with APR1400 development. This fuel assembly is enhanced in thermal hydraulic and nuclear performance
and the structural integrity compared with a conventional fuel assembly as follows. The mixing vanes with
high thermal performance, which induce a relatively small pressure loss, are adopted in all mid-grids to
increase the thermal margin by more than 10 % that has been confirmed in Critical Heat Flux (CHF) test.
The batch average burn-up is increased to 55,000 MWD/MTU through optimizing the fuel assembly and
fuel rod dimensions and adopting an advanced Zirlo alloy as a fuel clad. The neutron economy is improved
with the introduction of the axial blankets at both ends of the pellet region and the optimization of the fuel
Fuel Assembly (PLUS7)

Beyond your imagination_ APR1400 19
rod diameter. The mid-grid buckling strength has been increased using straight grid straps and optimizing
grid height. These design improvements increase the seismic resistance for the fuel assembly to maintain its
integrity even under severe seismic-related accidents.
The conformal type contact geometry between the mid-grid spring and fuel rod increases the in-between
contact area to improve the resistance capacity for fretting wear. The Debris-Filter Bottom Nozzle (DFBN)
is adopted to trap most debris before they enter the fuel assembly. This increases the debris filtering
efficiency to reduce fretting wear-induced fuel failures.
Major Design Characteristics
Increased thermal margin by adopting high thermal performance mixing vanes in all mid-grids
Enhanced batch average burn-up by optimizing the fuel assembly and fuel rod dimensions and
adopting an advanced Zirlo alloy as a fuel clad
Improved neutron economy by introducing the axial blankets at both ends of the pellet region
and optimizing the fuel rod diameter
Increased mid-grid buckling strength with the use of straight grid straps and the optimization of
the grid height to augment the seismic resistance
Enlarged contact area between the mid-grid spring and fuel rod with conformal type contact
geometry to improve the resistance capacity for fretting wear
Enhanced debris filtering efficiency by using Debris-Filter Bottom Nozzle to reduce fretting
wear-induced fuel failure

20 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
5. Control Element Assembly
The Control Element Assembly (CEA) is composed of 12 fingers full
strength, 4 fingers full strength, and 4 fingers part strength CEAs.
Neutron absorbing material is contained within a cylindrical sealed metal
tube. The absorber material used for full strength control rod is boron
carbide (B4C) pellets. Inconel 625 is used as the absorber material for the
part-strength control rods.
Control Element Drive Mechanism
6. Integrated Head Assembly
The reactor vessel upper closure head area of the conventional plant
consists of Control Element Drive Mechanism (CEDM) cooling system,
cooling shroud assembly, heat junction thermocouples, missile shielding
structure, and head lift rig. These components are usually disassembled,
separately stored, and reassembled during every refueling outage. The
Integrated Head Assembly (IHA) is applied to simplify the structure
configuration on the reactor vessel upper closure head region and to
improve the convenience of maintenance. All components on the head
region are handled as single unit to reduce the occupational exposure
dose of the maintenance personnel, the space required for equipment
storage, and to shorten the overhaul duration.
Integrated Head Assembly
Benefits of IHA Adoption
Enhanced maintenance convenience to reduce occupational exposure dose and component
storage area
Reduced overhaul duration to improve the plant economy

Beyond your imagination_ APR1400 21
7. Pressurizer
The pressurizer is a vertically mounted cylindrical pressure vessel. Replaceable direct immersion electric
heaters are vertically mounted in the bottom head. The pressurizer is furnished with nozzles for the spray,
surge, Pilot Operated Safety Relief Valves (POSRVs), and pressure and level instrumentation. A manway is
provided in the top head for access for inspection of the pressurizer internals. The pressurizer surge line is
connected to one of the reactor coolant hot legs and the spray lines are connected to two of the cold legs at
the reactor coolant pump discharge. The pressurizer maintains RCS pressure within specified limits in
conjunction with the Chemical and Volume Control System (CVCS) against all normal and upset
conditions without reactor trip.
The pressurizer volume relative to power is increased to enhance the transient response of the RCS to
reduce unplanned reactor trips. Four POSRVs are adopted instead of two Safety Depressurization (SDS)
valves and three Pressurizer Safety Valves (PSVs) of the conventional plant. POSRVs perform the
overpressure protection for the RCS, depressurize the RCS for the initiation of a feed and bleed operation
in the event of Total Loss Of Feed-Water (TLOFW), and allow remote manual operation under postaccident
conditions to prevent high pressure ejection of molten core. The outlet lines from the POSRVs
connect directly to the ring shaped section of the pressurizer discharge header, which transfers the
steam/water discharge to the In-containment Refueling Water Storage Tank (IRWST). This design prevents
the contamination of containment by RCS coolant discharge and ensures reliable valve operation without
chattering and leakage for any type of discharge flow condition.
Major Design Characteristics
Enhanced transient response by increased pressurizer
volume to reduce unplanned reactor trip
Benefits of POSRV adoption
* Minimized the probability of containment contamination due to
the fluid discharge from RCS
* Reliable valve operation without chattering and leakage
* Lowered valve stuck-open susceptibility
Pressurizer

22 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
8. Steam Generator
The Steam Generator (SG) is a vertically inverse
U-tube heat exchanger with an integral
economizer, which operates with the RCS coolant
in the tube side and secondary coolant in the shell
side. The two SGs are designed to transfer the heat
of 3,983 MWth from the RCS to the secondary
system.
The secondary system produces steam to drive the
turbine-generator, which generates 1,400 MWe of
electrical power. Moisture separators and steam
dryers in the shell side limit the moisture content
of the exit steam less than 0.25 w/o during normal
full power operation.
An integral flow restrictor has been provided in
each SG steam nozzle to restrict the discharge
flow in the event of a steam line break.
For the inspection and maintenance of primary
side, a 21 inch primary manway is provided in the
cold leg and hot leg side of the primary head,
respectively. For the secondary shell side, two 21
inch secondary manways allow access to
separator, dryer area, and internal hatch over the
top of the tube bundle, and two 8 inch handholes
are for sludge lancing on the top of tube-sheet.
To improve the integrity of SG tube, the SG tubes
are made of Inconel 690, which has high
resistance to corrosion and the loose parts trapping
feature inside of the feedwater nozzle is adopted to
prevent damage to the internals and tubes of the
SG by foreign materials originating from the
connected secondary system. The upper tube
support bar and plate are designed to prevent the
SG tube from flow induced vibration.
The SG tube plugging margin is increased by 10
%. This increases the SG tube heat transfer area,
which increases steam flow. This design
improvement accommodates increased core
power and compensates for the lower operating
temperature of the secondary side, which is
induced from the reduced core exit temperature.
This design also enhances the resistance of the SG
tube to stress corrosion and extends the lifetime of
the SG.
Steam Generator
The water volume of the SG secondary side is
enlarged to increase the dry-out time up to 20
minutes in the event of Total Loss Of Feed-Water
(TLOFW). This design enhances the capability of
alleviating the transients during normal operation
to reduce the potential for unplanned reactor trips
and enhance plant safety and operational
flexibility.

Beyond your imagination_ APR1400 23
To improve operability, the angle of the primary outlet nozzles is modified to enhance the stability of midloop
operation, and the SG water level control system is designed such that the water level is controlled
automatically over the entire operating range.
Major Design Characteristics
Enhanced SG tube integrity with the following design improvements
* Employed Alloy 690 tubes with excellent corrosive resistance
* Equipped loose parts trapping feature
* Improved design of the upper tube support bar and plate to prevent flow induced vibration
Increased water volume of the SG to increase the dryout time up to 20 minutes against
TLOFW event to enhance the plant safety and to reduce unplanned reactor trips
Modified primary outlet nozzle angle to improve the stability of mid-loop operation
Automatic control of steam generator water level over the entire operating range

24 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
9. Reactor Coolant Pump
The Reactor Coolant Pump (RCP) provides sufficient
forced circulation flow through the RCS to assure adequate
heat removal from the reactor core during power operation.
In addition, the RCP and motor assembly in conjunction
with the flywheel provide sufficient coastdown flow
following loss of power to the pumps to assure adequate
core cooling.
The RCP is a vertical, single stage bottom suction,
horizontal discharge, motor-driven centrifugal pump. The
shaft seal assembly consists of two face-type mechanical
seals, which reduce the leakage pressure from RCS pressure
to the volume control tank pressure. A third face-type lowpressure
vapor seal at the top is designed to withstand
system operating pressure when the RCP is not operated.
The head and the capacity are increased to accommodate
increased core power compared with those of the
conventional plant.
Reactor Coolant Pump

APR1400
Advanced Power Reactor 1400
Beyond your imagination_ APR1400 25

26 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
Secondary System
The secondary system consists of the main steam,
turbine generator, condensate, feedwater,
extraction steam, and auxiliary systems. The heat
balance of the secondary system is determined
through optimization studies considering system
operability, reliability, and economy. The
secondary system is designed to be capable of
operating at 3 ~ 6 % house load for a period of at
least 4 hours without any detrimental effects on
the systems and increasing the plant condition
from a cold condition to full power within 200
minutes, excluding rotor preheating.
Turbine Generator
52” Last Stage Bucket Design
The Main Steam Supply System (MSSS)
transports the steam from the steam generators to
the power conversion system and removes the
heat from the RCS. The steam flow is directed
from the Steam Generators (SGs) to the High
Pressure (HP) turbine, of which the inlet steam
pressure is maintained at 962 psia during full
power. The turbine generator consists of a doubleflow
HP turbine and three double-flow Low
Pressure (LP) turbines driving a direct-coupled
generator. The LP turbine rotors are of monoblock
type. The material used for the LP rotors is
Ni-Cr-Mo-V alloy steel and is treated to obtain
enough toughness. The 52 inch last stage buckets
of the LP turbine are designed to have low stress
and increased stiffness. The generator system
consists of the generator itself and auxiliary
systems such as a stator cooling water system, gas
control system and seal oil system. The stator of
the generator takes a highly reliable F-class
Micapal II insulation system and a highly reliable
brazing technology. The rotor of the generator also
adopts the highly reliable insulation system and
radial flow cooling method.
The condensate and feedwater systems transfer
condensate from the main condenser hotwells to
the SGs while the feedwater heaters raise the
condensate temperature by using the extraction
steam, and the deaerator removes the entrained
oxygen and non-condensable gases. The
feedwater heaters are installed in 6 stages and
arranged horizontally for easy maintenance and
reliability. The configuration of the Main Feed
Water Pump (MFWP) is designed to be 355 %
to allow more reliable operation. Even if one
MFWP is tripped during full power condition with
three MFWPs operating, the other two MFWPs
could recover the total feedwater flow to the
nominal value of the full power condition and the
plant is restored to the full power condition. This
design reduces unnecessary power cutbacks and
unplanned turbine trips.
The Auxiliary Feed Water System (AFWS)
supplies feedwater to the SGs for events resulting
in loss of normal feedwater and requiring heat
removal through the SGs. The AFWS is actuated
by an Auxiliary Feedwater Actuation Signal
(AFAS) from the Engineered Safety Features

Beyond your imagination_ APR1400 27
Actuation System (ESFAS) or the Diverse
Protection System (DPS). The ESF-Component
Control System (ESF-CCS) includes logic to close
the flow control valves when the SG water level
has risen above a high level setpoint, and to re-open
this valve when the SG water level drops below a
low level setpoint. Different from the conventional
plant, the AFW storage tank is installed in the
auxiliary building separated from the condensate
tank to enhance the system reliability in transients.
Generator
Major Design Characteristics
Capable of operating at 3 ~ 6 % house load for more than 4 hours and increasing the plant
condition from a cold condition to full power within 200 minutes, excluding rotor preheating
52 inch last stage buckets of LP turbine having low stress and increased stiffness
Adopted horizontally arranged feed-water heaters for easy maintenance and reliability
355 % MFWP configuration to reduce unnecessary power cutbacks and unplanned turbine trips
Installed AFW storage tank in the auxiliary building to enhance the system reliability in
transients

28 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
MMIS and Electrical System
1. I & C System
In order to come up with high performance and maintainability, I&C systems including control and
monitoring systems are based on proven, diverse, and commercial off-the-shelf hardware, network, and
software platforms such as Distributed Control System (DCS) and Programmable Logic Controller (PLC)
with operating experience of more than 3,000 years.
The I&C architecture consists of two platforms. One is a safety grade PLC platform, which is comprised of
Core Protection Calculator (CPC), Plant Protection System (PPS), and ESF-Component Control System
(ESF-CCS). These safety systems are designed to meet the licensing requirements for independence,
defense-in-depth, diversity analysis, failure mode analysis, and environmental qualification. The fail-safe
concept is implemented such that the system is allowed to operate safely.
The other is a non-safety grade DCS platform, which consists of Power Control System (PCS), NSSS
Process Control System (NPCS), Process-Component Control System (P-CCS), Diverse Protection System
(DPS), and Information Processing System (IPS). The DCS with multi-loop controllers is adopted for these
non-safety I&C systems. The number of I&C platform is minimized to increase the cost-effectiveness and
to decrease operator°Øs maintenance burden.
Defense against the Common Mode Failure (CMF) of digital plant protection systems is one of the key
requirements in designing digital I&C systems. The DPS is designed to be diverse from the PPS against the
CMF of the digital plant protection system. Diverse manual ESF actuation is also designed to keep the
plant safety against severe situations due to a simultaneous digital system failure of the PPS and the DPS.
The open architecture concept is applied to the configuration of the I&C system for high reliability and
maintainability. In addition, the stringent software & hardware qualification process is established and
followed for the life cycle.

Beyond your imagination_ APR1400 29
I&C Architecture
Major Design Characteristics
Applied proven technology by adopting commercial off-the-shelf hardware, network, and
software platforms with operating experience of more than 3,000 years
Applied a fault tolerant design by adopting a fail-safe concept
Implemented the DCS with the multi-loop controllers for simplified I&C facilities and for
enhanced cost effectiveness
Employed diverse product types and manufacturers in the PPS, DPS and manual ESF
actuation system to keep the plant safe against CMF
Adopted an open architecture concept for convenience of system modification and upgrade by
using diverse manufacturers
Performed stringent software and hardware qualification processes for a highly reliable digital
I&C system

30 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
2. Main Control Room
Main Control Room
The computerized Main Control Room (MCR) of the APR1400 adopts compact workstations. These
workstations are integrated with operator support systems with a human centered automation concept. The
MCR provides operating crews with an information-oriented operational environment that enables fast
situational awareness of plant status.
The design goal of the advanced MCR is to enhance the plant safety by extending operator coping time
against accidents and to reduce human errors by improving the operation readiness. The following
advanced technologies and design concepts are being incorporated to achieve the design goal. In addition, a
safety console is provided in the MCR as a backup facility for safe operation against a total failure of
workstations.

Beyond your imagination_ APR1400 31
Compact Workstation
The MCR contains workstations, a Large Display Panel (LDP), and a safety console. The workstations are
identical and reconfigurable so that an operator has a backup workstation to cope with when different types
of workstation failures are encountered. The workstations are placed near one another in a fixed location to
improve communication among the operators.
A unified computer based Man-Machine Interface (MMI) design is applied for all the major plant systems
for operator’s convenience. The compact workstation with a computer-based MMI reduces the interface
management tasks using the two-click access and format chaining. The MMI design adopts the system and
function based displays as well as the diverse information displays for operation. The enhanced display
includes dynamic logic display, P&ID, video data from CCTVs, and design data. In addition, the compact
workstation is designed to provide all operational means, including the computerized operator support
functions such as critical function monitoring, success path monitoring, signal validation, and computerized
procedures. This integrated compact workstation design is expected to reduce workload of the operating
crew.
Large Display Panel
The Large Display Panel (LDP) is large enough to be viewed from anywhere in the MCR. It presents the
plant level indications and alarms, which enable the operating crews to assess the plant situation related to
the critical safety and the power production functions. The LDP displays are developed to have a simplified
and fixed format with the °Ædark board°Ø concept. A °Ædark board°Ø concept means that no alarm
indicates the normal state of plant. This allows operating crews quickly and easily to assess the plant status
at a glance. The LDP is designed to quickly direct the operators°Ø attention to the exact trouble source and
to allow them to diagnose severity of plant incidents. The LDP provides sufficient information for
emergency operations. The LDP also continuously displays the critical function performance and the
success path availability.
Safety Console
The safety console is provided as a backup for safe operation against a total DCS failure. The safety
console indications are designed to provide the qualified information and alarms in the similar format to the
LDP display to enhance familiarity of the display.
Computerized Procedure System
The Computerized Procedure System (CPS) provides an integrated presentation of procedural instructions
and related process information needed to execute applicable procedures by both an operator and the
operating crews. The CPS also monitors the plant condition and supports the recovery actions for
inadvertent errors committed by operators. The CPS is designed to monitor continuously applied steps and
non-sequential steps. The CPS supports the multiple procedure execution as well as the multi-user
execution. The CPS is capable of accessing information displays and alarm system and evaluating the
instruction logic of a procedure.

32 Beyond your imagination_ APR1400 CHAPTER_2 DESIGN
Human Factors Engineering
The extensive Human Factors Engineering (HFE) program is incorporated to reduce the possibility of a
human error in the MCR. In the conceptual and basic design at the R&D stage as well as during the
construction phase of the plants, the MMI design has been analyzed and evaluated in an iteratively expanding
manner with participation of more than 100 licensed operators and human factors specialists to optimize the
design. During the APR1400 development, the evaluation for the MMI design has been performed seven
times with full scope dynamic mockups and an APR1400 specific dynamic mockup. The evaluation verified
that the MMI are suitable for the human factor principles and guidelines. The new MCR design was also
validated to support the normal and emergency operations appropriately.
Configuration of APR1400 HFE Evaluation System
Human Factors Engineering Validation with Dynamic Mockup
Major Design Characteristics
Adopted compact workstations to provide a unified MMI for control and monitoring of systems
Applied automation and operator support functions such as signal validation, safety
monitoring, and a computerized procedure
Applied redundancy for monitoring and control to enhance the robustness against digital
system failures
Provided a safety console as a backup facility for safety operation against the total failure of
workstation
Applied the Human Factors Engineering (HFE) principles and program to address the
complexity of the computer based operator interfaces

Beyond your imagination_ APR1400 33
3. Electrical System
The plant electrical system consists of the generator, the generator circuit breaker, the Main Transformer
(MT), the Unit Auxiliary Transformers (UATs) and the Stand-by Auxiliary Transformers (SATs). The
normal power source for non-safety loads are the off-site power through the main transformer and the onsite
power through the UATs from the generator.
The electric power for the safety-related systems is supplied from the following four alternative ways: the
normal power source of the normal off-site power through the MT or the on-site power through UATs
generated by the in-house generator, the stand-by off-site power connected through the SATs with the grid,
the on-site stand-by power supply from two Emergency Diesel Generators (EDGs), and an Alternative
Alternate Current (AAC) source from a backup diesel generator.
During normal operation, the electric power for the safety-related systems is supplied from the normal
power source of the normal off-site power through the MT or the on-site power through the UATs. If the
normal power source is not available, the safety loads are covered with the off-site power source via the
SATs. Then, if the off-site power source to the safety-related systems is interrupted, the safety loads are
backed up by two independent Class 1E EDG sets. Each of them is located in a separated room of the
auxiliary building and is connected to two 4.16 kV safety buses.
The non-class 1E AAC source adds more redundancy to the electric power supply for safety systems. It is
provided to cope with Station Blackout (SBO) situation which has a high potential of transients to severe
accidents. The AAC source has sufficient capacity to accommodate loads on the safety.
One Line Diagram of Station Power Block

CHAPTER 3

Safety Goal and Design Philosophy
Safety System and Feature
Seismic Design
Reactor Containment Building Design against Severe Accident
Severe Accident Mitigation Design
Reactor Containment Building Design against Aircraft Impact
NRC Design Certification Process of APR1400

36 Beyond your imagination_ APR1400 CHAPTER_3 SAFETY
Safety
1. Safety Goals and Design Philosophy
One of the APR1400 development policies is to dramatically increase the level of safety. The following
safety goals are established to improve the plant safety level by ten times. The APR1400 design has been
made to meet these safety goals with securing an additional margin of safety to protect the owner’s
investment as well as public health.
Major Safety Goals of APR1400
The total Core Damage Frequency (CDF) shall not exceed 10-5/RY for both internal and
external initiating events and 10-6/RY for a single event and an incident occurring in a high
pressure condition.
The containment failure frequency shall be less than 10-6/RY.
The whole body dose at the site boundary shall not exceed 0.01 Sv (1 rem) for 24 hours after
the initiation of core damage with a containment failure.
2. Safety Systems and Features
The safety systems consist of Safety Injection System (SIS), In-containment Refueling Water Storage
System (IRWST), Safety Depressurization and Vent System (SDVS), Containment Spray System (CSS),
and Auxiliary Feed Water System (AFWS).
The main design concept of the SIS is simplification and redundancy to achieve higher reliability and better
performance than the conventional plant. The SIS is comprised of four independent mechanical trains
without a tie line among the injection paths and two electrical divisions. Each train has one active Safety
Injection Pump (SIP) and one passive Safety Injection Tank (SIT) equipped with a Fluidic Device (FD).
For simplicity and independence of SIS, the common header installed in the SIS lines of the conventional
plant is eliminated. This design separates the functions of SIS and Shutdown Cooling System (SCS).
The passive flow regulator, that is, the FD is installed in the SIT. The basic concept of the FD is vortex flow
resistance. When water flows through the stand pipe, which is installed in a rectangular direction with the
exit nozzle, it creates low vortex resistance and a high flow rate. When the water level is below the top of
the stand pipe, inlet flow is switched to the control ports which are installed in a tangential direction with
the exit nozzle, and it makes a high vortex resistance and low flow rate. Thereby, the SIT discharges a large

Beyond your imagination_ APR1400 37
amount of water to fill the reactor vessel lower
plenum rapidly when water level is above the
stand pipe. However, when the water level is
below the stand pipe, the SIT injects a relatively
small amount of water for a long time. The FD
installed in SIT substitutes for the low pressure
SIPs such that the low pressure SIP is
eliminated.
Safety Injection System
The In-containment Refueling Water Storage
Tank (IRWST) is located inside the
containment and the arrangement is made in
such a way that the injected emergency cooling
water returns to the IRWST. This design
removes the operator action of switching the
suction of SIP from the IRWST to the
containment recirculation sump, required in the
conventional plant. This new design lowers the
susceptibility of IRWST to external hazards.
The IRWST provides the following functions:
storing refueling water, a water source for the
SIS, shutdown cooling system, and
containment spray system, a heat sink to
condense steam discharged from the pressurizer
Safety
Injection
Tank
Closure Plate
Partition Plate
Stand Pipe
Vortex Chamber
Middle & Lower Plate
Discharge Tube
Insert Plate
Fluidic Device

38 Beyond your imagination_ APR1400 CHAPTER_3 SAFETY
for rapid depressurization if necessary. This prevents high pressure molten corium ejection and enables
feed and bleed operation. This also allows supplying coolant to the cavity flooding system which mitigates
molten corium concrete interaction in severe accidents.
Through adopting the advanced features of the FD in SIT and the IRWST, the high pressure injection, low
pressure injection, and recirculation modes of the conventional SIS are merged into one operation mode of
safety injection. The SIS is designed for safety water to be injected directly into the reactor vessel so that
the discharge of injected flow through the broken cold leg is eliminated.
The Safety Depressurization and Vent System (SDVS) is a dedicated safety system designed to provide a
safety grade means to depressurize the RCS in the event that the pressurizer spray is unavailable during
plant cooldown to cold shutdown and to rapidly depressurize the RCS to initiate the feed and bleed method
of plant cooldown subsequent to the total loss of feedwater event. The Pilot Operated Safety Relief Valves
(POSRVs) are employed for feed and bleed operation. This system establishes a flow path from the
pressurizer steam space to the IRWST.
The Containment Spray System (CSS) consists of two trains and takes the suction of its pump from the
IRWST to reduce the containment temperature and pressure during accidents occurring in the containment.
The CSS is designed to be interconnected with the Shutdown Cooling System (SCS), which is also
comprised of two trains. The pumps of these systems are designed to have the same type and capacity. This
design makes the CSS have a higher reliability compared with the conventional plant.
The Auxiliary Feed Water System (AFWS) consists of two divisions and four train systems, and supplies
feedwater to the SGs for RCS heat removal in case of loss of main feedwater. In addition, the AFWS refills
the SGs following a LOCA to minimize leakage through pre-existing tube leaks. The reliability of AFWS
has been increased by the use of two 100 % motor-driven pumps, two 100 % turbine-driven pumps, and
two independent safety-related emergency feedwater storage tanks located in the auxiliary building instead
of a condensate storage tank of the conventional plant.
Major Design Characteristics
Improved reliability of SIS through the design of four trains mechanical equipment
Simplified operation of SIS by merging high pressure injection, low pressure injection, and recirculation
modes into one safety injection mode
Lowered susceptibility of IRWST to external hazards by locating IRWST in the containment
Enhanced plant safety by adopting advanced features such as the FD in SIT, the IRWST, and
the Direct Vessel Injection of SIS
Improved reliability of CSS through the interconnection design between CSS and SCS
Increased reliability of AFWS by using two 100 % motor-driven pumps, two 100 % turbinedriven
pumps, and two independent safety-related emergency feedwater storage tanks
located in the auxiliary building

Beyond your imagination_ APR1400 39
3. Seismic Design
The buildings and structures are designed with the application of the Safe Shutdown Earthquake (SSE) of
0.3g as a Design Basis Earthquake (DBE) to increase their ductility against earthquakes. The seismic input
motion enforced in the high frequency range is applied to envelope the design ground response spectrum of
the Reg. Guide 1.60 standard spectrum. In the meantime, the design load of Operating Base Earthquake
(OBE) is eliminated to improve design and verification according to 10 CFR 50 Appendix S.
Since the seismic evaluation is performed with the inclusion of the effects of soil-structure interaction on
soil sites, the APR1400 can be constructed on rock bed sites as well as soil sites.
Seismic Input Motion
Schematic Diagram of Soil-Structure Interaction
4. Reactor Containment Building Design against Severe Accident
In order to maintain the integrity of the Reactor Containment Building (RCB) and to prevent the leakage of
radioactive materials during Severe Accidents (SAs), the RCB is designed to have free volume enough so
that the structural load is maintained below ASME Section III Service Level C 24 hours after SAs. This
also helps keep hydrogen concentration under 13% in case of a 75% oxidation of fuel clad-steam. Another
design feature to prevent leakage is installation of the 1/4 inch steel liner plate on the inboard side of the
RCB. In addition, the RCB is constructed with the pre-stressed concrete having the high compressive
strength of 6,000 psi after 91 days of curing.
The reactor vessel cavity is designed such that molten core materials spread out for its heat transferable area
to be not less than 0.02MW and is cooled to solidify on the cavity floor. Also, the convoluted vent path
of the reactor vessel cavity prevents the molten core debris from releasing to the containment atmosphere.
Major Design Characteristics
Enlarged free volume of the RCB to maintain the structural load below the code requirement
and to keep the hydrogen concentration below the design limit
Lined with 1/4 inch steel plate on the inboard side of the RCB to prevent the leakage of
radioactive materials
Reinforced RCB with high compressive strength of 6,000 psi
Spacious reactor vessel cavity and convoluted vent path for the SAs mitigation

40 Beyond your imagination_ APR1400 CHAPTER_3 SAFETY
5. Severe Accident Mitigation Design
The Severe Accidents (SAs) management system prevents and mitigates SAs, and maintains containment
integrity. It is designed to meet the procedural requirements and criteria of US NRC regulations, including
the post Three Mile Island (TMI) requirements for new plants as reflected in 10 CFR 50.34 (f) and SECY-
93-087. This system includes a large dry pre-stressed concrete containment, Hydrogen Management
System (HMS), large reactor cavity and core debris chamber, Cavity Flooding System (CFS), In-Vessel
corium Retention and External Reactor Vessel Cooling System (IVR-ERVCS), Safety Depressurization
and Vent System (SDVS), Emergency Containment Spray Backup System (ECSBS), and Severe
Accidents Management Procedure (SAMP).
Hydrogen Mitigation System
The Hydrogen Mitigation System (HMS) consists of 26 Passive Auto-catalytic Recombiners (PARs) and
10 glow plug igniters. The capacity of the HMS is designed to accommodate the hydrogen production from
a 100% metal-water reaction of fuel cladding and to limit the average hydrogen concentration in
containment below 10% in accordance with 10 CFR 50.34 (f) for a degraded core accident.
PAR
Igniter
Reactor Cavity Design
The reactor cavity adopts a core debris chamber,
which is designed to have a heat transfer area of
corium more than 0.02MWt. The flow path
of the reactor cavity is designed to be convoluted
to hinder the transfer of core debris to the upper
containment. This design prevents Direct
Containment Heating (DCH) by core debris.
Reactor Cavity

Beyond your imagination_ APR1400 41
Cavity Flooding System
The APR1400 has two strategies for cooling the molten core: Ex-Vessel Cooling (EVC) and In-Vessel
Retention (IVR). The Cavity Flooding System (CFS) supplies the coolant for ex-vessel cooling. It consists
of two trains connected with IRWST and two isolation valves that are installed in each line. When the two
isolation valves are open during the SAs, the cavity cooling water is supplied from the IRWST to the
reactor cavity by the gravity. It cools down the core debris in the reactor cavity, scrubs fission product
releases, and mitigates the molten corium concrete interaction (MCCI).
In-Vessel Corium Retention through External Reactor Vessel Cooling
System
In-Vessel corium Retention through External Reactor Vessel Cooling System (IVR-ERVCS) retains
molten core in the reactor vessel by cooling the external surface of reactor vessel. This system submerges
the reactor vessel bottom head before molten core relocates to the bottom head. Cooling water is supplied
from the IRWST by a Shutdown Cooling Pump (SCP) and a Boric Acid Makeup Pump (BAMP). This
system maintains the reactor vessel integrity and reduces the threat of containment integrity.
Cavity Flooding System and IVR-ERVCS
Emergency Containment Spray Backup System
The Emergency Containment Spray Backup System (ECSBS) provides long-term coolability by supplying
spray water with containment for 48 hours so that the containment temperature and pressure are reduced
during the SAs. This system consists of spray nozzle, piping, and containment penetration. Spray water is
supplied by external pump from temporary external water source. The ECSBS contributes to relieving the
threat of containment integrity.

42 Beyond your imagination_ APR1400 CHAPTER_3 SAFETY
6. Reactor Containment Building Design against Aircraft
Impact
A design-specific assessment on the effects on nuclear power plant facilities of the impact of a large
commercial aircraft and military aircraft is performed in accordance with 10 CFR Part 52. The velocity of a
large commercial aircraft and a military aircraft for the aircraft impact assessment can be considered as 150
m/sec and 215 m/sec respectively.
Both single and double shell structure concept has application to protect the nuclear power plant facilities
from possible aircraft impact. The outer wall is strengthened to ensure safety of the structure in the event of
an aircraft impact.
Two distinct types of structural failure modes such as local failure and global collapse are evaluated for
reactor containment building and spent fuel pool. The local failures are caused by impact of the aircraft
engines and are largely independent from the global behavior characteristics of the impacted structure.
Three main evaluations such as the local analysis, the global analysis, and the vibration analysis are
performed for the design of nuclear power plant facilities against aircraft impact.
External fires caused by aircraft impacts are of relatively short duration and will not have a significant
impact on systems necessary to provide cooling of fuel in the reactor vessel or spent fuel pool. This
assumption is based on an abundance of oxygen available to support combustion of the fuel and a good
firefighter access to the fire. However, fire duration according to Standard Review Plan 9.5.1 is
conservatively applied to guarantee structural safety.
After the September 11th terrorists’ attack, Korea assessed the safety of operating nuclear power plants
against aircraft impact and reconfirmed the safety of structures. The evaluation methods applied to
APR1400 nuclear power plant, and the structural integrity were also confirmed. In case of APR1400
nuclear power plant, both single and double shell structures were applied to protect the nuclear facilities
from aircraft impact, and the safety of each case was assessedwith the evaluation procedure.
Deformed Shape (Global Behavior)

Beyond your imagination_ APR1400 43
Aircraft Impact Locations
7. NRC Design Certification Process of APR1400
Design Certification (DC) is the U.S. NRC (Nuclear Regulatory Commission)’s approval of a specific
standardized plant design on the basis of 10 CFR 52 which was developed to reduce licensing uncertainty
by resolving design issues at an earlier stage. The scope and contents of the application are essentially
equivalent to the level of details found in a Final Safety Analysis Report.
Starting from the Pre-application Review meeting and the application of DC, DC is acquired through the
U.S. NRC's rulemaking process, followed by the staff’s review of the application, which includes the
various discussions on safety issues associated with the proposed nuclear power plant design. During this
process, the U.S. NRC provides all stakeholders with opportunities to take part in public meetings and
rulemaking activities related to design certifications.
KEPCO is aspiring to attain opportunities to share our long years of experience and competence in the safe
and peaceful use of nuclear energy and to make a significant contribution to meeting the increasing nuclear
power demand across the world in the future. To this end, KEPCO is pursuing to acquire design
certification from the NRC with the APR1400, which is an optimal design to meet the NRC’s Severe
Accident and Safety Goal Policy Statements.
KEPCO notified the NRC in March 2009 of its intent to acquire the Design Certification of APR1400. And
then, the first NRC-KEPCO Pre-application meeting on the proposed certification was held on November
18, 2009. In the meeting, KEPCO suggested a tentative plan that it would prepare to file for Design
Certification of APR1400 in 2011. Accordingly, a series of Pre-application review meetings are expected
before the filing to arrange the NRC review schedule. Important design information will also be provided
in advance during the process. Although dependent upon the amount of time the NRC requires reviewing
DC applications, the NRC review process is expected to take approximately three years.
Since the APR1400 has already undergone a rigorous examination by the Korean Institute for Nuclear
Safety (KINS) to acquire the construction permit for the first APR1400 units (SKN 3&4), we are
convinced that it wold acquire the NRC DC in 2013 with less time and resources than the precedents.
Pre-application
Review Meetion
Topical Report
Submission
DC Application
Technical Report
Submission
NRC Staff Review
ACRS
Review Meetiong
DC Rulemaking
ACRS:Advisory Committee on Reactor Safeguard

CHAPTER 4 PROVEN &
EVOLUTIONARY TECHNOLOGY

Safety Injection Performance Test for Direct Vessel Injection
Performance Test for Fluidic Device in Safety Injection Tank
Performance Test for IRWST Sparger
Advanced Thermal Hydraulic Test Loop for Accident Simulation

46 Beyond your imagination_ APR1400 CHAPTER_4 PROVEN & EVOLUTIONARY TECHNOLOGY
Proven & Evolutionary Technology
In order to enhance safety, international competitiveness, operational convenience, and maintainability, the
APR1400 was developed by adopting advanced design features. These advanced design features are based
on the proven nuclear power plant design technology gained through many years of repeated constructions
and extensive operation experiences of the OPR1000. The new design features have been successfully
evaluated to ensure that they enhance the performance and safety of the APR1400. The following relevant
experimental activities have been conducted over several years by Korea Atomic Energy Research Institute
(KAERI).
Safety Injection Performance Test for Direct Vessel Injection
The Safety Injection (SI) nozzles in the APR1400 are located in the upper part of the Reactor Pressure
Vessel (RPV) downcomer. Due to this design feature, in a Loss-Of-Coolant Accident (LOCA), the thermalhydraulic
phenomena in the RPV downcomer differ from such phenomena in the case of a Cold Leg
Injection (CLI), and this difference is believed to govern a Large Break Loss-Of-Coolant Accident
(LBLOCA) reflood phase. In order to evaluate the Emergency Core Coolant (ECC) bypass during the
reflood phase of a postulated LBLOCA and to assess the contribution of the new SI system to safety
enhancement, the performance of the SI system was examined by using the Multi-dimensional
Investigation in Downcomer Annulus Simulation (MIDAS) facility. The MIDAS facility was designed to
be 1/5 length scale of the APR1400 and to use steam and water as test fluids at 190 psia and 572 as
design conditions.
MIDAS Facility
Schematic Diagram of MIDAS Facility
Performance Test for Fluidic Device in Safety Injection Tank
The APR1400 uses a Fluidic Device (FD), installed inside Safety Injection Tank (SIT) as a passive design
feature, to ensure effective use of the SIT water. This design feature enables the APR1400 to achieve the
goals of minimizing the ECC bypass during a blowdown, and of preventing a spillage of excess ECC water
during the refill and reflood phases of a LBLOCA.

Beyond your imagination_ APR1400 47
The FD provides a high discharge flow rate of SI water
when the FD starts to operate, which is required during
the refill phase of a LBLOCA. When the refill phase is
terminated, the discharge flow rate of the SI water drops
sharply but is still large enough to remove any decay
heat during the reflood phase. Because of the strong
vortex motion in the FD, the pressure loss coefficient of
the low flow rate period is almost ten times higher than
that of the high flow rate period. The difference in the
pressure drop helps extend the total duration of the SI
and also enables the low pressure safety injection pump
to be removed from the SI system.
In order to confirm the performance of the FD designs,
full-scale performance tests were carried out in the Valve
Performance Evaluation Rig (VAPER) facility, which was
designed with the same size and operating conditions as
those of the APR1400 SIT. It was verified from these fullscale
tests that the performance of the FD satisfies the
standard design requirements of the APR1400.
VAPER Facility
Performance Test for IRWST Sparger
To cope with transients such as a RCS overpressure, the
Safety Depressurization and Vent System (SDVS),
which enables a feed and bleed operation, and the SIS
are incorporated in the APR1400 to maintain the
integrity of the RCS and the core. Actuation of Pilot
Operated Safety Relief Valves (POSRVs) results in a
transient discharge flow of air, steam or a two-phase
mixture to the IRWST through the spargers.
The discharge of these fluids induces complicated
thermal-hydraulic phenomena such as a water jet, air
clearing, and steam condensation. These phenomena
impose relevant hydrodynamic forces on the IRWST
structure and the components of the SDVS. These
structures shall be designed so as to withstand these
hydrodynamic loads and to maintain their structural
integrity as well as the safety functions of the engineered
safety features systems. Hydrodynamic loads on the
IRWST wall and the components of SDVS are induced
by air clearing and steam jet discharge through a
prototype sparger of the APR1400. The relevant test of
these loads was conducted at the Blowdown and
Condensation (B&C) facility.
IRWST Sparger
Schematic Diagram of B&C Facility

48 Beyond your imagination_ APR1400 CHAPTER_4 PROVEN & EVOLUTIONARY TECHNOLOGY
Advanced Thermal Hydraulic Test Loop for Accident Simulation
The Advanced Thermal hydraulic test Loop for Accident Simulation (ATLAS) is a thermal hydraulic
integral effect test facility for simulations of various transients and accident conditions of the APR1400 and
the OPR1000. It simulates various accident scenarios at actual pressure and temperature conditions of the
APR1400. The integrated safety of the APR1400, which adopts new design features, has been verified
through accident simulation in the ATLAS.
The major accident scenarios are reflood phase of the LBLOCA, small-break LOCA scenarios including
the DVI line breaks, steam generator tube ruptures, main steam line breaks, feedwater line breaks, mid-loop
operation, and other transient conditions.
Schematic Diagram of ATLAS

Beyond your imagination_ APR1400 49
Major Design Characteristics of ATLAS
1/2-height, 1/144-area, and 1/288-volume
scale
Full-pressure simulation of the APR1400
Same geometrical configurations as those
of the APR1400 including 24 reactor
coolant loops, a direct vessel injection,
and an integrated annular downcomer
Maximum 8% of the scaled nominal core
power
ATLAS Facility

CHAPTER 5
RADIOACTIVE WASTE
MANAGEMENT

Vitrification of Low and Intermediate Level Waste
High Density Spent Fuel Storage Rack

52 Beyond your imagination_ APR1400 CHAPTER_5 RADIOACTIVE WASTE MANAGEMENT
Radioactive Waste Management
1. Vitrification of Low and Intermediate Level Waste
Development of Low and Intermediate Level Wastes (LILW) vitrification technology started in the early
1990s separately from the APR1400 development. The base technologies for vitrification were developed
from a feasibility study and an international joint program from 1994 to 1998. The pilot vitrification facility
was constructed by KHNP and has operated since October 1999. A commercial vitrification facility was
constructed at the Ulchin NPP site. The vitrification technology can reduce the waste volume to less than
1/20 of the initial bulk volume of LILW.
Induction Cold Crucible Melter
The Induction Cold Crucible Melter (ICCM) vitrifies the combustible LILW in the following way. Glass
frits are loaded into the ICCM and then melted using Joule energy induced by an electromagnetic field.
After the glass is completely melted at about 2,012, shredded combustibles are fed onto the hot glass
melt. The waste is decomposed on hot glass melt and radionuclides in the waste are incorporated into the
stable glass matrix through chemical bonds with glass formers.
The ICCM can vitrify almost any kind of combustible LILW including protective clothes, gloves,
polyethylene, and spent ion exchange resin generated from the water purification system. Dried borate
concentrate, which is non-combustible, can be also vitrified since it contains a large amount of glass former
elements. The radionuclides, which are bonded concretely with the glass matrix, hardly come out to the
environment under any condition.
Plasma Torch Melter
The Plasma Torch Melter (PTM) can handle non-combustible LILW such as concrete, soil, metal scraps,
spent filters, and sludge generated from the liquid waste treatment system. The non-combustible LILW is
transformed into a more environment-friendly waste form, monolith, through high temperature melting by
the PTM. The radionuclides and hazardous heavy metals are incorporated into the waste form which is
produced from inorganic material melted in the PTM.
Off-Gas Treatment Process
The off-gas generated from the ICCM or PTM contains a very small amount of dioxin, volatile radioactive
nuclides, and acid gases such as sulfer oxides, hydrogen chloride, and nitrogen oxides. It is released into the
environment after the hazardous materials are removed and/or decomposed with the Off-Gas Treatment
Process (OGTS). The dust, which contains most of the radioactive elements volatilized from the melter, is
captured in a High Temperature Ceramic Filter (HTCF) and then recycled to the ICCM. Dioxin is removed
and/or decomposed in the HTCF, a Post Combustion Chamber (PCC), and an activated charcoal bed.

Beyond your imagination_ APR1400 53
Sulfer oxides and hydrogen chloride are removed by absorption with a scrubbing solution and nitrogen
oxides are converted into harmless nitrogen at a Selective Catalytic Reduction (SCR) bed.
Cold Crucible Melter
Plasma Torch Melter
Glass Frit Feeder
DAW Feeder
Resin Feeder
Cold Crucible Melter
Pipe Cooler
High Temperature Filter
Post Combustion Chamber
Off-gas Cooler
Scrubber
Reheater
Activated Carbon / HEPA Filter
Extraction Fan
Reheater
DeNOx System
Stack
Schematic Diagram of Vitrification Facility
2. High Density Spent Fuel Storage Rack
The APR1400 adopts a high density fuel storage
rack by using neutron poison material. The
storage racks are stainless steel honeycomb
structures with rectangular fuel storage cells. In
order to maximize the spent fuel storage
capacity, the neutron absorbing material is
mechanically attached to the outside of each
individual cell and covers the full height of the
active spent fuel assembly. High density spent
fuel storage rack increases the capacity of spent
fuel storage approximately twice.
High Density Spent Fuel Storage Rack

CHAPTER 6

Reactor Containment Building Work
Modularization
Design Verification by 3D CAD System
APR 1400 Construction Schedule

56 Beyond your imagination_ APR1400 CHAPTER_6 CONSTRUCTION
Construction
A new construction schedule and constructability enhancement methods were developed based on the
experience gained from repeated OPR1000 constructions. The power block foundation of APR1400 is
seismically enhanced with the application of 0.3g Safe Shutdown Earthquake (SSE) as a Design Basis
Earthquake (DBE). Reactor Containment Building (RCB) and the auxiliary building are built on a common
basemat. This design requires a highly increased mat size and the amount of concrete. Thus, the
construction method for this massive concrete structure is reviewed to meet the target duration. The
common basemat foundation is simplified as a flat type so that it may benefit the concrete works.
Modularization has been introduced to reduce the construction period and the cost. There are three types of
modules: the structural module, mechanical equipment module, and composite module. To expand the
modular construction, the research is being done for the Steel-plate Concrete (SC) structure module, the
mechanical equipment, and the composite module. If the composite module is applied to all buildings in
the nuclear power plant, the construction period will be dramatically reduced to less than 40 months
through prefabrication at both the factory and the site.
1. Reactor Containment Building Work
There are two ways to bring big components such as steam generators into the RCB and to place them in
the proper location. One is the Over the Top Method (OTM), in which massive equipment is placed into its
proper position through the top of the RCB by a large capacity crane. The other is the conventional Side
Method (SM), in which major components are brought into the RCB through an equipment hatch and
positioned with a polar crane in containment. Since the SM requires two steps, it takes longer.
Since the RCB is wrapped around the auxiliary building and has bigger and heavier components than those
of conventional nuclear power plants, the OTM is favorable for the installation of major components.
To reduce the construction duration, the OTM has been adopted after a comparative study and assessment.
The detailed installation procedure has been verified through simulation by using 3-D CAD models to
confirm its feasibility.
Shear Wall SC Modules in RCB
Over the Top Method

Beyond your imagination_ APR1400 57
2. Modularization
To reduce the construction schedule, fabrication work at the factory for mechanical and electrical
equipment needs to be increased. Approximately 80 items of the APR1400, including auxiliary and
containment building water chillers and pumps, feedwater pumps and turbine drives, charging pumps,
turbine building component cooling water heat exchangers, and condensers, have been identified to be
suitable for modularization.
Reactor internal assembly is manufactured into three pieces in the conventional plant: Core Support Barrel
(CSB), the Lower Support Structure (LSS) with Core Shroud (CS), and the Upper Guide Structure (UGS).
The assembling of reactor internals at the construction site takes longer and is on the critical path of the
construction schedule. Reactor internals of the APR1400 are fabricated into two parts by integrating the
CSB, LSS, and CS. This modularization of reactor internals is estimated to reduce the construction
schedule by approximately 3 months.
In the condenser modularization, three shells and transitions are assembled in the factory, and the low
pressure feedwater heaters and the water boxes are assembled at the construction site.
Reactor Internal Module
Condenser Module

58 Beyond your imagination_ APR1400 CHAPTER_6 CONSTRUCTION
3. Design Verification by 3D CAD System
To successfully accomplish the APR1400 development from conceptual design to construction, the entire
plant design process has been reviewed by using a 3D CAD model. Design output was produced with a
frozen model after the verification with the 3D CAD model. This design verification improved both the
quality and the timeliness of the project design.
Tri-dimensional Design Verification System
We developed a new 3D CAD system called Tri-dimensional Design Verification System (TDVS) to
improve and streamline the existing engineering process for the main 2D design work and subsidiary 3D
review work. In the TDVS, all engineers should use 3D models at every stage of the design process and
review the 3D models from various points of view, and produce deliverables based on the verified 3D
models. Each 3D design is controlled and managed through the TDVS to implement design work
procedures, and to share and distribute the correct information to the right people without delay.
3D Design Verification System
Modeling
Engineers of each discipline make 3D CAD models themselves. The 3D CAD system is connected with
the Engineering Data Base (EDB) system. When the engineer routes the pipe with 3D CAD modeling
software, the data for this pipe line, such as line number, line size, specifications, pressure and temperature,
come from the EDB so that engineers don't need to input this engineering information.

Beyond your imagination_ APR1400 59
Review
Review of 3D CAD model is done with various exclusive types of software to verify design information
and configuration
Interference Check
Engineers from each discipline can check physical
interferences using exclusive interference detection software,
which can detect interferences automatically.
Interference Check
Animation
Since lots of concrete structures are built in a nuclear power plant, it is very important to check the
accessibility of the main equipment prior to construction. Therefore, engineers should use the animation
program to make a scenario and check ingress and egress path of equipment.
Walk-through
The virtual character technique of computer games is applied
to plant design. An engineer enters the 3D integrated model as
a virtual character, and can look for various components, as if
the engineer is actually performing a walk-through inspection.
Virtual Walk-through
Data Navigation
The 3D CAD system is connected to the EDB and Drawing and Document Management System (DDMS).
When an engineer reviews a 3D CAD model, the engineer can review the engineering information and
related drawings together at the same time.

60 Beyond your imagination_ APR1400 CHAPTER_6 DESIGNCONSTRUCTION
Production
Engineers can produce the following various deliverables by using a verified 3D CAD model.
Piping design drawing
After verifying the piping model, an engineer can create a piping plan and section drawing with exclusive
software. The software performs hidden-line removal to convert the view from an incomprehensible wireframe
into a standard line-drawing. The software allows automatic annotation and dimensioning of the
drawing.
Piping isometric drawing
This Piping Modeling software extracts and converts piping data and transfers them to drawing generation
software. The degeneration software creates fully annotated piping fabrication isometric drawing with the
related bill of material.
Pipe support drawing
After verifying the support model, an engineer can create support drawings. Drawing generation software
creates plan, section, and isometric views on the drawing with the related bill of material.
Cable tray layout drawing and HVAC duct layout drawing
After verifying the piping model, the engineer can create a piping plan and section drawings with exclusive
software. The software allows the semi-automatic annotation of the drawing. The designer points at items
in the drawing and the software retrieves their label from the 3D CAD model and allows the designer to
place them in the drawing.
Other deliverables from 3D CAD model
Bill of material
Piping spool index and location
Welding data and location
Design review data
Discrepancy list (between model and database )
Analysis data deck

Beyond your imagination_ APR1400 61
4. Apr1400 Construction Schedule
Along with the many new construction methods, the modularization of reactor internals and the mechanical
and structural composite modularization technologies have been applied to the construction of APR1400.
The Over the Top method allows the major components in the containment to be manufactured as large
modules and installed in the early phase of construction. The modular construction method is applied to the
Containment Liner Plate (CLP) and Stainless Steel Liner Plate (SSLP) to reinforce the steel and the
structural steel module. This method is also applied to the fabrication of equipment such as the reactor
internals and the condenser.
The deck plate construction method is applied for the construction and the installation of mechanical and
electrical equipment to be carried out simultaneously in the auxiliary building and compound building.
Thus, it is estimated that Shin-Kori 3 & 4 will be constructed in less than 51 months for the first unit and 49
months for the second unit. By using the new construction method, it is expected that succeeding units
could be constructed within 41.5 months.
F/C
34
3.5 13.5 17 4 3.5
38
38
41.5-Months Construction Target Schedule

CHAPTER 7
PLANT OPERATION &
MAINTENANCE

Improvement for In-Service Inspection
Enhanced Refueling Work
Design Feature for Reducing Unplanned Trip
Excellent Operation Performance of Korean Nuclear Power Plant

64 Beyond your imagination_ APR1400 CHAPTER_7 PLANT OPERATION & MAINTENANCE
Plant Operation & Maintenance
The APR1400 incorporates a number of design modifications and improvements to meet the utility's needs
for enhanced safety and economic goals and to address new licensing issues such as the mitigation of
severe accidents. In addition, the APR1400 design is optimized to achieve high operation performance and
to enhance the convenience of maintenance by incorporating the following improvements. Availability
factor of 92 % and more is achieved for the APR1400 over its design lifetime.
1. Improvement for In-Service Inspection
The reactor head is manufactured as one piece by integrating the flange and upper shell based on the
advanced forging capacity of the manufacturer. In the conventional plant, the flange and upper shell are
fabricated separately and welded to each other. This new forging technique reduces the girth seam, which
reduces the amount of In-Service Inspection (ISI) that has to be performed over its lifetime. Also, work
platforms are installed to enhance the convenience of ISI for steam generators.
2. Enhanced Refueling Works
The fuel handling devices are improved to reduce
the refueling time. In particular, a fuel transfer tube,
which connects the containment building and the
fuel handling area in the auxiliary building, is
improved to be opened quickly by remote control so
that the exposure dose is reduced. In addition, a
temporary fuel storage rack can be installed inside
the refueling pool to be used under an abnormal
condition during the refueling. The design of the In-
Core Instrument (ICI) cable tray is improved to not
be installed and disassembled for every refueling.
This improvement reduces the polar crane load and
simplifies the task related to ICI cable.

Beyond your imagination_ APR1400 65
3. Design Feature for Reducing Unplanned Trip
Primary System
To reduce unplanned reactor trips, the core thermal margin is increased by more than 10 % through
lowering the core outlet temperate and increasing the RCS coolant flow. In addition, the pressurizer volume
relative to power is enlarged to enhance the capability of coping with the transients.
Turbine
The turbine rotor is manufactured as one piece by forging to reduce the susceptibility of Stress Corrosion
Cracks (SCCs). The turbine control system is improved to enhance the reliability and maintainability by the
redundant design of controllers and the strengthening of the diagnostic functions.
The vibration monitoring functions are improved by strengthening the self-diagnostic functions of the
detectors and multi-directional measurements. In addition, earthquake-proof structures are installed to
prevent a turbine trip caused by high vibration.
Generator
Static excitation type is adopted to reduce mechanical wearing. The Auto-Voltage Regulator (AVR) is
placed in a dedicated room to minimize its malfunction by protecting it from heat and humidity. Also, the
filtration abilities of the stator cooling water pipelines are strengthened to not heat up by the reduced
coolant flow.
Feedwater System
The feedwater flow control system is designed to control the feedwater flow automatically over the full
operation range and to operate three turbine driven main feedwater pumps during normal power operation.
When one main feedwater pump is tripped during the full power condition, the other two main feedwater
pumps would be able to provide the total feedwater flow to the full power condition. This design reduces
unnecessary power cutback and unplanned turbine trip.

66 Beyond your imagination_ APR1400 CHAPTER_7 PLANT OPERATION & MAINTENANCE
4. Excellent Operation Performance of Korean Nuclear
Power Plant
Outstanding Performance of Nuclear Power Plants
Korea has made great strides in achieving
capacity factor over 90% and One Cycle
Trouble Free (OCTF) operation. The
outstanding performance exceeding the
global average could be attributed to the
excellent operation and maintenance
techniques accumulated for the last 30
years.
Highest Class Capacity Factor in the World
Korea made great achievements in nuclear power plant operation in 2008, including a nuclear capacity
factor of 93.4%. This capacity factor is approximately 14 % higher than the world average. This
achievement resulted from the dedication of Korea's highly trained engineers and their prominent operation
and maintenance techniques.
Overseas Nuclear Energy Group
Korea is a member of three nuclear energy owners groups, PWROG, COG, and FROG. It contributes to
improving the reliability of Korean nuclear plants by exchanging technical information on operation
experience, benchmarking nuclear power plants by overseas members, taking part in technical meetings
and seminars, and implementing joint projects to resolve common problems facing nuclear energy owners
such as receiving licenses and endorsements for operation. Taking part in the steering committee meetings
or annual meetings of the groups has led to the identification of issues on the nuclear plant operations of
overseas members to improve the operation capability of domestic plants. Korea hosted the steering
committee meeting of FROG and the technical committee meeting of COG in 2003 and 2006, respectively

Beyond your imagination_ APR1400 67
Atomic Energy Research and Development Plan
The atomic energy research and development plan, supervised by the Ministry of Knowledge Economy
(MKE), has propelled Korea to be an advanced country in nuclear power technology in the 21st century. In
general, the financial resources would be provided by nuclear power plant construction enterprises and
operation companies. Korea has paid the amount of KRW1.20/kWh from last year's earnings for a research
and development fund. Its major research and development achievements are the development of the
radioactive safety inspection unit for a nuclear power facility and nuclear power plant steam generator tube
inspection and maintenance robot development. In 2003, the radiation technology development project was
funded. In 2004, the hydrogen production system on nuclear power use was made for establishing hydric
energy production technology. In 2007, the Advanced Power Reactor + (APR+) development project was
launched to upgrade the technology of the APR1400 for the continuous enhancement of its safety and
economy.
In the future, the path of the atomic energy research and development plan will includes: system integrated
modular advanced reactors, proton base engineered technology, and hydrogen production gas turbine and
light water reactor type new atomic fuel development.

Korean Nuclear Group Synergy
NPP CONSTRUCTION MANAGEMENT, COMMISSIONING & OPERATION
Korea Hydro & Nuclear Power Co. owns & operates 20 units of nuclear power plant, with
additional 8 units under construction and ranks 3rd among the world nuclear power companies
in terms of capacity. KHNP has accumulated world-class nuclear technologies through the
construction, operation and maintenance of diversified nuclear reactors.
KHNP provides all necessary services to the utilities from conceptualization to commercial
operation of overseas nuclear power projects. Our optimized and integrated services ensure
maximum value for each project and provide partners with a competitive edge.
Korea Hydro & Nuclear Power Co., Ltd.
167 Samseong-dong, Gangnam-gu, Seoul, 135-791, Korea
Tel: 82-2-3456-2070, Fax: 82-2-3456-2819, E-mail: yoonyy@khnp.co.kr
www.khnp.co.kr
EQUIPMENT DESIGN & MANUFACTURING
Doosan Heavy Industries & Construction Co., Ltd., while being Korea’s only company that
specializes in nuclear power plants, employees the highest level of technology especially in the
area of nuclear power generation. With its turnkey production lines and management systems for
materials, design, construction, testing and services, maintenance and repair, Doosan is
strengthening its position as a leading provider of nuclear power systems in both the Korean and
international markets.
DOOSAN Heavy Industries & Construction Co., Ltd.
1303-22 Seocho-dong, Seocho-gu, Seoul, 137-920, Korea
Tel: 82-2-513-6327, Fax: 82-2-513-6678, E-mail: jaechoon.hwang@doosan.com
www.doosanheavy.com
PLANT DESIGN & ENGINEERING
Korea Power Engineering Company, Inc.(KOPEC) has developed its own nuclear reactor and
plant design engineering technologies in the field of nuclear and thermal power plants, and
standardized Korea nuclear/fossil power plants. Based on the generated output, nearly 60% of all
Korea’s power plants have been designed by KOPEC. KOPEC is now exporting its engineering
and design skills abroad helping the current nuclear renaissance. Clients countries include
advanced nations such as USA as well as others.
Korea Power Engineering Company, Inc.
360-9 Mabuk-dong, Giheung-gu, Yongin-si, Gyeonggi-do, 446-713, Korea
Tel: 82-31-289-3425, Fax: 82-31-289-3167, E-mail: jmyu@kopec.co.kr
www.kopec.co.kr

CORE DESIGN, FUEL DESIGN & FABRICATION
Korea Nuclear Fuel Co., Ltd. (KNF) provides all types of nuclear fuels and related services with
highly qualified resources and advanced technology.
KNF has also performed the initial core, reload core designs and safety analysis of all currently
operated PWR power plants in Korea. With valuable experience and state-of-the-art technology,
KNFC is ready to meet customer requirements for nuclear fuel.
Korea Nuclear Fuel Co., Ltd.
493 Deokjin-dong, Yuseong-gu, Daejeon, 305-353, Korea
Tel: 82-42-868-1320, Fax: 82-42-863-4430, E-mail: hjkim@knfc.co.kr
www.knfc.co.kr
PLANT MAINTENANCE & SERVICES
Korea Plant Service & Engineering Co., Ltd. (KPS) provides high quality maintenance services
for nuclear power and industrial plants throughout the Korea and overseas markets. KPS covers
full range of maintenance for nuclear power plants and other various types of power plants
including thermal, hydro, combined cycle/co-generation such as maintenance in commissioning
phases, routine maintenance, planned outage maintenance and modification & rehabilitation in
operation.
Korea Plant Service & Engineering Co., Ltd.
Migum-ro 1, Bundang-gu, Seongnam-si, Gyeonggi-do, 463-726, Korea
Tel: 82-31-710-4441, Fax: 82-31-710-4449, E-mail: junhk@kps.co.kr
www.kps.co.kr
NUCLEAR TECHNOLOGY R&D
As a government-sponsored R&D institute, the Korea Atomic Energy Research Institute
(KAERI), over the past fifty years since its establishment in 1959, has been greatly devoted to
the self-reliance of nuclear technologies in Korea. KAERI performed the development of various
types of nuclear fuel technologies and the design of PWRs. KAERI also successfully designed
and constructed a multi-purpose 30MW research reactor ‘HANARO’, which plays a critical role
in nuclear R&Ds. One of the major projects KAERI is now focusing on is the development of a
small-and-medium sized reactor ‘SMART’, which can be used for electricity generation, sea
water desalination, and district heating
Korea Atomic Energy Research Institute
150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353, Korea
Tel: 82-42-868-2000, Fax: 82-42-868-2702, E-mail: webmaster@kaeri.re.kr
www.kaeri.re.kr

Conclusion
The customers would gain the following profits through introducing the APR1400;
First, the owner could have the APR1400 with low construction cost and short construction
time by virtue of the large amount of nuclear power plant design and construction techniques
available in Korea.
Second, the experience and expertise in design, manufacturing, and project management
accumulated for the past three decades could facilitate the timely delivery of equipment, the
prompt provision of maintenance services, and the proper support for construction and
operation in compliance with the requirements of customer.
Third, the owner could be supported in self-reliance and localization of nuclear technology.
Korea is a unique country which has the experience from introduction of nuclear power plant in
turn-key contract to development of new concept reactor.
Fourth, the customer could apply advanced nuclear technologies to the plant operation and
maintenance. Korea consistently develops and upgrades nuclear technologies through national
research and development plan.
Korea is a unique country which has steadily constructed and updated nuclear power plants since
the 1980s. Based on our self-reliant technologies and experiences from the design, construction,
operation and maintenance of OPR1000, the APR1400 has been developed by adopting
advanced design features to enhance the plant safety and economical efficiency. The APR1400 is
an embedment of all these experiences and technologies for the past 30 years in Korea. Thus, we
can definitely assert that the plant safety and the economic
efficiency of the APR1400 is the best in the world.
APR1400 is the best choice for customers who
want an advanced, safe, and proven nuclear
power plant.

You need more copies? Any questions?
Please contact :
Overseas Nuclear Projects Development Department, KEPCO
TEL : 82-2-3456 - 6400
FAX : 82-2-3456 - 6469
www.kepco.co.kr / www.apr1400.com