Overview of Mitsubishi Advanced PWR 2... · APWR design to improve economy, safety, reliability,...

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved.８AS-EXP-20130003

Overview of

Mitsubishi Advanced PWR

IAEA INPRO 7th Dialogue Forum in Vienna

Sumio FUJIIActing General Manager,

Nuclear Systems Engineering Department

November 19, 2013

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Remarks on proprietary and copyright information

• The data and information in the presentation materials are proprietary of

Mitsubishi Heavy Industries.

• Utilization of the data and information is limited to the purpose of INPRO

7th Dialogue Forum.

• The data and information concerning advanced PWR in the presentation

materials are never permanent. They may be changed in the process of

nuclear power plant construction for individual site.

• Mitsubishi Heavy Industries has the copyright of all figures and photos in

the presentation materials.

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 1

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Contents

1. Development History of Mitsubishi PWR

2. Advanced PWR for Global Deployment

3. Safety of Mitsubishi Advanced PWR

4. State-of-Art Technology for Digital Instrumentation &

Control System of Mitsubishi Advanced PWR

5. Safety Enhancements for Beyond Design Basis

Accident of Mitsubishi Advanced PWR

6. Construction Management of Mitsubishi Advanced

PWR

7. Operation and Maintenance Management of


8. Conclusion


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Development History

of Mitsubishi PWR


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Line-up of Mitsubishi PWR for Japanese utilities

Unit C/O Outpt

Mihama1 1970.11 340 MkWe


Genkai1 1975.10 559 MkWe

Ikata1 1977.09 566 MkWe


Ikata2 1982.03 566 MkWe

Tomari1 1989.06 579 MkWe


Unit C/O Output

Takahama1 1974.11 826 MkWe

Takahama2 1975.11. 826 MkWe


Sendai1 1984.07 890 MkWe



Senfai2 1985.11 890 MkWe

Ikata3 1994.12 890 MkWe


Unit C/O Output

Ohi1 1979.03 1175 MkWe Ohi2 1979.12 1175 MkWe

Turuga2 1987.02 1160 MkWe

Ohi3 1991.12 1180 MkWe

Ohi4 1993.02 1180 MkWe Genkai3 1994.03 1180 MkWe


Tsuruga3 201X～ 1538 MkWe

Tsuruga4 201X～ 1538 MkWe

2Loop（（（（300-600MWe）））） 3loop（（（（900-1000 MWe））））4Loop（（（（1200-1500MWe））））

Tsuruga 3 & 4 are Advanced PWRs.


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Development of Advanced PWR (J-APWR)

• The development of the Advanced PWR (APWR) was started in late

1980s as a joint cooperative development project by five Japanese PWR

owner utilities and Mitsubishi Heavy Industries (MHI), financially

supported by the Ministry of International Trade and Industry (Currently

the Ministry of Economy, Trade, and Industry) as a part of the Phase III

Improvement and Standardization Program of Japanese PWR.

• The design of the APWR was based on MHI’s conventional 4-loop plant

technologies, on which MHI has accumulated significant operating

experiences, and was scaled up to achieve higher electrical output.

• In addition to adopting those proven technologies after the first step

development, further modifications were also made on the prototype

APWR design to improve economy, safety, reliability, operability, and

maintainability by incorporating advanced technologies.

• The first APWR plant (Japanese APWR) is Tsuruga-3 producing 1538

MWe operated by the Japan Atomic Power Company. The Tsuruga-3 is

now under safety review for construction permission, which got

suspended due to the Fukushima Daiichi accident.


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GⅢ＋ Reactors Development

70’s 80’s 90’s 2000’s 2010’s 2020’s

Development & Improvement of PWR

Technology

Enhanced up to APWR

APWR Tsuruga 3, 4licensing process

US-APWRUS NRC Licensing

EU-APWR

US UtilitiesComanche Peak

3, 4

European Utilities

ATMEA1

by ATMEA (Joint Venture)

European and Global Utilities

AREVA


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Development of US-APWR & EU-APWR

• By referring the J-APWR, MHI has designed the US-APWR which meets

regulatory requirements in United States of America as well as Utilities

Requirements Document (URD). The turbine-generator system and all

electric equipment of the US-APWR are designed for 60 Hz electric grid.

• Basic designs of the US-APWR except for the fuel length are same as

those of the J-APWR whose design has completed. The US-APWR has

been developed as a larger-output version of the J-APWR, aiming at

higher electrical outputs and improved economics, by modifying some

design features mainly in the secondary side without increasing core

thermal output.

• The fuel length was changed to 4.2 meters in place of 3.66 meter, and

the electric output was increased to about 1700 MWt.

• EU-APWR is a sister plant of the US-APWR and is aiming to satisfy

European Utilities Requirements (EUR). Nevertheless the turbine-

generator system and all of electric equipment are designed as 50 Hz,

designs of core and other major equipment on the reactor coolant system

are same as those of the US-APWR except the reactor coolant pumps.


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Advanced PWR for

Global Deployment


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US-APWR & EU-APWR : Course of Design


� To contribute toward safe, stable

and flexible power production

through satisfying American

“Utilities Requirements Document”

and “European Utilities

Requirements”.

A course of design for the US-APWR and EU-APWR is to;

�To achieve the best performance in safety, reliability,

operability, and maintainability by incorporating advanced

technologies.

�To provide the world’s largest NPP for large electric grid

utilities aiming at economical production costs.

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Major Features on safety and reliability

� Top-mounted ICIS for avoiding penetrations at

the RV bottom

SHSH

SHSH

RWSP

RVACC ACC


� Full 4-train safety systems

with best-mix of passive

and active systems, which

allows on-power

maintenance

� Full digital I&C technology enabling

one-man operation

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Major Features on economy

� Large reactor producing thermal output of 4,466MWt

� 14-ft fuels creating additional thermal margin and flexible core operation

� 14-ft fuels making 24-month operationwithout deterioration in fuel economy

� Enhanced SG heat transfer performance by enlarged heat transfer area with triangular lattice arrangement of SG tubes

� High-performance steam-water separators generating high quality steam in SG

� High performance LP-turbine having last stage blades of 70 inches length

� Secondary system enabling high thermal efficiency over 36%


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Comparison of Major Specifications


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Reactor Vessel and Internals


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Reactor Core


Fuel and Rod cluster control assemblies

A, B, C, D : Control group bank

SA, SB, SC, SD : Shutdown group bank

Control rod drive

mechanism

Reactor

vessel

Fuel

assembly

Outlet

nozzle

Inlet

nozzle

Reactor Vessel Core design

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Major Core Parameters

Design limitsBest estimates

of initial core

Active Core

Active core equivalent diameters (cold) 3.89 m

Active fuel height (cold) 4.2 m

Hydrogen/Uranium atomic ratio (cold) 5.57

Core average linear power density 15.2 kW/m

Maximum linear power density 39.5 kW/m 31.2 kW/m

Nuclear enthalpy rise hot channel factor FN∆H 1.73 1.50

Delayed neutron fraction βeff (%) 0.44 to 0.75 0.50 to 0.69

Prompt neutron lifetime, l* (µsec) 8 to 20 14.0 to 15.3

Reactivity coefficient

Doppler power coefficient (pcm/%power) BOC

EOC

-12.4 to -7.4

-12.1 to -7.6

Moderator temperature coefficient (pcm/℃℃℃℃) -71.1 to -1.4

Moderator density coefficient (pcm/g/cm3) < 0.51x105 < 0.32x105

Boron coefficient (pcm/ppm) - -9.3 to -8.0

Neutron multiplication factor

Maximum core reactivity keff (BOC,cold,noXe) - 1.223

Maximum fresh fuel assembly k∞ - 1.456

Boron concentration (ppm)

Refueling boron concentration > 4000 -

Cold shutdown, BOC, noXe, ARI, keff<0.95 - 1850

Cold shutdown, BOC, noXe, ARO keff=0.99 - 1796

Hot shutdown, BOC, noXe, ARO keff=0.99 - 1706

Hot zero power, BOC, noXe, ARO keff=1.0 - 1579

Hot full power, BOC, noXe, ARO keff=1.0 - 1444

Hot zero power, BOC, equilibrium Xe, ARO - 1086

Shutdown margin

BOC

EOC> 1.6

2.94

2.15


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Major Thermal-Hydraulic Parameters

Design Parameters Design Values

Coolant condition

Primary coolant system pressure (MPa(absolute)) 15.51

Thermal design flow rate (ton/hr) 76,300

Effective flow rate for core cooling (ton/hr) 69,500

Reactor vessel inlet temperature (℃) 288.1

Average rise temperature in reactor vessel (℃) 36.9

Heat transfer at normal condition

Fraction of heat generated in fuel (%) 97.4

Core average linear heat rate (kW/m) 15.2

Maximum local peak linear heat rate at FQ=2.6 (kW/m) 39.5

Power density (kW/l) 89.2

Specific power (kW/kg uranium) 32.0

Minimum DNBR by WRB-2 correlation

At nominal condition; for typical hot channel

for thimble hot channel

2.05

1.98

During AOO; for typical hot channel

for thimble hot channel

> 1.35

> 1.33

Maximum fuel centerline temperature during AOO (℃℃℃℃) < 2548


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Safety of



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Structure of Emergency Core Cooling System

Conventional 4-loop PWR US-APWR

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 18S

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RWSP

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Features and Strengths in ECCS

� Four independent safety trains

• Four-train mechanical safety systems

• Four-train electrical safety systems

• No inter-connecting piping between trains

� Direct vessel injection (DVI)

• The cooling water from the safety injection pumps is directly injected into

the reactor vessel.

� Refueling water storage pit (RWST) inside the containment

• Operation of changing the suction from the RWST to the containment

recirculation sump is eliminated.

� Advanced accumulator (ACC)

• An advanced accumulator is connected to each cold leg to refill the

reactor vessel lower plenum and down-comer immediately after a LOCA.

� No low-head safety injection system

• Installation of ACCs enabled us to eliminate low-head safety injection

systems, which resulted in reduction of the number of active components.


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Schematic Flow Diagram of ECCS


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Advanced Accumulator having two flow rates

Nitrogen

Injection

Water

Flow Damper

High Flow Rate Mode

Injection Water

Nitrogen

Flow Damper

Low Flow Rate Mode

Main stand pipe

Side inlet

Side inlet

• A short while high flow rate of cooling water is required just at early stage

of LOCA, a large volume of water is injected to fill up the down-comer.

After that, the flow rate is reduced to the amount necessary for decay heat

removal. Advanced accumulator system has a longer supplying duration

that can cover the time expected to the low pressure injection system.

• This design makes it possible to eliminate the LPIS, which leads to the

reduction of active components.

Conventional Design

Advanced Accumulator


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State-of-Art Technology

for Digital Instrumentation &

Control System of



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Full Digital Instrumentation & Control System

• Fully digitalized technology is applied to the instrumentation and control

system (I&C system) for both safety and non-safety functions.

• The reactor protection system and actuation system of engineered

safeguard features (ESF) have 4-time redundancy in the I&C system.

• Conventional operating and monitoring devices such as switches and

indicators have been eliminated, except devices for diverse actuation

system (DAS).


System Features

Main control board

・Fully computerized

・Visual Display Units (VDU) for safety and non-safety channels

・Limited number of conventional switches and indicators

Safety I&C

・Fully digitalized with Mitsubishi digital controllers

・Four-train redundancy in reactor trip system, ESF actuation system and safety logic

system for component control

Non-safety I&C・Fully digitalized with Mitsubishi digital controllers

・Duplex digital architecture for each control and process monitoring sub-system

Data communication

・Fully multiplexed, including class 1E signals

・Multi-drop data bus and serial data link

・Fiber optics communication networks for noise immunity and required isolation

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Full Digital Instrumentation & Control System


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• A large display panel is installed

to display major parameters for

normal and abnormal situations.

Thus, the current status of the

entire plant can be understood by

all crews and their communication

is improved.

• Thus, the digital system provides

significant benefits to the safety of

nuclear power through reduction

in operation works and

maintenance work loads, which

reduces the potential for human

error.

Human Engineering & Control Room Design

• Plant parameters and associated operating switches are displayed on

same screens, and touch screen operations are applied. Thus, the

operator’s work load is reduced and the reliability of operation is

increased.


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Diverse Actuation System (DAS)

• Diverse Actuation System (DAS) is installed to

provide back-up actuations for safety and non-

safety components as countermeasures against

common cause failure (CCF) in the software of

the digital I&C system.

• The DAS was designed for beyond design basis

CCF incidents. The DAS consists of diverse

devices from the digital safety system so that a

CCF in the digital system doesn’t impair the DAS

functions.

• The DAS initiates safety functions independent

from the output of the digital safety system.

Manual actuation is provided for all functions.

Automatic actuation is also provided for functions

if the time for manual operator action is

inadequate.


Diverse Human-System-

Interface (HSI) Panel

Switches

Indicators

Alarms

Diverse Human-System-

Interface (HSI) Panel

Switches

Indicators

Alarms

Automatic actuation

system

Automatic actuation

system

Diverse

Trip

Diverse

Trip

Each

Remote

I/O

Each

Remote

I/O

Components

such as pumps, valves

Components

such as pumps, valves

Measured

parameter

signals

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Safety Enhancements for

Beyond Design Basis

Accident of



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Design against Airplane Crash

� Countermeasures against Airplane Crash

• The US-APWR is designed against the airplane crash (APC) by taking

into account the potential effects of the impact of a large commercial

airplane for the building strengths and plant layout.

• The countermeasures for the APC in the US-APWR design:

- Physical / functional separation and segmentation in the layout design

for safety related structures, systems and components (SSCs)

- Reinforcement of structural design of the buildings, which contain

safety related SSCs

• It was confirmed by the assessment that the inherent robustness of the

US-APWR design would be maintained for the followings;

- The reactor core remains cooled or the containment remains intact

- Spent fuel pit integrity is maintained


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� Prevention of severe accidents

1. Reduction of potential precursor of design basis accidents and severe

accidents.

• Reduction of latent LOCA precursors. E.g.;

- Adoption of top-mounted in-core instrumentation system (ICIS) to

eliminate penetrations at the reactor vessel bottom.

- Reduction of reactor vessel welding seams and improvement of

welding methodologies

- Improvement of piping bypassing the containment to a higher rating to

reduce probability of interface LOCA

2. Enhanced reliability in safety functions

• Installation of four-train structure for safety systems

• Installation of refueling water storage pit (RWSP) in the containment

• Installation of advanced accumulators to reduce the number of active

components

• Countermeasures against Anticipated Transient Without Scram (ATWS).

E.g.; Installation of diverse actuation system to cope with common cause

failure in the reactor protection system

Countermeasures against Severe Accidents


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• Countermeasures against station blackout (SBO)

E.g.; Installation of four-train emergency gas turbine generators (GTGs) and

additional alternate power source by two-train GTGs.

• Countermeasures against fire

E.g.; Clear physical separation between the redundant trains of safety

systems

• Countermeasures against the intersystem LOCA

E.g.; Prevention of over-pressurization of the residual heat removal system

� Frequency of severe accidents (PRA result for core damage frequency)

• The result of calculation for US-APWR core damage frequency (CDF)

meets the NRC goal and the Utility Requirements Document (URD) goal.


Requirement NRC URD US-APWR

CDF 1×10-4/RY 1×10-5/RY 3.0×10-6/RY


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� Mitigation of severe accidents (To keep the integrity of the containment

vessel)

• For debris dispersion, reinforcement of depressurization function of the

primary system and an improvement of RV cavity geometry are

considered.

• For quasi-static over pressurization, an usual containment vessel air

recirculation system and an alternative containment spray supplied from

the fire service water system can be used. These systems can be used to

reduce the pressure if the containment spray system is not available.

• As countermeasures to avoid containment vessel damage due to

hydrogen combustion, a hydrogen control system (igniters) is installed to

control the hydrogen concentration.



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Construction Management

of Advanced PWR


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Total Engineering Capability with “Single Responsibility” in the full areas;

� Conceptual/Basic/Detailed Engineering

� Manufacturing of components

� Erections and installations of structures, systems and components

� Support of human resource development for operation & maintenance

� Technical support after start of commercial operation

Construction Management of Advanced PWR (1)


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Material procurement

and management

CAM*

Welding

inspections

On-site

installation

inspections

Construction

process

management

•CAM : Computer-Aided Manufacturing

Integration


� To keep quality and schedule,

MHI’ Data management system

integrates;

� Design

� Procurement

� Manufacturing & Inspection

� Construction


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Reactor vessel is

now under processing in

upright installation

position with high-

accuracy & high-quality.

Major Components (Reactor Vessel, Steam Generator, Reactor

Coolant Pump, Reactor Internals, Control Rod Drive Mechanism,

Pressurizer, Turbine, etc.) are manufactured in our hands.

Works and machines have been updated/enlarged and are

prepared for the global deployment.

(Photo)

A reactor vessel is

manufactured by a super-

large combined machine

tool named “Super Miller”.


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�Rational designs like;• Internal structures using steel plate

reinforced concrete (SC) (left)

• Large modular (prefabricated) block

construction (Right)

�Tools for efficient construction like

Super-large-capacity cranes

which enables on-site welding and

formation of containment vessel.

�Ability for comprehensive coordination

of civil & installation work

40m-dia.

upper

containment

Brilliant Successes

(1st Concrete to Fuel Loading)

2 loop : 34.5 months



Reduction of on-site work volume and construction period


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Operation and Maintenance

Management of



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・・・・Application of world’s first submerged automated guided vehicle with robot・・・・Using a manipulator with 7 shafts

Inspection Technologies for Advanced PWR

Pressurizer

RV

Piping inspection unitUT / ECT for BMI nozzle inspection unit

TOFD-UT ( for sizing )

RV head nozzle inspection unit

ECT probe for J-type welds surfacePhased array type probe for

J-groove welds

SG tubes inspection unit

•Lightweight and compact unit

•Point focus phased array enables high accuracy flaw sizing. (Approx.±±±±3mm)

Censer

Inspections can be done for 4 days by utilizing two units simultaneously. (for 4 loops)

One unit （（（（more than 3,000 tubes）））） can be inspected for around 4 days.

RV inspection unit A-UT machine

MHI point focus probe Automated UT unit for piping

Achievement of high accuracy sizing with point focus

capability

Phased array censer

SG

ECT (for detection：：：：0.5mm)


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Conclusion

MHI developed Advanced PWRs based on accumulation of half-century experiences in construction and operation of conventional PWR power plants.

Proven and advanced technologies have been introduced in APWR to improve economy, safety, reliability, operability and maintainability.

Advanced accumulator was employed to reduce the number of active components, and low pressure injection system was eliminated.

US- and EU-APWR were developed to let them meet regulator’s and user’s requirements including flexible operation. The fuel length was changed to 4.2 meters.

ECCS and supporting systems have 4 divisions in redundancy.

Due to enhancements in safety systems, the total of the CDF is estimated 3.0x10-6 /RY in a PRA calculation.

MHI’s manufacturing and construction management is integrated to one data base system to accomplish them with high quality, high reliability and strict keeping of schedule.


© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved.８AS-EXP-20130003

Thank you for kind attention!

Design Control Documents of the US-APWR are found on;

http://www.nrc.gov/reactors/new-reactors/design-cert/apwr.html

Overview of Mitsubishi Advanced PWR 2... · APWR design to improve economy, safety, reliability,...

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