For the Long-Term Stable
Supply of Electric Energy-Yesterday, Today, and Tomorrow of
Nuclear Power GenerationYOSHIO MIYAKEHIROSHI MUKAI

ing the highest safety level. As one of specific efforts toattain this purpose, Mitsubishi Heavy Industries, Ltd(MHI) has been proceeding with the design of the newplants based upon the experience accumulated to date,meanwhile taking measures for maintaining the stable
operation of the aged plants. Also MHI has been focusing to develop technology required and measuresavailable for the effective use of equipments and plants.
In the mid- and long term planning, MHI is pursuing the development of advanced reactor, including
realization of local distributed electric power sources
as well as multi-purpose energy sources in order to meetthe diversifying needs for future nuclear power plants.
This paper presents an overview of the current activities of MHI in these area, including a brief history of
introduction and acquisition of light water reactor techFig
nology (Fig
Fig.. 1 ).

1. IntroductionMore than thirty years have passed since nuclear
power generation started on commercial basis in Japan.
Beginning with introducing the technology from abroad,acquisition and improvements of the technology havebeen progressed in every field of design, manufacturing,
construction, operation and maintenance through thediligent efforts of all of the relative people concerned.
As a result, now the 52 light water reactor plants supply approximately one-third of the electric power of ourcountry and play an important role as major electric
power source.
For continuing to play the role successively, it is important that the light water reactor plants assurecompetitiveness against other electric power sources,
such as combined cycle thermal power plants, maintain-

Category

Light-water
reactor(PWR)

1980's1970's[Newly established plant]
1st generation PWRs

1990's

2nd generation PWRs

Standardization of LWR improvements by MITI

3rd generation PWRs

Next generation plant

Advancement of LWR technology by MITI

Future LWR

Replacement and renewal of large equipment (SG etc.)
Advancement of operationand maintenance[Operation plant][FBR]

Correspondence to aging,long cycle operation,
increase in electrical output

MKIII

Construction of Joyo
Advanced
reactor2000'sConstruction of Monju[Advanced thermal reactor]
Construction of Fugen

Advanced loop type FBRConstruction of high-temp.
[High temp. gas[Fusion
engineering test reactor
cooled reactor]
reactor]Development of nuclear fusion test apparatus

Helium gas turbine
power reactor

ITER[Nuclear fuel]
39 GWd/t fuel

48 GWd/t fuel (step 1)55 GWd/t fuel(step 2)

Fuel withhigher burnup
MOX fuel

Construction of Reprocessing Plant (Rokkasyo-mura)[Fuel cycle]
Nuclear fuel andfuel cycle

Tokai
Reprocessing
Plant

Development of innovative reprocessing technologies[Cask]Development of transport and storage cask
Low-level Radioactive Waste Disposal Center

Spent FuelInterim Storage
Facility

Development of radioactive waste treatment and disposal technologies
Fig. 1 Progress of nuclear energy systems of MHI

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

1

eration plants such as Mihama Unit 1, MHI had promoted domestic production of equipment andimprovement of the technology introduced in early stage
in line with actual domestic conditions. During the ten
years plus after nuclear power generation was put in
use, MHI experienced such troubles as bowing of fuelrod, cross flow from core baffle plate clearance and degradation of steam generator (SG) tubes which resulted
in decrease of plant availability. To overcome these problems, MHI made thorough investigation of the root
causes, and took comprehensive measures such as improving the design of equipment in terms of materials
and structures, improving operational controls, water
quality control for example, as well as developing andadopting the improved inspection and corrective maintenance techniques.
Some measures taken by MHI against degradation to
SG tubes are presented below as an example case fortechnical improvements.
The steam generator is a vertical U-tube heat-exchanger to transfer the heat generated in the reactor
core to the secondary turbine-generator system, andformed of several thousands of the tubes. Since the introduction of the technology from Westinghouse, MHI
had suffered from various types of degradation withtubes. Countermeasures against those troubles implemented by MHI were improvement of the water quality
in the secondary system, structural modification of the

2. Yesterday
Yesterday--From introduction to taking root of lightwater reactor technology
2.1 Introduction of light water reactor technology
MHI entered into an agreement with Westinghouse
Electric Company of the United States in 1959, and introduced technology on the pressurized water reactor(PWR) as the commercial nuclear power plant. Basedupon those technology, MHI has been disseminating
PWRs throughout Japan. MHI has thus far built total of
23 PWR plants over years, including Mihama Unit 1,
which entered into commercial operation in 1970 and
first supplied nuclear generated electric power to EXPO
Osaka held in 1970, and Genkai Unit 4 , the most recent
PWR plant which commenced commercial operation in
Table 1).
1 These PWR plants can be classified into
1997 (Tthree generation groups, the first group started commercial operation in 1970's, the second in 1980's and thethird in 1990's. The first generation group including nineunits were imported plants or domestic plants constructed based on the imported technology. The second
generation group includes seven units constructed basedon MHI's own technology that were developed throughthe experience in construction and operation of the firstgroup plants. The third generation group is consisted ofseven units which have been further improved based on
domestic technology.
Ever since introducing the technology for the first gen-Table 1 Japanese PWR plants
Type

Plant Name
Mihama No. 1*
Mihama No. 2
Genkai No. 1

Electric
Output
(MWe)
340
500
559

2-loop Ikata No. 1
PWR Genkai No. 2

566

Ikata No. 2
Tomari No. 1
Tomari No. 2
Takahama No. 1*
Takahama No. 2
Mihama No. 33-loop Kawauchi No. 1
PWR Takahama No. 3
Takahama No. 4
Kawauchi No. 2
Ikata No. 3
Ohi No. 1*
Ohi No. 2*
Tsuruga No. 24-loopPWR Ohi No. 3
Ohi No. 4
Genkai No. 3
Genkai No. 4

566

559
579
579
826
826
826
890
870
870
890
890
1 175
1 175
1 160
1 180
1 180
1 180
1 180

Commercial
Operation

OwnersThe Kansai Electric Power Co., Inc.
The Kansai Electric Power Co., Inc.
Kyushu Electric Power Co., Inc.
Shikoku Electric Power Co., Inc.
Kyushu Electric Power Co., Inc.

Nov. 1970
July. 1972
Oct. 1975
Sep. 1977
March. 1981
Shikoku Electric Power Co., Inc.
March. 1982
Hokkaido Electric Power Co., Inc.
June. 1989
Hokkaido Electric Power Co., Inc.
April. 1991
The Kansai Electric Power Co., Inc.
Nov. 1974
The Kansai Electric Power Co., Inc.
Nov. 1975
The Kansai Electric Power Co., Inc.
Dec. 1976
Kyushu Electric Power Co., Inc.
July. 1984
Jan. 1985
The Kansai Electric Power Co., Inc.
June. 1985
The Kansai Electric Power Co., Inc.
Kyushu Electric Power Co., Inc.
Nov. 1985
Dec. 1994
Shikoku Electric Power Co., Inc.
March. 1979
The Kansai Electric Power Co., Inc.
The Kansai Electric Power Co., Inc.
Dec. 1979
The Japan Atomic Power Company Co., Ltd. Feb. 1987
Dec. 1991
The Kansai Electric Power Co., Inc.
Feb. 1993
The Kansai Electric Power Co., Inc.
March. 1994
Kyushu Electric Power Co., Inc.
July. 1997
Kyushu Electric Power Co., Inc.

*: Primary system: Westinghouse Electric Company,
Secondary system: MHI,
Others of primary and secondary systems: MHI

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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tube supports, use of the improved materials for tubes
and improvement of tube manufacturing methods.
The purposes of the secondary system water quality
control are to prevent the SG tubes from being corroded
and the secondary system equipment from being agedegraded. The water quality control technology first
introduced from the U.S, was phosphate (PO4) treatment
which was on the extension of conventional thermal
power plants. However, this method caused a corrosive
reduction of heat transfer tube wall thickness that resulted from a concentration of phosphate at narrow
region (clevis) between the heat transfer tube and tube
support plate. In order to prevent this problem, phosphate treatment method was replaced with the allvolatile treatment (AVT) method which was expected toreduce the concentration at the clevis. After replacement,
although a stress corrosion crack (SCC) was experienced
due to the remaining phosphate at clevis, AVT method
had successfully precluded the heat transfer tube degradation due to phosphate.
Because the tube supports structures is susceptible
to the impurity condensation at clevis between tube andsupport plate, MHI determined to improve the shape ofthe hole on the support plate for reducing the enrichment of impurity. Based on the results of evaluations ofvarious characteristic tests, MHI devised the shape of
Broached Egg Crate (BEC) for holes which could reducethe clevis region and verified their concentration characteristics by tests. As a result, BEC holes were foundeffective for reducing the impurity enrichment to 1/100
or less compared to conventional round shaped holes.
Accordingly this revised design was applied to Takahama
Unit 3 and the subsequent plants as well as to the replacement steam generators. Another modification to
steam generator is for the anti-vibration bars (AVB)
which are inserted to suppress the vibration of U-bend
portion of tubes for preventing from the fretting wear of
tubes due to insufficient support. In this case, the num-bers of supporting points were increased.
In order to increase the strength of the tube against
SCC, research and development were carried out forchanging the conventional materials into new ones. As aresult, it was found that the resistance of Ni-Cr-Fe alloy
to SCC in the high temperature and high pressurizedwater increases as the amount of Cr was increased. After TT alloy (thermal treated alloy) 690 containing 27 to
31% Cr and subjected to the special thermal treatment(aging treatment for approximately 15 hours under
700oC) after full solution treatment was verified to haveexcellent SCC resistance, it was adopted for Ohi Unit 3
and the subsequent plants as well as for the replacement steam generators.
As for manufacturing methods, stress corrosion cracking from outside of tube (secondary side) was experienced
due to residual alkali which came to be concentrated at
the clevis between tube and tube plate. To prevent thisproblem, a method of expanding the tube inner diameter along the entire length embedded in the tube plateby using a roller was adopted for eliminating the clevis.
Associated with this problem, tube expansion by roller
brought high residual stress inside the tube and stresscorrosion crack occurred inside the tube (in the reactor
coolant side). For resolving the problem, MHI determined
to use the hydraulic pressure expansion method to reFig
strain the residual stress (Fig
Fig.. 2 ).
2.2 Establishment and expansion of the improved
technology
In the second generation plants, many improvements
were reflected based upon the experience obtained in thefirst generation. In addition, the modifications of inspection and maintenance procedures through the adoptionof integrated reactor vessel head structures, the increaseof space for maintenance and inspection and automatedmaintenance/inspection works resulted the reduction ofoccupational radiation exposure and shortening the periodical inspection duration. Equipments were also

Secondary side water quality, PO4

AVT treatment

Increase in number of support points by anti-vibration bar
Modification of holes intube support plate

BEC hole(Quadruple hole)Round hole Reactor
coolant
Tube support
plate

Secondarycoolant

Secondarycoolant

Tube
hole

Reactorcoolant

Tube
support
plate
Tube

Tube

Tube plate
Reduction in stress, Increased tube for fluidpressure + Increased tube for one-step roller
Fig. 2 Improvement of steam generator

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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Exposed dose
equivalent

60
50

2

40
30

1

Experience of initial trouble

20
10
0
70

0
75

80

1st generation PWRs
. imported plants anddomestically supplied
plants based on
foreign technology. Localization of PWR
Technology

85

90

95

00

2nd generation PWRs

. Improvement and development of PWR technologies. Increase in reliability, safety, and operation rate
. Reduction in exposure and shorteningof periodic inspection interval

20 000
15 000

3rd generation PWRs
. Improved operability, reliability,
safety and economy

70

75

80

85
90
95
Year
Fig. 3 Trend of operation rate and exposed dose to employees

Plant construction cost

simplified through the adoption of a super-sized moisture separator and heater with increased unit capacity.
Plant reliability was also increased by application of an
integrated low-pressure turbine rotor. Along with the reflection of utility's excellent maintenance and operating
technique, they contributed to an increase of the plantFigavailability remarkably (Fig
Fig.. 3 ). In the third generation
plants that followed the second generation plants, thematured Japanese version PWR plants realized further
improvements in operability, economic efficiency andplant performance through the application of advanced
technology such as digital controllers and advanced main
control board, and optimization of the original system/
equipment designs and plant layout designs whichbrought the simplification of systems and equipments
and a reduction of equipment installation spaces in thebuildings.

2-loop600 MWe
class

10 000
5 000

Total installed capacity of PWRs
in Japan (MWe)Availability factor (%)

70

equivalent (mSv)

3

Availability factor

80

Average exposed dose90

0

Scale-up of major
components
3-loop900 MWe
class

4-loop1 200 MWeclass

Increase in numberof loop modules

APWR
1 500 MWe
class

Electric output
Fig. 4 Reactor model strategy

ogy and streamlined construction methods in responseto the utilities' needs. Tsuruga Unit 3 and 4, which aresupposed to be the first APWR units developed as thethird advanced standard design during the period from1981 to 1985 under the intimate cooperation of government, utilities and manufacturers. Since 1986, various
studies have been voluntarily continued for further advancement by the utilities and MHI. The aim of theplants, which provide the largest output in the world, isto realize significant increases in reliability, safety, operability and economic efficiency through introducing thesophisticated technology into the equipment and systems.
Although an APWR is of four loop plant, same as
the conventional large PWR plant, power output canbe increased to approximately 1 530 MWe by scale-up
Figof major components (Fig
Fig.. 4 ). The prime specificationsof the APWR and conventional four loop plant are com-

3. Today-Assuring the role of nuclear energy as abasic electrical power source
electrical
3.1 Development of APWR (Advanced PWR)
Today, from the viewpoint of assuring competitivenesswith other electrical power sources, such as combined
cycle thermal power plants, light water reactor plants
must be able to provide with high performance, facilitated operation and maintenance, and increased
economic efficiency as well as assured safety. In this regard, MHI has positively adopted numerous advanced
technology in the design of the plants, advanced design
techniques, such as three dimensional CAD methodol-

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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Table 2 Main specifications of APWR plant
Item

Current 4-loop

APWR

1 180
3 420

1 530
4 450

(m)
(t)

193
17 x 17
3.66
89

257
17 x 173.66
121(m3/h /loops)(kg/cm2a)42.01 x 104
157

42.58 x 104
157

52F type70F-1 type4

4

Reactor coolant system

ReactorElectric output(MWe)
NSSS power reactor system thermal output (MWe)
Number of fuel assemblies
Fuel rod lattice
Active core lengthCore loading (uranium weight)
Number of loops
Reactor coolant flow
Reactor coolant pressure
Steamgenerator

Type
NumberSteam pressure (kg/cm2a)
Type

ReactorNumbercoolant pump
Motor output

(kW)

Type
TurbineGenerator

62.5

62.5

93A-1 type
4

100A type

Approx. 4 480

Approx. 6 000

TC6F44 typeTC6F54 type

4

Moisture separator & reheater 2-stage reheat 2-stage reheat
Capacity

1 310

(MVA)

1 715

Upper radialreflector alignment pin
Tie rod
Alignment pin
Flow holes

Bolt

Tie rod
Fuel assembly

Tie rod

Core barrel
Radialreflector
ReactorvesselFlow hole

Ring block
Radial reflector

Core baffle

Fig. 5 Radial reflector

pared in T abl e 2 . Major design features of APWR
mainly consisting of advanced technology are presented below:(1) Improvements in reliability
In order to increase the reliability of the APWR
while assuring maintainability, improvements weremade to the major components such as the coreinternals and the steam generators.
With regard to the core internals, instead of thecore baffle formed with stainless steel plates assembled with a large number of bolts, it was decided
to use the neutron reflector which is composed witheight stainless steel ring blocks stacked around thereactor core. Adoption of the neutron reflector led
to a reduction of bolts used for assembling fromsome 2000 to about 50 and eliminated the weldingFiglines (Fig
Fig.. 5 ). In this way, the core internals were

simplified and the neutron dose to the reactor vessel could be reduced to approximately one-third,
thus the reliability of the reactor vessel was further improved.
A large-sized steam generator (Model 70F-1) was
adopted in which the heat transfer area was increased
from 5 000m2 (conventional 4 loop plant) to approximately 6 500m 2 in accordance to reactor powerincreased. The diameter of the heat transfer tube, on
the other hand, was decreased from the conventional
7/8 in. to 3/4 in. in order to suppress any increase of
the shell. This is an example of applications of advanced technology to limit the increase of theequipment size followed with equipment capacity increase. The height of steam generator was limited byadopting the upgraded secondary separator with reduced numbers of separation stages. Through these
Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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Steam outlet nozzle
Upper head

.
.
.

1

Single tier secondary separator
Reduction of SG (height)

Upper shell
Feedwater inlet nozzle

2 Compact moisture separator
Improvement of water/steam separating performance

Transition cone

3 Improved anti-vibration bars
Enhancement of allowance to flow induced vibration of tubes

Middle shell* Increase the number of points (from 6 to 9 points ) supportedby the anti-vibration bars
* All tubes are supported by the central anti-vibration bars

Lower shell
Tube sheet
Channel head.

4 3/4 inch tube
Minimization of the increase of the steam generator
size which copes with increased plant output

Reactor coolant
inlet/outlet nozzle

Fig.6 70F-1 type steam generator

Current 4-loop plant(2 divisions)
CS/RHRP

ACC

Reactor
ACC

SG

SG

ACC

Sump
LP HP
CSP

RWSP
SG

ACC
ACC

SG

Reactor
SG

SG

SIP

SIP

Spray header

SG

CS/RHRP

ACC
ACC

Spray header

SG

ACC

SIP: Safety injection pump
HP: High head safety injection pump
LP: Low head safety injection pump
CSP: Containment spray pump
CS/RHRP: Containment spray/
residual heatremoval pump
SG: Steam generator
ACC: Accumulator tank/
Advanced accumulator tank
RWSP: Refueling water storage pit

Spray header

Spray header

Spray header

Spray header

APWR plant(4 divisions)
SIP
SIP

HP LP
CSP
CS/RHRP
RWSP

CS/RHRP

Fig. 7 Emergency core cooling systemFig
was reduced (Fig
Fig.. 7 ).
In addition, the refueling storage pit previously installed outside containment was moved to the bottom
floor inside containment to use as the water source
for the emergency core cooling systems during loss of
coolant accident (LOCA). Thus, the coolant injected
into the core during accident can be collected in thispit for using as a water source again, and switching
operation of water source can be eliminated for enhancing the safety.
High performance accumulator tanks were applied
as an advanced technology to the emergency corecooling system. This accumulator tank is equippedwith a device which can switch, without any external power, between large injection flow required in
the early stage of LOCA and low flow in the latterstage. As a result, the conventional low pressure injection pump for low flow injection could be

modifications, the weight of steam generator was reduced by approximately 10% or more compared withconventional one with same capacity. In addition, thenumbers of anti-vibration bars installed at the U-bend
portion of the heat transfer tube bundle were increased, for improving the reliability, in order toincrease the tube supporting points from 6 for the latFig
est plants to 9 for APWR (Fig
Fig.. 6 ).
(2) Enhancement in safety
For enhancing the safety features of APWR, thedesign of the emergency core cooling system was
modified from the conventional 2 train configuration (2 x 100% capacity) to 4 train configuration(4 x 50% capacity) to increase the reliability ofequipment operation during accident. By installing each channel near the associated loop,
separation and independence of the safety-related
equipments were reinforced and the pipe volume

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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Current 4-loop plant
Long termcore cooling

Conventional
accumulator tank
Request for
injectionLong termcore cooling

Blow downCore refilling
and refill

Advanced accumulator tank

Low-head
injection pump

Injection rate

Blow down
and refill Core refilling

Injection rate

APWR plant

High-headinjection pumpTime

Request for
injection
Safety injection pumpTime

Fig. 9 Characteristics of core injection at loss-of-coolant accident (LOCA)
Nitrogen High water level
gas
Injected Large flow bywater smooth flowing

Vortex damper

Nitrogengas

Low water level

Small flow by
Injected
occurrence of swirl
water

Vortex damper

During large flow

Fig. 10 Operator console, large display panel
and shift supervisor console

During small flow
Switching of flow

Elimination of low pressure injection pump
Fig. 8 Mechanism of advanced accumulator

Figeliminated (Fig
Fig.. 8 and 9 ).
(3) Improvement of plant operability
The advanced main control board was designed toincorporate the operation system utilizing the compact consoles equipped with flat and touch-typecomputer displays. Additionally system monitoringand operations are integrated with each other by collectively displaying the monitoring information on theoperation screen to allow easier operational control.
The digital control/protection systems were adopted
to facilitate the maintenance works using the self-diagnosis and auto-test equipments, as well as theadvanced annunciator system which can identify thecriticality and priority of the alarms was introduced
to support the operator's activities. By introductionof those improved equipments and systems, the plantcan be maneuvered by just one operator during normal and accident conditions. A large display has beenplaced in the main control room to indicate the status and parameters of the major systems so that all
personnel, operators and shift supervisor, can easily
identify the status of the entire plant for monitoringand operation. These design improvements will bringsignificant contribution to enhancement of the plantFig
operability (Fig
Fig.. 10
10).

Fig. 11 Final-stage 54 inch blade of lowpressure turbine(4) Large capacity, high performance turbine
Since an advancement of turbine performance willcontribute greatly to higher plant economic efficiency,
the steam turbine used for APWR was designed toattain the higher efficiency and increase the capacity
in compliance with the reactor power increased as wellas the reliability of large rotary machine by introducing the advanced technology. High performance blades
designed using the complete three dimensional flow
analyses were adopted for higher efficiency. A groupof 54 in. blades was adopted as the final stage blades
Figof the low pressure turbine (Fig
Fig.. 11 ) so as to maximize the output under large steam flow conditions due
to increased capacity. Accompanying to adoption of alarge sized low pressure turbine, the bearings which
Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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Corrosion-resistantadvanced claddingHigh density pellet
Zircaloy grid

Debris filter

Fig. 12 Main features of step 2 fuel
(55 GWd/t fuel)Fig. 13 Steam generator replacement

were originally supported from the turbine casing
were changed to be supported directly on the base inorder to increase the supporting rigidity of bearings
so as to enhance the reliability against vibration.
3.2 Improvement and development of nuclear fuelSince MHI manufactured nuclear fuel assemblies for
Mihama Unit 1 in 1970, MHI has manufactured approximately 15 000 PWR fuel assemblies as of Dec. 2002.
During the last eleven years, the fuel has had excellent
operating experiences without any leakage.
In order to reduce the fuel cycle cost, the burn-up of
the fuel was increased from 39 GWd/t to 48 GWd/t (step1) following extension of core cycle length. At the sametime, various design improvements were made with theaim of increasing reliability. At present time, the development of 55 GWd/t fuel has been completed as step-2
fuel, and the safety reviews by the utilities are on theway for commercial use.
The distinctive features of the step-2 fuel are use of
corrosion-resistive cladding to meet the higher burn-up
and high-density pellet to increase the amount of uranium loaded for further reduction of the numbers of thereplaced fuel assemblies. Zircaloy, which has lower neutron absorption characteristics, was adopted as gridmaterial in the fuel assembly design in place of the conventional Inconel. As a result, the reactivity of the corecan be maintained longer compared to the Inconel usedfuel with the same uranium enrichment, thus operation
cycle duration can be extended. In order to reduce fuelleakage due to foreign materials (debris) in the primarycoolant, the design of fuel was changed to apply the newly
developed debris filter in the lower section of the fuelFig
(Fig
Fig.. 12
12).
MHI is striving for development of high burn-up fuelaiming further reduction of fuel cycle cost and increaseof the reliability.
The design developments of Mixed Oxide (MOX) fuelhave already been completed in line of the national recycle policy, and the MOX fuel is now ready for operation

Fig. 14 Reactor vessel headreplacement(before shipment)in the core with the burn-up up to 45 GWd/t. On theother hand, careful investigation of the fuel manufacturing plant and coordination of the specifications
required for the quality control to assure that the MOX
fuel has a quality equivalent to that of domestic uranium have been made for the case when MOX fuel may
be manufactured in the large-scale MOX manufacturing
plant in France and/or in United Kingdom. At the sametime, preparatory works for licensing the fuel manufacturing plant are under progress from the standpoints of
the quality assurance.
3.3 Activities for stable operation of existing plants
Commercial operations of Mihama Unit 1 (PWR) and
Tsuruga Unit 1 (BWR) commenced in 1970. There has
been an increase of the numbers of aged plants whichhave been operated for more than twenty years. In order to reduce the power generation cost through theirlives, it is important to operate these plants by using
the existing equipments as effectively as possible while
maintaining the stable operation. In order to meet thoserequirements, extension of the operating period and increase of generated power in the possible range areconsidered. A brief summary of the technical developments for these purposes is presented below:(1) Measures for the aged plants
In order to extend the plant life maintaining safety
and reliability, preventive maintenance must be carried out regularly and thoroughly so as to prevent theoccurrence of failures or troubles. From the viewpointof preventive maintenance, MHI has replaced thesteam generators and the reactor vessel upper heads
based upon the experience in oversea plants, and alsoplans to replace the core internals of aged plants inFig
the future (Fig
Fig.. 13 and 14
14).
Also MHI is actively addressing the developmentof inspection technology to detect any indications as
early as possible that could lead to failures or troubles.
Notable examples include the intelligent Eddy Current Test (ECT) equipment that allows accurate and
Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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Containment vessel
Pressurizer

Main steam

SG

ReactorvesselIncrease in performance of SG
and increase in flow to RCP
Change to highperformance rotor

Main feed water
TurbineSupplywater
flowmeter
FI

Generator

Condenser

Phased array
UT machine

ReactorFig. 15 Application of phased array
UT into reactor vessel

Sea water

RCP

Appropriation of safety
allowance of core

Circulating water pump
Feed water pump
SG: Steam Generator
Increase in measurement
RCP: Reactor Coolant
accuracy for supply water flow
Pump

Fig. 16 Method for increasing electrical output

high speed inspection of steam generator heat transfer tubes and phased array Ultrasonic Test (UT)Figequipment (Fig
Fig.. 1515) that is capable of shortening flaw
detection time for equipments such as the reactor vessel and reducing the undetectable region. The studies
are currently underway on the feasibility of practicaluse of these equipments.
In addition, MHI is positively addressing the development of evaluation methods for the occurrence
and progress of stress corrosion cracks in Ni-based
alloys in the primary coolant and advanced crack
growth analysis methods that can take into accountthe residual stresses. Each plant is being evaluatedin turn and confirmed the operability for sixty years
without any problem through the adoption of suitablemeasures including the development and applicationof the techniques mentioned above.
(2) Long cycle operation
The longest operating cycle permitted in Japan
until now is thirteen months, however the review ofextension of the operating cycle has started and is
expected to realize in the near future. In the U.S.,
most nuclear power plants have already been approved to operate for eighteen months or more andare now implementing. In Japan, specific studies are
now underway for extension of the operating cycle asan effective measure to increase the plant availability.
To realize long cycle operation, the plant must be
evaluated and verified from the viewpoints of the fuelperformance, the reliability of plant facilities and theoverall plant safety. With regard to the fuel performances, the development of the high burn-up fuel has
already been completed, and every effort toward practical use is now promoted. The evaluation of the effects
on the reactor core design and the safety analysis,
and the countermeasures to the evaluation results are

now under consideration. As to the reliability of facilities, the integrity evaluation of equipments, which
are maintained during periodical inspection, has been
performed, partly under way, and verified the fundamental applicability to long cycle operation. In orderto evaluate the effects of long cycle operation on plantsafety, the incremental of the risk has been assessed
through the application of probabilistic safety analysis (PRA) approach, and various measures for
improving the facilities and operation are now underinvestigation.
(3)Increase of electrical output
For operating plants, the license amendment forincreasing the electrical output by making changesof the plant equipment or optimization of the safetymargin are frequently applied aiming at the improvement of economical efficiency in the U.S. and Western
Europe. As an engineering measure to increase theelectrical output without changing the reactor thermal output, consideration can be given to improving
the thermal efficiency by replacement of the existing
equipments, such as replacement with large capacity
and high performance steam generators and high efficiency turbines. Equipment replacement works, suchas change to advanced turbine, are already performed
in Japan for better performances. Also, increasing thethermal output of the reactor by optimizing the design margin and/or equipment margin necessary for
maintaining the integrity can be effective engineering measures. MHI is proceeding with technicalstudies into those measures, including the verification and evaluation of the safety and integrity of theplants, referring to the experience in the U.S. andWestern Europe. The increase of electrical output of
15 to 20% at most, depending on the number of loops,
is expected through the combination of the above meaFig
sures taken (Fig
Fig.. 16
16).
Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

9and apply nuclear energy to various uses other than power
generation. Six concepts were selected at the forum: GasCooled Fast Reactor, Lead-Cooled Fast Reactor, Molten Salt
Reactor, Sodium-Cooled Fast Reactor, Supercritical WaterCooled Reactor and Very High- Temperature Reactor. In
Japan, innovative reactor designs were also studied by the
Atomic Energy Commission under the Long-term Nuclear
Energy Plan of 2000.
Taking these efforts in Japan and overseas into account,
MHI is actively working to develop the new APWR+ plant
concept to meet the future needs for large capacity power
production and the small integrated module reactor IMR(Integrated Modular Water Reactor) which can respondflexibly to power demands. MHI has also participated in a
number of development projects and initiatives in cooperation with other major power producers around the world.
These include the simplified AP1000 plant being developed
under the initiatives of Westinghouse, the high temperature He gas-cooled small reactor PBMR (Pebble Bed
Modular Reactor) being developed by PBMR Ltd. of South
Africa and the high temperature gas-cooled reactor capableof being available for multi purposes such as hydrogen production. As overview of the basic philosophy for realizing
an increase of economic efficiency for each concept is schematically shown in Fig
Fig.. 1 7 . A summary of the concept ofplant are also presented below:(1) APWR+
The reconstruction of the existing nuclear power
plants will be demanded in the near future. MHI basically believes that large capacity plants can respond tothese demands from the viewpoints of utilizing the construction site more effectively and reducing the overall
construction costs. MHI is developing APWR+ as a promising concept for accommodating these needs. APWR+
is a 4 loop plant designed based upon the APWR in whichelectrical output can be increased to approximately 1 750
MWe by incorporating the longer fuel and increased
capacity of major equipments. The safety systems aswell as the electrical systems are of four channel configuration so that maintenance at power can be
performed. In this concept, higher economic efficiency
is attained by enhancing the safety and pursuing themerits of large scale production.
(2) Small Integrated Modular Reactor: IMR
If a nuclear power plant suitable for small distributed
power source which is located in close proximity to theservice area and requires no provision of the specific installations such as transmission lines, etc. and can becompatible with both the safety and the economy could be
realized, a new nuclear energy market could be developedin the future. And if such a reactor could be of modular
type, it could be able to respond flexibly to the electrical
power demands and the site conditions. From the above
viewpoint, MHI is developing the small integrated moduFig
lar reactor (IMR) as a concept to meet these needs (Fig
Fig.. 18
18).

4. Tomorrow-Aiming at long-term stable supply ofenergyFrom the viewpoint of finding suitable solutions to environmental problems and attaining independence and longterm security of energy supply, nuclear power generation
is believed to continue to play an important role as ever inJapan which has virtually poor natural energy resources.
Accordingly, light water reactor plants will be required
to increase economic efficiency while assuring sufficient
safety through the introduction of new concepts. In addition to light water reactors, it is expected that medium or
small size reactors will come to be realized as a locally distributed power source which would allow the risk of new
installation to be reduced. There will also be increased
demands for multi-purpose reactors in the future that operate at high temperature. In addition, there will also be aneed to develop the nuclear fuel cycle based on the fast
breeder reactor in order to enhance the long-term utilization of uranium resources.
These trends have already begun. At the generation IV
international forum (GIF) on the generation IV nuclear systems held under the initiatives of the U.S. and participatedby many countries including Japan, numerous proposals havebeen presented to realize the introduction of new types ofinnovative reactors by 2030 as a target date. Proposals contained the new type of reactors which would increase economic
efficiency and safety, reduce the radioactive waste, prevent
nuclear proliferation, use uranium in ever more effective ways,
Applied new technologies
Plant construction price

Increase in thermal efficiency
by He gas coolingIntegration of primary system
Adoption of statically safe system

AP600 Scale
merit

PBMR

IMR

500

AP 1 000

APWR
APWR+

1 500
2 000
1 000
Electric output (MWe)Fig. 17 Concept of increase in economic efficiency
of future reactor

Built-in steam
generator

ReactorFig. 18 Integrated modular water reactor (IMR)

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Technical Review Vol.40 No.1 (Feb. 2003)

10

The IMR is a plant with electrical output of 300 MWe in
which the steam generator is integrated into the reactor vessel and the reactor is cooled by natural circulationof the primary coolant. Anticipated accidents can be limited by eliminating the primary coolant pipings. It willbring the enhanced safety as well as the improved
economy. This reactor concept and compatibility withthe safety are now under verification in the public research sponsored by the Ministry of Economy, Trade andIndustry (METI). Since a small reactor thus standardized can be completely prefabricated in a factory andshipped by barge, a remarkable reduction of the construction duration and improvement of the economic
efficiency can be expected to attain in combination with
offshore siting. In addition, if the development, construction and operation of this reactor could be performed incooperation with other countries such as Southeast Asia,
it would play a role in terms of the international contributions by Japan.
(3) AP 1000
It is expected that the plant economies can be improved by using the simplified safety systems whichuses the passive equipments operated by gravity or
water head instead of the conventional safety systems
driven by the dynamic forces. Westinghouse developed
AP 600 plant based on this concept with electrical
output of 600 MWe and acquired the Design Certification by the U.S. Nuclear Regulatory Commission(NRC) in 1999. Westinghouse is also currently developing AP 1000 plant with increased electrical output
Figof 1 000 MWe (Fig
Fig.. 19
19). MHI is now co-developing thedesign with Westinghouse aiming at acquiring design
certification for AP 1000 concept for a medium capacity plant of high safety and economic efficiency. These
plants are expected to be put in the practical use in
the near future.
(4) High temperature gas cooled reactor
PBMR Ltd. of South Africa has developed a small
reactor (PBMR) with output of 165 MWe. It is characterized by using the direct cycle gas turbine with highthermal efficiency (higher than 40%) that uses heliumgas heated to approximately 900 oC and by high levelof safety achieved in part by adoption of coated parFig
ticle fuel (Fig
Fig.. 2020). In the context of the advanced gas
turbine technology, this design is intended to achieve
high efficiency and compact nuclear power plant which
generates electrical power through the gas turbine
generator directly driven by high temperature He gas
heated in the reactor. MHI has participated in thedevelopment project of the world's first vertical helium gas driven turbine generator in response torequest from PBMR Ltd. combining with both nuclear
energy technology and gas turbine technology.
In addition, MHI has positively participated in thedevelopment of a high temperature gas cooled reac-

tor known as GTHTR 300 which is promoted by theJapan Atomic Energy Research Institute and the development of the hydrogen production systems coupledwith the high temperature engineering test reactor(HTTR).
(5) Fast Breeder Reactor (FBR)
The nuclear fuel cycle technology associated withthe fast breeder reactor can remarkably increase theutilization factor of uranium resources as comparedwith that of the light water reactor. It can also reduce the level of radioactivity that remains for long
period in the high level radioactive waste. Accordingly,
from the viewpoint of realizing a long-term stable andsafe supply of energy, it is urgently required to put
those technology into practical use as soon as reasonably possible.
In "Feasibility Study on Commercialized Fast Rector Cycle System" currently undertaken by Japan
Nuclear Cycle Development Institute, nine electricalutilities, Electric Power Development Co. Ltd, The
Japan Atomic Power Company and the associated organizations, the concepts of the fast breeder reactor
with economic efficiency comparable to the light water reactor are now under construction throughintroduction of the innovative technology. MHI is also

Fig. 19 AP1000 (conceptual drawing of
expected completed plant)

Reactor
High/lowcompressor
turbo unit

Generator

Power turbine

Recuperator

Reactorfuel outlet
Inter-cooler

Precooler
Fig. 20 Bird's eye view of PBMR

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Technical Review Vol.40 No.1 (Feb. 2003)

11

Fig
primary system equipments (Fig
Fig.. 2121). This concept has
also been nominated as a future prospective candidate model at the International Forum on Generation
IV Nuclear Energy Systems. It is hoped that the"Monju" will start operation again at an early date so
as to establish a base for practical use and the concept of practical fast breeder reactor will be verified
at plant level.

Steam generator
Secondary main pipe

Primary main pipe

Reactor structurePrimary integrated component(intermediate heat exchanger+ primary pump)

Fig. 21 Sodium-cooled, advanced loop type fast breeder reactor

actively engaged in the developments of this project.
Of the technology incorporating MHI's innovative
idea, the concept of the sodium cooled fast breeder
reactor, as a typical example, is presented below:
For the practical use of the fast breeder reactor, itis the most important point how it can approach to
the light water reactor, particularly with respect toeconomic efficiency. The sodium cooled advanced loop
type reactor has been selected as a candidate concept
for this requirement. Using the same type of sodium
as in prototype "Monju" reactor for cooling system,
this concept aims at significant reduction of the construction cost by introducing the novel ideas andtechnology such as compact reactor structures, minimized piping length by using the new materials,
elimination of the numbers of loops and integrated

5. Conclusion
A brief overview of the technical development activities performed by MHI, in order to assure the nuclear
energy as a basic power source and realize the long-term
stable energy supply in the future, has been presented
above. Half a century is going to pass since Japan decided to use the nuclear power for peaceful purposes. In
the mean time, Japan had experienced big changes indomestic and international economical, political and social environments. However, the starting point where
Japan, as a resource-poor country, had decided to introduce the nuclear power has remained unchanged. To
make this basic policy more secured, the role played by
the nuclear power plant manufacturers is supposed to
be important and extensive. Under the further cooperative guidance of the government, utilities and otherentities concerned, MHI will continue to make every effort to assure and reinforce its technical bases forsupporting the nuclear power plants, huge accumulationof various technology, and advance the new and challengeable technology toward the future.

Nuclear Energy Systems Headquarters
Managing Director,
General Manager,
Nuclear Energy Systems
Headquarters
Yoshio Miyake
Hiroshi Mukai

Mitsubishi Heavy Industries, Ltd.
Technical Review Vol.40 No.1 (Feb. 2003)

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