1
Technical Document
Graphic Viewing on FR Technology
・made by the data of
IAEA FAST REACTOR DATABESE 2006 Update・
April, 2015
International Atomic Energy Agency
FBR Senior Research Laboratory, Japan
Representative Tadao TAKAHASHI
The University of Tokyo, Japan
Professor Naoto KASAHARA
2
Preface
The graphs in this document are made by the data of ‘IAEA FAST REACTOR DATABESE 2006 Update.’
From this reason, ・All the data from the IAEA data book are treated having same technical worth. Total
number of plants adopted in the IAEA data book are not so many and also many data are
lacked, because the development of FR is now in progress.
・But, thermal power are written for the data of all plants, so, many graphs are made using
thermal power. These graphs show the effect of power size of plants, and also all data can
be shown certainly.
・Some graphs are made using specifications calculated from application of some basic
physical equations.
・The name of technical specifications written in this document are used those of IAEA
FAST REACTOR DATABESE 2006 Update and no modification are made.
・Almost of graphs are made by using colored marks. In these graphs, circular blue
symbols show the data of experimental reactors, cubic red symbols show
demonstration or prototype reactors and triangle green symbols commercial sized
reactors.
・Sizes of graphs, with combination of graph title, names of x-axis and y-axis, are decided
under consideration of the case of referring the graphs as state as themselves.
・This Document was originally made for educational purpose on FR plant system design
in the University of Tokyo, Japan.
3
Table of contents
0 FR plant (adopted in this document)・・・・・・・・・・・・・・・・・・・ 5
0.1 FR plants・・・・・・・・・・・・・・・・・・・・・・・・・ 5
0.2 Nominal full power of FR plants・・・・・・・・・・・・・・・ 6
1 General information of FR plants・・・・・・・・・・・・・・・・・・・・ 7
1.1 Classification of Plants・・・・・・・・・・・・・・・・・・・・7
1.2 Kind of fuel and coolant・・・・・・・・・・・・・・・・・・・ 7
1.3 Kinds of reactor(Primary circuit configuration)・・・・・・・・ 8
1.4 Development of plants・・・・・・・・・・・・・・・・・・・ 9
1.5 Development state in each nation or organization・・・・・・・ 11
2 Formation of reactor・・・・・・・・・・・・・・・・・・・・・・・・・・ 14
2.1 Fuel element・・・・・・・・・・・・・・・・・・・・・・・ 14
2.2 Fuel subassembly・・・・・・・・・・・・・・・・・・・・・ 17
2.3 Reactor・・・・・・・・・・・・・・・・・・・・・・・・・・ 19
3 Formation of cooling system・・・・・・・・・・・・・・・・・・・・・・・ 27
3.1 Cooling system・・・・・・・・・・・・・・・・・・・・・・・ 27
3.2 Component of cooling system (1. Circulation pump)・・・・・・ 31
3.3 Component of cooling system (2.Intermediate heat exchanger)・ 32
3.4 Component of cooling system (3.steam generator)・・・・・・・ 34
3.5 Steam turbine generator・・・・・・・・・・・・・・・・・・ 37
4 Form of other system and equipment・・・・・・・・・・・・・・・・・・・ 39
4.1 Refueling system etc.・・・・・・・・・・・・・・・・・・・・ 39
4.2 Secondary Containment・・・・・・・・・・・・・・・・・・・ 40
4.3 Coolant purification system・・・・・・・・・・・・・・・・・ 40
5 Nuclear characteristics・・・・・・・・・・・・・・・・・・・・・・・・・ 42
5.1 Volume ratio in the core・・・・・・・・・・・・・・・・・・・ 42
5.2 Neutron flux・・・・・・・・・・・・・・・・・・・・・・・・ 42
5.3 Linear power・・・・・・・・・・・・・・・・・・・・・・・・ 43
5.4 Power density・・・・・・・・・・・・・・・・・・・・・・・ 43
5.5 Enrichment of Plutonium・・・・・・・・・・・・・・・・・・ 44
5.6 Total breeding gain・・・・・・・・・・・・・・・・・・・・・ 44
5.7 Fuel burn up・・・・・・・・・・・・・・・・・・・・・・・・ 46
5.8 Reactivity and Doppler coefficient・・・・・・・・・・・・・・ 46
6 Cooling characteristics in the core・・・・・・・・・・・・・・・・・・・・ 48
6.1 Hydrodynamics of coolant in the core・・・・・・・・・・・・・ 48
6.2 Maximum coolant temperature・・・・・・・・・・・・・・・ 49
6.3 Maximum surface temperature of fuel cladding・・・・・・・・ 49
7 Plant cooling characteristics・・・・・・・・・・・・・・・・・・・・・・・ 452
7.1 Temperature of cooling system・・・・・・・・・・・・・・・・ 52
7.2 Temperature of intermediate heat exchanger・・・・・・・・・ 53
7.3 Temperature of steam generator・・・・・・・・・・・・・・・ 54
7.4 Plant thermal efficiency・・・・・・・・・・・・・・・・・・・ 55
8 Structural integrity(including material properties) ・・・・・・・・・・・・57
8.1 Fuel cladding tube・・・・・・・・・・・・・・・・・・・・・ 57
8.2 Fuel element spacer, cladding of blanket, wrapper tube・・・・ 59
8.3 Neutron absorber・・・・・・・・・・・・・・・・・・・・・・ 59
4
8.4 Reactor vessel・・・・・・・・・・・・・・・・・・・・・・・ 60
8.5 Primary piping・・・・・・・・・・・・・・・・・・・・・・・ 61
8.6 Primary main pump・・・・・・・・・・・・・・・・・・・・ 61
8.7 Intermediate heat exchanger・・・・・・・・・・・・・・・・ 62
8.8 Steam generator・・・・・・・・・・・・・・・・・・・・・・ 63
8.9 Secondary containment building・・・・・・・・・・・・・・・ 64
9 Safety・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 66
9.1 Plant shutdown・・・・・・・・・・・・・・・・・・・・・・ 66
9.2 Reactor scrum・・・・・・・・・・・・・・・・・・・・・・・ 67
9.3 Decay heat removal・・・・・・・・・・・・・・・・・・・・ 68
9.4 Detection of fuel failure and location of failed fuel・・・・・・・ 70
9.5 Detection of coolant leakage・・・・・・・・・・・・・・・・・ 71
9.6 Water leakage in steam generator・・・・・・・・・・・・・・ 71
10 Management of operation・・・・・・・・・・・・・・・・・・・・・・・ 73
10.1 Operation method・・・・・・・・・・・・・・・・・・・・ 73
10.2 Mean length of reactor run(Operation period)・・・・・・・・ 74
10.3 Exchange of fuel and others・・・・・・・・・・・・・・・・ 75
10.4 Preheating ・・・・・・・・・・・・・・・・・・・・・・・ 76
10.5 Purity of coolant・・・・・・・・・・・・・・・・・・・・・ 77
10.6 In-service inspection・・・・・・・・・・・・・・・・・・・・ 78
Literature・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 81
Index・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 82
5
0 FR plant (adopted in this document)
0.1 FR plant
0.1.1 Classification of FR plants
FR plants adopted in ’IAEA FAST REACTOR DATABESE 2006Update’ are shown
in the following table, and they are classified Experimental Reactors, Demonstration or
Prototype Reactors, and Commercial sized Reactors as follows.
Experimental Reactors Plant ReactorRapsodie RapsodieKNK-Ⅱ Kompakte Natriumgekuhlte kernreaktoranlageFBTR Fast Breeder Test ReactorPEC Prova Elementi di CombustibileJOYO JOYODFR Dournteay Fast ReactorBOR-60 Bystrij Opytnyj Reactor（Fast Experimental Reactor)EBR-Ⅱ Eｘperimental Breeder Reactor ⅡFermi FermiFFTF Fast Flux Test FacilityBR-10 Bystrij Reactor(Fast Reactor)CEFR China Experimental Fast Reactor
Demonstration or Prototype Fast Reactors Plant ReactorPhenix PhenixSNR-300 Schneller Natriumge kuhlte ReacorPBFR Prototype Fast Breeder ReactorMONJU MONJUPFR Prototype Fast ReactorCRBRP Clinch River Breeder Reactor PlantBN-350 Bystrie Neytrony(Fast Reactor）BN-600 Bystrie Neytrony（Fast Reactor)ALMR(Prism) Advanced Liquid Metal ReactorKALIMER-105 Korean Advanced Liquid Metal ReactorSVBR-75/100 Svinete-Vismus Bystrij Reactor（Lead-Bismuth Fast Reactor)BREST-OD-300 Bystrie Reactor Esteestvennoy Bezopasnosti （Fast Reactor Natural Safety)
Commercial Size Reactors Plant ReactorSPX-1 Super PheniX-1SPX-2 Super PheniX-2SNR 2 Schneller Natriumge kuhlte ReactorDFBR Demonstration Fast Breeder ReactorCDFR Commercial Demonstration Fast ReactorBN-1600 Bystrie Neuyrony（Fast Reactor ）BN-800 Bystrie Neuyrony（Fast Reactor ）EFR Europian Fast ReactorALMR Advanced Liquid Metal ReactorSVBR-75/100 Svinetc-vismuth Bvstiri Reactor（Lead-Bismuth Fast Reactor)BN-1800 Bystriij Neytrony（Fast Reactor ）Brest-1200 Bystrii Reacto Estesvennoy Bezopasnosti （Fast Reactor Natural Safety)JSFR-1500 JNC Sodium-cooled Fast Reactor
6
0.2 Nominal full power of FR plants
0.2.1 Power size of plants
Thermal power and electric power of each plant are shown in the following tables and
graphs.
0.2.2 Thermal power
0.2.3 Electric power
0
200
400
600
800
1000
1200
1400
1600
1800
Elec
tric
Po
wer
MW
e
Graph 0.2.3 Electric Nominal Full Power
Experimental FR Demonstration or Prototype FR Commercial Size FRPlant Country
etc.ElectricMWe
ThermalMWｔ
Plant Country etc.
ElectricMWe
ThermalMWｔ
Plant Country etc.
ElectricMWe
ThermalMWｔ
Rapsodie France 0 40 Phenix France 255 563 SPX-1 France 1242 2990KNK-Ⅱ Germany 20 58 SNR-300 Germany 327 762 SPX-2 France 1440 3600FBTR India 13 40 PBFR India 500 1250 SNR 2 Germany 1497 3420PEC Italy 0 120 MONJU Japan 280 714 DFBR Japan 660 1600JOYO Japan 0 140 PFR UK 250 670 CDFR UK 1500 3800DFR UK 15 60 CRBRP USA 380 975 BN-1600 USSR 1600 4200BOR-60 USSR 12 55 BN-350 USSR 130 750 BN-800 USSR 870 2100EBR-Ⅱ USA 20 62.5 BN-600 USSR 600 1470 EFR Euro 1580 3600Fermi USA 61 200 ALMR(Prism) USA 303 840 ALMR USA 303 840
FFTF USA 0 400 KALIMER-105
Korea 162 392 SVBR-75/100
Russia 101.6 280
BR-10 USSR 0 8 SVBR-75/100
Russia 80 265 BN-1800 Russia 1800 4000
CEFR China 23.4 65 BREST-OD-300
Russia 300 700 Brest-1200 Russi 1200 2800
JSFR-1500 Japan 1500 3530
0
500
1000
1500
2000
2500
3000
3500
4000
4500
Th
erm
al P
ow
erM
Wth
Graph 0.2.2 Thermal Nominal Full Power
7
1 General information of FR plants
1.1 Classification of plants
1.1.1 Size and classification of plants
Plant size of each plant is shown by using the value of thermal power or electric power.
In this document, value of thermal power or electric powers is used for indicating
actual size of plant in case by case.
1.1.2 Thermal power and plant classification
Graph 1.1.2 shows the relation between thermal power and plant classification.
In the graph, number of plant classification
indicates as follows.
The plant classifications (Experimental,
Demonstration or Prototype, and Commercial
sized FR) seem to depend on the value of
thermal power in general.
This means the step of development are
made consequently power upgrading of plants.
1.1.3 Electric power and plant classification
Graph 1.1.3 shows the relation between electric power and plant classification.
Number of plant classification used in graph
is same as in the previous graph.
This graph is similar to previous graph.
In these graphs, target powers of
commercialized plants are assumed to be 600
～1800MW electric, and power of
demonstration or prototype and experimental
plant were decided from these target power.
This relation can be confirmed later from
graphs shown the development step by
nations.
1.2 Kinds of fuel and coolant
Kind of fuel and coolant of FR have been investigated for many candidate materials and
decided respectively for each plant.
1.2.1 Kinds of fuel
Relation between starting time of plant construction and kind of fuels are shown in
graph 1.2.1.1. Kinds of fuel for thermal power are shown in Graph 1.2.1.2. Numbers of
kind of fuel are as follows.
In some smaller plants earlier constructed, uranium is used
naturally for fuel. But in the afterword stage,
plutonium-uranium mixed oxide fuels are mainly adopted.
In one plant, plutonium carbide fuel is used and this selection
3 Commercial Sized FR2 Demonstration or Prototype FR1 Experimental FR
1
2
3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Pla
nt
Cla
ssif
icat
ion
Thermal Power MW
Graph 1.1.2 Plant Classification for Electric Power
1
2
3
0 200 400 600 800 1000 1200 1400 1600 1800
Pla
nt
Cla
ssif
icat
ion
Electric Power MW
Graph 1.1.3 Plant Classification for Thermal Power
4 PuC-UC3 U-Mo, UN, UO2, U-Pu-Zr, 2 PuN-UN1 PuO2-UO2
8
has been evaluated as an attractive trial.
1.2.2 Kinds of coolant
The relation between starting times of plant construction and kinds of coolant in graph
1.2.2.1, and relation between thermal power and kinds of primary coolant are sown in
graph 1.2.2.2. Kinds of coolant for primary and
secondary are same in all reactors.
Numbers of kind of coolant are as follows.
Liquid metal sodium is used mainly as coolant of FR plant.
Sodium has good properties from the view point of nuclear and heat transfer
characteristics, but has inferior properties on strong chemical reaction against water,
and so, preventing method for chemical accident has been progressing.
Lead Bismuth alloy is one of back up candidate materials, but no plant has been
constructed yet.
1.3 Kind of reactor (Primary circuit configurations)
1.3.1 Size and type of reactor
Relations between type of reactor and thermal power are shown in graph 1.3.1.1,
electric power in graph 1.3.1.2. Numbers of type of reactor are as follows.
As conclusions, reactor type of small reactors are loop type, but on the
other hand, large reactors, except one reactor, are pool type because of expecting core
compactness.
Reactors type of plants having construction experience, are shown in graph 1.3.1.3.
In early stage of development, many reactors were loop type, but after 2000 year pool
type are mainly adopted.
1
2
3
4
1950 1960 1970 1980 1990 2000 2010
Kin
d o
f Fu
el
Start of Construction CY
Graph 1.2.1.1 Drive Fuel charge on Start of Construction
1
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Kin
d o
f Fu
el
Thermal Power MW
Graph1.2.1.2 Drive Fuel Charge
1
2
3
1950 1960 1970 1980 1990 2000 2010
1ry
Co
ola
nt
Start of Construction CY
Graph 1.2.2.1 1ry Coolant on Start of Construction
1
2
3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
1ry
Co
ola
nt
Thermal Power MW
Graph1.2.2.2 1ry Coolant
3 Lead-Bismuth2 Sodium-Potassium1 Sodium
9
This tendency may prove the development of FBR makes progress on possibility to
design the compact plant.
1.4 Development of plants
1.4.1 Classification of plant and its development
The relations between classification of plant and time of its first criticality are shown
in graph1.4.1. Classification of plant is shown as follows.
At the end of 1950’s experimental reactors
reached their criticality, and at early 1970’s
demonstration or prototype reactors gained
their criticality. And at the middle of 1980’s a
commercial sized reactors reached their
criticality. Their intervals are about 15 years
and this shows favorable progress of
development of FR plants.
But from the middle of 1990’s FR development was seemed to be standing. This
tendency can be seen from other graphs.
1.4.2 First criticality
The relations between the time of first criticality and thermal power of plants are shown
in graph1.4.2.
In this graph1.4.2, development of FR had been slowing down after 1985, like as
shown in graph 1.4.1. But after 2000 years, it seems to return to revival stage again.
1
2
0 500 1000 1500 2000 2500 3000 3500 4000 4500
1ry
Cir
suit
Co
nfi
gura
tio
n
Thermal Power MW
Graph 1.3.1.1 1ry Circuit Configuration
1
2
1950 1960 1970 1980 1990 2000 2010
1ry
Circ
uit C
onfig
urat
ion
Start of Construction CY
Graph 1.3.1.3 1ry Circuit Configuration
1
2
0 200 400 600 800 1000 1200 1400 1600 1800
1ry
Cir
cuit
Co
nfi
gura
tio
n
Electric Power MW
Graph 1.3.1.2 1rｙ Circuit Configuration
1
2
3
1950 1960 1970 1980 1990 2000 2010 2020
Pla
nt
Cla
ssif
icat
ion
First Criticality CY
Graph 1.4.1 First Criticality depend on Plant Classification
3 Commercial Sized FR2 Demonstration or
Prototype FR1 Experimental FR
10
1.4.3 First electric power generation and first full power operation
Relations between electric power and first electric power generation are shown in
graph1.4.3.1, and first full power operation in graph 1.4.3.2.
As shown in graph1.4.3.1, the development had been made progress till 1990 year, like
shown in graph 1.4.2. But after that, progress of development has been slowed down and
stayed. This phenomenon is clear especially for full power operation in graph 1.4.3.2.
From these information there is no plant reached the state of full power operation
except some experimental reactors after 1990.
1.4.4 Time of major plant events (start of construction, first criticality, first electricity
generation, first full power operation and final shutdown of plant)
Time of major plant events are shown in graph1.4.4. Number of major events is shown
as follows.
From middle of 1950’s to 1995 years, major
plant events, namely construction, criticality,
electric power generation and full power
operation, had been made progress
satisfactory. But after that, some plant had
stopped their operation and abolished after
short terms.
On the other hand, after 2000 year, some
commercial sized reactors started their
operation, but start of real restoration has not
been yet.
1.4.5 Accumulated development terms of each plant
Graph1.4.5 shows the operational history of each plant. Blue part marks operation
term from first criticality to shut down or in operation in 2003. So, the reactors shown
0
500
1000
1500
2000
2500
3000
3500
1950 1960 1970 1980 1990 2000 2010 2020
The
rmal
Po
wer
MW
First Criticality CY
Graph 1.4.2 First Criticality
0
200
400
600
800
1000
1200
1400
1950 1960 1970 1980 1990 2000 2010 2020
Ele
ctri
c P
ow
er
MW
First Electricity Generation CY
Graph 1.4.3.1 First Electricity Genaration
0
200
400
600
800
1000
1200
1400
1950 1960 1970 1980 1990 2000 2010 2020
Elec
tric
Po
wer
M
W
First Full Power Generation CY
Graph 1.4.3.2 First Full Powe Generation
1
2
3
4
5
1950 1960 1970 1980 1990 2000 2010 2020
Maj
or
Even
ts
Dates of Major Events CY
Graph1.4.4 Major Events
11
by black bar are in operation in 2003.
Value of reactor-years is the product of number of plant multiplied by its operation
years from first criticality.
Value indicated by blue area shows total reactor years till 2003 whether in operation or
not, so this value shows the actual operational result of each plant.
But black bar shows that the plant is in operation in 2003, and its value of reactor
years will be increasing by its operation.
Experimental reactors have suitable reactor years, but demonstration or prototype
reactors have not enough reactor years. Commercial sized reactors have very small
reactor years today.
1.4.6 Accumulated reactor years of FR
Values of accumulated reactor years of FRs are shown in graph 1.4.6. X axis of graph is
classified by plant operation history as follows.
Value of reactor years is used for evaluating
the development state or operation experience.
By this definition FR has about 400 reactor
years.
But this estimation of FR reactor years were
calculated by using IAEA data till 2003, and the calculated values are not accurate
because IAEA data are shown only the unit of calendar year for all events.
1.5 Development state in each nation or organization
Histories of FR Development are different for each nation or organization, because their
state of technical ability, economical capability and political state are different.
But in this document all data are treated without these considerations.
1.5.1 Development in each nation
Scales of plants, classified by nations or organization, are shown in graph 1.5.1.1 for
thermal power, in graph 1.5.1.2 for electric power. Numbers of classification are shown
as follows. But designs by 1France, 2Germany and 3England are succeeded by design by
10Euro.
1 First Criticality～Final Shutdown・2003 In Operation2 First Criticality～2003 In Operation3 First Electricity Generation～Final Shutdown4 First Criticality ～Final Shutdown5 Start of Construction～Final Shutdown
0
5
10
15
20
25
30
35
40
45
50
Term
ye
ars
Graph 1.4.5 Plant Operation Term
383
162128
221
261
0
50
100
150
200
250
300
350
400
450
1 2 3 4 5
Pla
nt
year
s
Clasification of Plant Operation Terms
Graph 1.4.6 Accumelated Reactor Years
12
In general, each nation has its special
development step from experimental reactors to
demonstration or prototype reactors, and also
commercial sized reactors. But some nations
have several designs for commercial sized
reactors under considerations of technical
competition or process for commercialization.
These graphs show 7Russian Federation,
10Euro and 5Japan as advanced nations for FR
development.
1.5.2 Development step in each nation
Relations between plant scale and development step of each nation are shown in
graph 1.5.2.1 for first criticality and in graph 1.5.2.2 for first power generation.
Developments from first criticality to first power generation were in favorable progress
until 1980’s, but slowed down after that.
1.5.3 Development step of each nation
Ratios of demonstration or prototype reactor
plant scale to experimental reactor, and that of
commercial sized reactor to demonstration or
prototype reactor plant scale are shown in
graph 1.5.3.
The ratios of demonstration or prototype
reactor plant scale to experimental reactor are
in the range of 11~14, and those of commercial
sized reactor to demonstration or prototype
reactor plant scale are 4~6.
1
2
3
4
5
6
7
8
9
10
11
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nat
ion
Thermal Power MW
Graph 1.5.1.1 Plant by Nation
1
2
3
4
5
6
7
8
9
10
11
0 200 400 600 800 1000 1200 1400 1600 1800
Nat
ion
Electric Power MW
Graph 1.5.1.2 Plant by Nation
0
2
4
6
8
10
12
14
16
18
20
France Germany Japan UK Russia USA
Rat
io o
f P
ow
er
Nation
Graph 1.5.3 Ratio of Plant Development
Demonstration or Prototype /Experimental
Commercial Size/Demonstration or Prototype
11 Republic of Korea 10 Euro 9 China8 USA7 Russian Federation, Kazakhstan 6 UK 5 Japan4 India3 Italy 2 Germany 1 France
0
500
1000
1500
2000
2500
3000
3500
1950 1960 1970 1980 1990 2000 2010 2020
Th
erm
al P
ow
er
MW
First Criticality CY
Graph 1.5.2.1 Step for Development of Plant
France
UK
Germany
Japan
India
Russia
USA
0
200
400
600
800
1000
1200
1400
1960 1965 1970 1975 1980 1985 1990 1995 2000
Ele
ctri
c P
ow
er
MW
First Electlicity Generation CY
Graph 1.5.2.2 Step for Development of Plant
France
UK
Germany
Japan
India
Russia
USA
13
For development of FR commercialization, scale up factor of plant power is about 5.
This technical view for scale up factor is common in almost all nations.
14
2 Formation of reactor
2.1 Fuel element
2.1.1 Foam of fuel pellet
There is no description about foam of fuel pellet in IAEA data book, but the diameters
of fuel cladding tubes are descripted for fuel specification, so foam of fuel can be
supposed as round bar. But this specification is specified only for cladding tube not for
fuel itself. Because the size of fuel pellet diameter or gap between pellets and claddings
are changed depending on their temperature or state of fuel burning, the diameters of
fuel pellets are not published. But using cladding diameter is convenience for predicting
fuel geometry.
2.1.2 Geometry of cladding tube
Diameters of fuel cladding tubes are shown in graph 2.1.2.1 for thermal power and in
graph 2.1.2.2 for electric power. Diameter of cladding seems to have a little relation to
plant scale.
Thicknesses of cladding tubes are shown in graph 2.1.2.3 for thermal power and in
graph 2.1.2.4 for electric power. Thickness of cladding seems to have a little relation to
plant scale.
The relation between diameter and thickness of cladding is shown in graph 2.1.2.5,
Moreover, ratios of thickness and diameter of cladding are shown in graph 2.1.2.6,
them the ratios are almost constant value.
These relations are described in the chapter of structural integrity of fuel cladding.
0
2
4
6
8
10
12
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Cla
dd
ing
Dia
met
er
mm
Thermal Power MW
Graph 2.1.2.1 Outer Diameter of Fuel Cladding
0
2
4
6
8
10
12
0 500 1000 1500 2000
Cla
dd
ing
Dia
met
er
mm
Electric Power MW
Graph 2.1.2.2 Outer Diameter of Fuel Cladding
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Cla
dd
ing
Thic
kne
ss m
m
Thermal Power MW
Graph 2.1.2.3 Thickness of Fuel Cladding
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0 500 1000 1500 2000
Cla
dd
ing
Thic
kne
ss m
m
Electric Power MW
Graph 2.1.2.4 Thickness of Fuel Cladding 2
15
2.1.3 Density of fuel
Intrinsic density of fuel pellets are shown in graph 2.1.3.1, and smeared density in
graph 2.1.3.2. The definition of smeared density is the density of fuel with fuel assumed
to occupy whole space inside the cladding tube.
2.1.4 Foam of fuel pellet
It is well known that there are two types of foam of fuel, namely solid and hollowed,
but no description in IAEA data book. Only some published documents describe
classifications about these foams. For this reason, no classification is given whether
solid or hollowed in this document.
The smeared density is used for calculating of physical properties in reactor physics.
But for calculating of heat generation and conductivity in fuel, using smeared density
makes error and calculation heat conductivity from fuel to coolant is impossible.
In this document, the smeared density is used for assuming diameter of pellet and for
understanding on physical and thermal properties of fuel.
2.1.5 Geometry of fuel pellet
Calculated diameters of fuel pellet are shown in graph 2.1.5.1, and the result
diameters of pellet are about 4 to 8 mm.
Ratios of diameter of pellet and inner diameter of cladding tubes are shown in graph
2.1.5.2. In this graph the ratio is higher value over 0.86, but data scatter in wide range.
Values of difference between 1.0 and this ratio relate to gap geometries for solid fuel,
and both gap and diameter of center hole for hollowed fuel, but it is impossible to
distinguish the foam of fuel from the value of this difference.
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
4 5 6 7 8 9 10 11
Cla
dd
ing
Thic
kne
ss m
m
Cladding Diameter mm
Graph 2.1.2.5 Thickness of Fuel Cladding
0.00
0.01
0.02
0.03
0.04
0.05
0.06
0.07
0.08
0.09
4 5 6 7 8 9 10 11
Rat
io o
f T
hic
kne
ss/D
iam
ete
r
Cladding Diameter mm
Graph 2.1.2.6 Ratio of Cladding Thickness/Diameter
80
82
84
86
88
90
92
94
96
98
100
0 500 1000 1500 2000 2500 3000 3500 4000 4500
De
nsi
ty o
f Fu
el
%T
D
Thermal Power MW
Graph 2.1.3.1 Intrinsic Density of Pellet
70
72
74
76
78
80
82
84
86
88
90
92
94
96
98
100
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Sme
are
d D
en
sity
of
Fue
l%
TD
Thermal Power MW
Graph 2.1.3.2 Smeared Density of Fuel
16
2.1.6 Length of fuel element
Lengths of fuel element are shown in graph 2.1.6.1. For large scale plant the lengths
are about 2.5 to 3.0 m.
The relation between height of fuel element
and these of core are shown in graph 2.1.6.2,
and the ratios of them in graph 2.1.6.3.
Lengths of fuel element reach about 2 to 3
times of height of core.
2.1.7 Fuel and blanket in fuel element
Lengths of fuel, upper and lower blanket in fuel element are shown in graph 2.1.7. The
data without blanket length in the graph means no prediction for blanket in IAEA data
0
500
1000
1500
2000
2500
3000
3500
4000
0 200 400 600 800 1000 1200 1400
Fue
l Ele
me
nt
Len
gth
mm
Core Height mm
Graph 2.1.6.2 Core Height--Fuel Element Length
0
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Ele
me
nt
Len
gth
/Co
re H
eig
ht
Thermal Power MW
Graph 2.1.6.3 Ratio of Fuel Element Length/Core Height
3
4
5
6
7
8
9
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Dia
met
er
of
Pe
llet
mm
Thermal Power MW
Graph 2.1.5.1 Diameter of Pellet(calculated using smeared density)
0.84
0.86
0.88
0.90
0.92
0.94
0.96
0.98
1.00
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Pe
llet/
Cla
dd
ing
Inn
er
Dia
met
er
Thermal Power MWGraph 2.1.5.2 Ratio of Pellet/Cladding Inner
Diameter
0
500
1000
1500
2000
2500
3000
3500
4000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Len
gth
of
Fue
l Ele
me
nt
mm
Thermal Power MW
Graph 2.1.6.1 Length of Fuel Element
0
200
400
600
800
1000
1200
1400
1600
1800
2000
40
58
40
12
0
14
0
60
55
62
.5
20
0
40
0 8
65
56
3
76
2
12
50
71
4
67
0
97
5
75
0
14
70
84
0
39
2
26
5
70
0
29
90
36
00
34
20
16
00
38
00
42
00
21
00
36
00
84
0
28
0
40
00
28
00
35
30
Len
gtu
mm
Thermal Power MW
Graph 2.1.7 Fuel/Blanket Length in Element
Upper Blanket LengthCore LengthLower Blanket Height
17
book or having no blanket.
2.1.8 Gas plenum
Relations between maximum fuel burn up and volume of gas plenum in fuel element
are shown in graph 2.1.8.1, and the relations between calculated length of gas plenum
and length of fuel element in graph 2.1.8.2. Gas plenum occupies enough long length in
fuel element, and their ratios attain about 1/3.
2.1.9 Length of fuel, blanket and gas plenum in fuel element
Ratios of length of fuel, blanket and gas plenum in fuel elements are shown in graph
2.1.9.
But no blanket and/or gas plenum in the graph means no prediction in IAEA data book
or no blanket and/or gas plenum.
2.2 Fuel subassembly
2.2.1 Structure of fuel subassembly
Fuel subassembly is foamed by hexagonal tube called wrapper tube, and contain
bundled fuel elements in it. Fuel elements are arranged in the foam of triangle structure
for high heat generation, so, the fuel subassembly is composed in the foam of hexagonal
shape. This hexagonal shape is profitable for arrangements in the core and
insert/pullout of fuel subassemblies in/from the core.
2.2.2 Type of spacer among fuel elements
For preventing direct contact among heated fuel elements, wire wrapped spacer
method or grid plate method is used.
The former is effective for coolant flow dynamics and the later for fabrication,
assembling of fuel elements bundle.
Types of spacer used in plants are shown in graph2.2.2. Classifications of type of
spacer are shown as follows.
Grid spacers are used in small reactors or a parts of large plants, but wire wrapped
spacers are used in many plants as typical spacer type for FR.
0
10
20
30
40
50
60
0 50,000 100,000 150,000 200,000 250,000
FP G
as V
olu
me
cm3
Maximum Fuel Buenup MWd/t
Graph 2.1.8.1 Fission Product Gas Volume per Pin
0
200
400
600
800
1,000
1,200
1,400
0 500 1,000 1,500 2,000 2,500 3,000 3,500 4,000
Gas
Ple
nu
mLe
ngt
hm
m
Fuel Element Length mm
Graph 2.1.8.2 Length of FP Gas Plenum
0.0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1.0
Rat
io o
f Le
ngt
h
Average Fuel Burnup MWd/t
Graph 2.1.9 fuel /Blanket /Gas PlenumGas Plenum Blanket Fuel
18
2.2.3 Length of fuel subassembly and fuel element, height of core
Lengths of fuel subassembly are shown in graph 2.2.3.1. It shows the length of fuel
subassembly are almost 4～5 m. So, the ratios
between length of fuel subassembly and fuel
element are shown in graph 2.2.3.2. The
ratio is 1.5~2.0 times.
This is because the fuel subassembly has
coolant entrance nozzle on down part and
handling head on upper part.
The ratios between the length of fuel
subassembly, fuel element and height of the
core are shown in graph 2.2.3.3.
Lengths of fuel element are 2～3 times
longer than height of core, and lengths of subassembly more than 4 times. These ratios
make affects for heightening coolant level in the core and longing height of reactor
vessel.
2.2.4 Number of fuel elements per subassembly
Number of fuel elements in a subassembly is decided by heat generation adapting
plant scale. Numbers of fuel elements are shown in graph 2.2.4. Their number increases
for scaling up of plant power, this means size of subassembly increases according to
plant power.
Because of triangle arrangement of fuel elements in hexagonal wrapper tube, number
of fuel element in subassembly are selected one of number series 1, 7, 19, 37, 61, 91, 127,
169, 217, 271, 331, 397, 489, 547, 631, 721, 817, 919, 1027, 1141・・・.
1
2
3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Typ
e o
f P
in S
epar
atio
n
Thermal Power MW
Graph 2.2.2 Type of Mechanical Separation of Pins
0
1000
2000
3000
4000
5000
6000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Len
gth
of
Sub
asse
mb
ly m
m
Thermal Power MW
Graph 2.2.3.1 Length of Subassembly
0.0
1.0
2.0
3.0
4.0
5.0
6.0
7.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Sub
asse
mb
ly/F
ue
l Ele
me
nt
Len
gth
Thermal Power MW
Graph 2.2.3.2 Ratio of Length Subassembly/Fuel Element
0
1
2
3
4
5
6
7
8
0 200 400 600 800 1000 1200 1400
Rat
io o
f Le
ngh
or
Hei
ght
Height of Core mm
graph 2.2.3.3 Ratio of Length, Height
Fuel Element/ Core
Subassembly/ Core
19
2.2.5 Wrapper tube of fuel subassembly
Relations between thermal power and width across flat of subassembly are shown in
graph 2.2.5.1, and widths across subassembly are mostly smaller than 200 mm.
This may be considered that width of
subassembly is equal or smaller than neutron mean free path.
The relations between number of fuel elements and width across subassembly are
shown in graph 2.2.5.2. This shows width across subassembly is proportional to
number of fuel elements in a subassembly and its diameter.
2.3 Reactor
In IAEA data book, the area containing only fuel subassemblies is called “core”. On the
other hand, the definition of “reactor” seems to be plant itself or plant name, but it is not
so clear.
In this document, the definition of “reactor” is the area including fuel, control and
blanket subassemblies, and moreover reflector etc..
2.3.1 Composition of reactor
Following the definition described above, reactor contains fuel, control, blanket and
reflector subassemblies, and also coolant and cover gas. Reactor vessel is the structure
mainly containing reactor.
2.3.2 Clearance between subassemblies
The relations between clearance and width across subassembly are shown in graph
2.3.2.1, and the ratios of them in graph 2.3.2.2. The value of the ratio are around 0.03,
so, the clearance is about 3 mm for width across subassembly100 mm, about 6 mm for
200 mm.
0
50
100
150
200
250
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Wid
th a
cro
ss F
lat
mm
Thermal Power MW
Graph 2.2.5.1 Width across Subassembly Flat 1
0
50
100
150
200
250
0 50 100 150 200 250 300 350
Wid
th a
cro
ss F
lat
mm
Number of Fuel Elements
Graph 2.2.5.2 Width across Subassembly Flat 2
0
50
100
150
200
250
300
350
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er
of
Fue
l Ele
me
nt
Thermal Power MWGraph 2.2.4 Number of Fuel Element per
Subassembly
20
2.3.3 Number of core zone
For flatting neutron flux distribution, core is divided a few zone containing fuels of
different Plutonium enrichments.
Numbers of core zone are shown in graph
2.3.3, and these numbers are1, 2 or 3 in
general.
For large power plant, core is divided mainly
in two zones.
2.3.4 Size of core
0
5
10
15
20
25
0 20 40 60 80 100 120 140 160 180 200Cle
aran
ce b
etw
een
Su
bas
sem
blie
sm
m
Width across Subassembly mm
Graph 2.3.2.1 Clearance between Subassemblies
0.00
0.02
0.04
0.06
0.08
0.10
0.12
0 20 40 60 80 100 120 140 160 180 200
Cle
aran
ce/W
idth
Width across Subassembly mm
Graph 2.3.2.2 Clearance/Width across Subassembly
1
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er o
f Zo
nes
Thermal Power MW
Graph 2.3.3 Number of Fuel Enrichment Zones
0
1000
2000
3000
4000
5000
6000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Dia
met
er
of
Co
rem
m
Thermal Power MW
Graph 2.3.4.1 Equivalent Diameter of Core
0
200
400
600
800
1000
1200
1400
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Hei
ght
of
Co
rem
m
Thermal Power MW
Graph 2.3.4.2 Height of Core
0.0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
He
igh
t/D
iam
ete
r
Thermal Power MW
Graph 2.3.4.3 Height/Diameter of Core
0
5
10
15
20
25
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Co
re V
olu
me
m3
Thermal Power MW
Graph 2.3.4.4 Volume of Core
21
Size of core, for example, the diameters of the core are shown in graph 2.3.4.1, the
heights of fissile zone in graph 2.3.4.2, the ratios of them in graph 2.3.4.3, and the
volumes in graph 2.3.4.4.
The diameters are proportional to the plant power, but the heights reach the limit 1.2
m. For large power plants, the ratios of height and diameter reach about 0.2, this shows
shape of core are flat circular cylindrical. From this reason the values of volume show
scattered values of diameter.
2.3.5 Region of inner and outer core zone
Many plants have the core of 2 zones, so, following investigations are focused in inner
and outer core region.
The diameters of inner core zone are shown in graph 2.3.5.1, and they show the inner
diameters are proportional to the power, like outer diameters shown in graph 2.3.4.1.
The ratio of inner and outer diameter are shown in graph2.3.5.2. These ratios are the
value of around 0.7～0.8 for large reactors, and so, cross sectional area of inner and
outer core zones are almost equal.
2.3.6 Number of fuel subassemblies
The numbers of fuel subassemblies in each zone are shown in graph 2.3.6.1.
For easy understanding, the ratios of them are shown in graph 2.3.6.2. Except special
plants, the numbers of fuel subassemblies in both zones are almost equal.
But for large plants, the ratios are larger than 1.0, this means that the numbers of
inner zone, having lower plutonium enrichment, are larger than those of outer zone.
2.3.7 Geometry of blanket
The outer diameters of blanket pins are shown in graph 2.3.7.
Except special case, the outer diameters of blanket pins seem to have a little
correlation with plant power like the diameter of fuel cladding tube.
0
500
1000
1500
2000
2500
3000
3500
4000
4500
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Dia
met
er
of
Inn
er
Co
rem
m
Thermal Power MW
Graph 2.3.5.1 Diameter of Inner Core
0.0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Inn
er/
Ou
ter
Co
re D
iam
eter
Thermal Power MW
Graph 2.3.5.2 Ratio of Inner/Outer Core Diameter
0
100
200
300
400
500
600
700
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er o
f Su
bas
sem
bie
s
Thermal Power MW
Garph 2.3.6.1 Number of Subassemblies
Inner Outer
0.0
0.5
1.0
1.5
2.0
2.5
3.0
3.5
4.0
4.5
5.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Rat
io In
ner
/Ou
ter
Thermal Power MW
Graph 2.3.6.2 Ratio of Number of Subassemblies Inner/Outer
22
2.3.8 Fuel and blanket
The relations between diameter of fuel and blanket are shown in graph 2.3.8.1, and the
ratios of them in graph 2.3.8.2. The ratios seems to be around 2.0.
Relations between number of fuel and blanket in graph 2.3.8.3. Both has mutual
proportional correlation, so the ratios of them are shown in graph 2.3.8.4 and they are
about 0.3～0.5.
2.3.9 Subassemblies in core
Number of fuel, blanket and reflector subassemblies are shown in graph 2.3.9.1.
This graph is complex, so the vertical axis is enlarged in graph 2.3.9.2.
It seems that order of values of number are as follows, reflector>fuel>blanket.
If numbers of subassemblies are expressed by the numbers of layer, these relation
seems to be easy to understand.
0
5
10
15
20
25
30
35
40
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Bla
nke
t D
iam
eter
mm
Thermal Power MW
Graph 2.3.7 Outer Diameter of Blanket Pin
0
2
4
6
8
10
12
14
16
18
20
0 2 4 6 8 10 12
Dia
met
er
of
Bla
nke
t m
m
Diameter of Fuel mm
Graph 2.3.8.1 Diameter of Fuel and Blanket
0.0
0.5
1.0
1.5
2.0
2.5
3.0
3.5
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Bla
nke
t/Fu
el D
iam
ete
r
Thermal Power MW
Graph 2.3.8.2 Ratio of Blanket/Fuel Diameter
0.0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Rat
ioo
Fuel
/Bla
nke
tN
um
ber
Thermal Power MW
Graph 2.3.8.4 Ratio of Fuel/Blanket Pins Number
0
50
100
150
200
250
0 50 100 150 200 250 300 350
Bla
nke
t P
ins
Nu
mb
er
Fuel Pins Number
Graph 2.3.8.3 Number of Pins of Fuel and Blanket
23
2.3.10 Weight of plutonium and uranium in the core
Total weights of plutonium in the core are shown in graph 2.3.10.1. Their weights are
very large for large plants.
Weights of plutonium-239 and uranium-235 are shown in graph 2.3.10.2. As the result,
the ratios of their weights are nearly equal to the ratios of their enrichment, namely
0.003/0.20=1/67.
2.3.11 The ratio of number of absorber element and diameter of element in control rod
subassembly.
Control rod subassemblies are divided in three types, coarse rod regulates rough
reactivity caused by fuel burn up, fine rod controls detailed reactor power level, and
safety rod actives rapid insertion in an emergency. Each reactor has various
compositions of them.
Diameter and numbers of elements in coarse, fine and safety rod subassembly are
shown in graph 2.3.11.1, 2.3.11.2 and 2.3.11.3 respectively.
0
200
400
600
800
1000
1200
1400
1600
1800
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er o
f su
bas
sem
blie
s
Thermal Power MW
Graph 2.3.9.1 Number of Fuel・Blanket・Reflector
Fuel
Blanket
Reflector
0
100
200
300
400
500
600
700
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er o
f Su
bas
sem
blie
s
Thermal Power MW
Graph 2.3.9.2 a part of Graph 2.3.9.1
Fuel
Blanket
Reflector
0
2,000
4,000
6,000
8,000
10,000
12,000
14,000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Tota
l Pu
Wei
ght
kg
Thermal Power MW
Graph 2.3.10.1 Total Pu Weight in Reactor
0
50
100
150
200
250
300
350
400
0 1000 2000 3000 4000 5000 6000 7000 8000
U2
35
Wei
ght
kg
Pu239 Weight kg
Grap 2.3.10.2 Weight of Pu239-U233 in Reactor
0
5
10
15
20
25
30
35
40
0 10 20 30 40 50 60 70 80 90
Dia
met
er o
f C
oar
se R
od
s m
m
Number of Coarse Rods Elements
Graph 2.3.11.1 Number of Coarse Rods Elements/Diameter
0
5
10
15
20
25
30
35
40
0 10 20 30 40 50 60 70
Dia
met
er
of
Fin
e R
od
s m
m
Number of Fine Rods Elements
Graph 2.3.11.2 Number of Fine Rods
Elements-Diameter
24
In general, rod elements of coarse and fine
are used same size but safety has much fatter
elements.
2.3.12 Number of control rod assemblies
From the view point of its effectiveness for neutron absorption, control rod assemblies
are located in the area of inner or inner/outer zone of the core. But there are no data for
their location, so total numbers of them are shown in graph 2.3.12.1.
Numbers of coarse, fine and safety rod
subassemblies are shown in graph 2.3.12.2
respectively.
Their total numbers increase according to
the thermal power, but scatter in wide range.
And the ratios of number of fuel and control
rod subassemblies are shown in graph
2.3.12.3. This graph shows one control rod
subassembly seems to shear control ability of
about ten fuel subassemblies. These
phenomena are enough supposed from the geometry and arrangement of fuel
subassemblies and length of mean neutron free path.
2.3.13 Composition of control rod subassemblies
Kinds and numbers of control rod subassemblies in each plant are shown on graph
2.3.13. It seems that some large plants have only safety control rod
subassemblies, but no more detail data is predicted in the IAEA data book.
0
20
40
60
80
100
120
0 10 20 30 40 50 60
Dia
met
er o
f Sa
fety
Ro
ds
ｍｍ
Number of Safety Rods Elements
Graph 2.3.11.3 Number of Safety Rods Element-Diameter
0
10
20
30
40
50
60
70
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er
of
Co
ntr
ol r
od
s
Thermal Power MW
Graph 2.3.12.1 Number of Control Rods Subassemblies
0
5
10
15
20
25
30
35
40
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er
of
Ro
ds
Thermal Power MW
Graph 2.3.12.2 Number of Control Rods
Safety Rods
Fine Rods
Coarse Rods
0
10
20
30
40
50
60
70
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Fuel
/Co
ntr
ol N
um
ber
Thermal Power MW
Grsph 2.3.12.3 Ratio of Fuel/Control Subassembies
25
2.3.14 Reactor vessel (Primary vessel)
Size of reactor vessel, especially its diameter, is depends on the type of reactor.
Inner diameters of reactor vessel are shown in graph 2.3.14.1 with the type of reactor.
Naturally diameters of reactor vessel of
pool type are considerably larger than those
of loop type.
The relations between diameter and height
of reactor vessel are shown in graph 2.3.14.2.
The heights are expected to be nearly
constant for large reactor vessel, so shapes of
reactor vessel of loop type are extremely flat.
Thicknesses of reactor vessels are shown in
graph 2.3.14.3. Thicknesses of reactor
vessels increase with increasing the
diameter for loop type, but thicknesses are nearly constant for pool type.
2.3.15 Kind of cover gas
Kinds of cover gas of primary cooling system are shown in graph 2.3.15.1, and
0
2,000
4,000
6,000
8,000
10,000
12,000
14,000
16,000
18,000
20,000
22,000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Inn
er D
iam
eter
of
Rea
cto
r V
esse
lm
m
Thermal Power MW
Graph 2.3.14.1 Inner Diameter of Reactor Vessel
Inner Diameter (Loop)
Inner Diameter (Pool)
0
2,000
4,000
6,000
8,000
10,000
12,000
14,000
16,000
18,000
20,000
22,000
0 2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18,000 20,000 22,000
Hei
ght
of
Rea
cto
r V
esse
lm
m
Inner Diameter Of Reactor Vessel mm
Graph 2.3.14.2 Diameter and Height of Reactor Vessel
Height (Loop)
Height (Pool)
Diameter＝Height
0
10
20
30
40
50
60
70
80
0 2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18,000 20,000 22,000
Thic
kne
ss o
f R
eac
tor
Ve
sse
lm
m
Inner Diameter of Reactor Vessel mm
Graph 2.3.14.3 Thickness of Reactor Vessel
Thickness(Loop)
Thickness(Pool)
1
2
0 500 1000 1500 2000 2500 3000 3500 4000 4500
1ry
Co
ver
Gas
Thermal Power MW
Graph 2.3.15.1 1ry Cover Gas
1
2
0 500 1000 1500 2000 2500 3000 3500 4000 4500
2ry
Co
ver
Gas
Thermal Power MW
Graph 2.3.15.2 2ry Cover Gas
0
10
20
30
40
50
60
70
40
58
40
12
0
14
0
60
55
63
20
0
40
0 8
65
56
3
76
2
1,2
50
71
4
67
0
97
5
75
0
1,4
70
84
0
39
2
26
5
70
0
2,9
90
3,6
00
3,4
20
1,6
00
3,8
00
4,2
00
2,1
00
3,6
00
84
0
28
0
4,0
00
2,8
00
3,5
30
Nu
mb
er
Co
ntr
ol R
od
s
Thermal Power MW
Coarse Rods
Fine Rods
Safety Rods
Graph 2.3.13 Kind and Number of Control Rods
26
secondary in graph 2.3.15.2.
Kinds of cover gas are 1 helium 2 argon. For primary, some plant use hellium, but all
of others use argon gas. Argon gases are used for secondary in all plants.
2.3.16 Pressure of cover gas
Pressures of cover gas in reactor vessel are shown in graph 2.3.16.1. The pressures
spread from slightly above atoms (shown pressure=0 in graph) to 0.2 MPa.
And pressures of primary and secondary cover gas are shown in graph 2.3.16.2.
Naturally gas pressures of secondary are about 1.5 times higher than these of primary.
0
0.05
0.1
0.15
0.2
0.25
0.3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
1ry
Co
ver
Gas
Pre
ssu
reM
Pa
Thermal Power MW
Graph 2.3.16.1 1ry Cover Gas Pressureslightly above atoms=0
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0 0.05 0.1 0.15 0.2 0.25 0.3
2ry
Co
ver
Gas
Pre
ssu
re M
Pa
1ry Cover Gas Pressure MPa
Graph 2.3.16.2 1ry-2ry Cover Gas Pressure
2ry-=1.5×1ry
2ry＝1ry
27
3 Formation of cooling system
3.1 Cooling system
3.1.1 Number of cooling system
Numbers of primary cooling system are shown in graph 3.1.1.1, and secondary in
graph 3.1.1.2.
The relations of number of primary and
secondary cooling system are shown in graph
3.1.1.3. Many plants have same number of
both cooling system.
3.1.2 Flow rate of coolant
Flow rates of primary coolant are shown in graph 3.1.2.1, secondary in graph 3.1.2.2.
The ratios of them are shown in graph 3.1.2.3. The ratios are about 0.8～1.0, this
indicates the flow rates of secondary are smaller than these of primary. These relations
are introduced from the ratios of primary and secondary hot/cold leg temperature
differences.
0
2,000
4,000
6,000
8,000
10,000
12,000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Flo
w R
ate
of
1ry
Co
ola
nt
kg/s
Thermal Power MW
Graph 3.1.2.1 Flow Rate of 1ry Cooloant
0
1,000
2,000
3,000
4,000
5,000
6,000
7,000
8,000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Flo
w R
ate
of
2ry
Co
ola
nt
kg/s
Thermal Power MW
Graph 3.1.2.2 Flow Rate of 2ry Coolant
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
1ry
Co
ola
nt
Loo
ps
Nu
mb
er
Thermal Power MW
Graph 3.1.1.1 Number of 1ry Coolant Loops
1
2
3
4
5
6
7
8
0 500 1000 1500 2000 2500 3000 3500 4000 4500
2ry
Co
ola
nt
Loo
ps
Nu
mb
er
Thermal Power MW
Graph 3.1.1.2 Number of 2ry Coolant Loops
0
1
2
3
4
5
6
7
8
0 1 2 3 4 5 6 7 8
2ry
Co
ola
nt
Loo
ps
Nu
mb
er
1ry Coolant Loops Number
Graph 3.1.1.3 Ratio of 2ry/1ry Coolant Loops Number
2ry＝1ry
28
3.1.3 Diameter of coolant piping
Diameters of piping of primary cooling system are shown in graph 3.1.3.1, secondary in
graph 3.1.3.2 and third (steam/water system) in graph 3.1.3.3. And all of these data
are shown together in graph 3.1.3.4.
In general, diameters of primary piping are larger than these of secondary piping, and
secondary are larger than those of third cooling system. The diameters of primary
piping are about 1000 mm, secondary about 800 mm and third about 500 mm. These
diameter sizes are related to each flow rate respectively.
3.1.4 Thickness of cooling piping
The diameter and thickness of primary hot leg cooling piping are shown in graph
3.1.4.1, secondary in graph 3.1.4.2 and third in graph 3.1.4.3.
Then, the ratios of thickness and diameter of hot leg cooling piping are shown in
graph3.1.4.4 for primary, graph 3.1.4.5 for secondary and graph 3.1.4.6 for third cooling
system. For primary and secondary, the ratios are approaching to a constant value for
large plants, but for third, these ratios are scattered in wide range.
0
200
400
600
800
1000
1200
1400
0 500 1000 1500 2000 2500 3000 3500 4000 4500
1ry
Pip
ing
Dia
met
er m
m
Thermal Power MW
Graph 3.1.3.1 1ry Hot Leg Cooｌant Piping Diameter
0
200
400
600
800
1000
1200
0 500 1000 1500 2000 2500 3000 3500 4000 4500
2ry
Pip
ing
Dia
met
er
mm
Thermal Power MW
Graph 3.1.3.2 2ry Hot Leg Cooｌant Piping Diameter
0
100
200
300
400
500
600
700
800
900
0 500 1000 1500 2000 2500 3000 3500 4000 4500
3rd
Pip
ing
Dia
met
er m
m
Thermal Power MW
Graph 3.1.3.3 3rd Hot Leg Cooｌant Piping Diameter
0
200
400
600
800
1000
1200
1400
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Dia
met
er o
f P
ipin
gm
m
Thermal Power MW
Graph 3.1.3.4 Diameter of Coolant Piping
1ry Hot Leg Cookant Piping Diameter
2ry Hot Leg Cookant Piping Diameter
3rd Hot Leg Cookant Piping Diameter mm
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
1.8
2.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
2ry
/1ry
Co
ola
nt
Flo
w R
ate
Thermal Power MW
Graph 3.1.2.3 Ratio of 2ry/1ry Coolant Flow Rate
29
3.1.5 Structure of piping
For structure of primary piping, double wall tube or single wall tube accompanied with
guard vessel are adopted. The former is
effective for receiving leaked coolant, the latter
is effective for keeping coolant level in the case
of coolant leakage accident. The piping
structures related to its piping diameter are
shown in graph 3.1.5.1. Single wall tubes
are used for large tube diameter.
1 Single 2 Double1
2
0 100 200 300 400 500 600 700 800 900 1000
1ry
Pip
ing
Stru
ctu
re
1ry Piping Diameter mm
Graph 3.1.5.1 1ry Piping Structure
0
2
4
6
8
10
12
14
16
18
20
0 200 400 600 800 1000 1200 1400
Thic
kne
ssm
m
1ry Piping Diameter mm
Graph 3.1.4.1 1ry Hot Leg Piping Thickness－Diameter
0
2
4
6
8
10
12
14
16
18
20
0 200 400 600 800 1000 1200 1400
Th
ickn
ess
mm
2ry Piping Diameter mm
Graph 3.1.4.2 2ry Hot Leg Piping Thickness－Diameter
0
20
40
60
80
100
120
0 100 200 300 400 500 600 700 800
Th
ickn
eaa
mm
3rd Piping Diameter mm
Graph 3.1.4.3 3rd Hot Leg Piping Thickness－Diameter
0.00
0.01
0.02
0.03
0.04
0.05
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Th
ickn
ess
/Dia
met
er
Thermal Power MW
Graph 3.1.4.4 1ry Piping Thickness/Diameter
0.00
0.01
0.02
0.03
0.04
0.05
0.06
0.07
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Th
ickn
ess
/Dia
met
er
Thermal Power MW
Graph 3.1.4.5 2ry Piping Thickness/Diameter
0.00
0.05
0.10
0.15
0.20
0.25
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Thic
knes
s/D
iam
eter
Thermal Power MW
Graph 3.1.4.6 3rd Piping Thickness/Diameter
30
The structures against coolant leakage for primary cooling
system are shown graph 3.1.5.2 and for secondary in graph 3.1.5.3.
Leak jacket structures are used for primary of many plants but
none for secondary of a lot of plants.
3.1.6 Valves in the cooling system
Some kinds of valves are installed in cooling system for corresponding to their effect
for coolant flow characteristics.
Kinds of valves are shown in graph 3.1.6, but hollow type maker shows no valve for
them.
For large plant, stop valves are hardly
installed in cold leg but installed in hot leg of
some plants. Check valves are little installed
for large plants. Steam generator isolation
valves are installed in almost all plants.
These decisions for installing valve are
selected from the view point of reactor type and safety.
3.1.7 Inventory of coolant
Coolant inventories contained in primary and secondary cooling system are shown in
graph 3.1.7.1, and the relation between primary and secondary in graph 3.1.7.2.
For large plant, coolant inventories reach amount of several thousand tons.
Inventories of primary system are larger than those of secondary in general.
0
1,000
2,000
3,000
4,000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Co
ola
nt
Inve
ntr
yto
n
Thermal Power MW
Graph 3.1.7.1 Coolant Inventry
1ry Ｉｎｖｅｎｔｒｙ
2ry Ｉｎｖｅｎｔｒｙ
1
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Val
ves
Thermal Power MW
Graph 3.1.6 Valving
4 1ry Cold Leg Stop Valve3 1ry Hot Leg Stop Valve2 !ry Check Valve1 SG Isolation
0
1
2
0 200 400 600 800 1000 1200 1400
1ry
Pip
ing
Sod
ium
Lea
k P
rote
ctio
n
1ry Piping Diameter mm
Graph 3.1.5.2 1ry Piping Sodium Leak Protection
2 Guard Vessel1 Leak Jacket0 none
0
1
2
0 200 400 600 800 1000 1200
2ry
Pip
ing
Sod
ium
Lea
k P
rote
ctio
n
2ry Piping Diameter2 mm
Graph 3.1.5.3 2ry Piping Sodium Leak Protection
0
500
1000
1500
2000
2500
3000
0 500 1000 1500 2000 2500 3000 3500 4000
2ry
Co
ola
nt
Inve
ntr
y t
on
1ry Coolant Inventry ton
Graph 3.1.7.2 1ry-2ry Coolant Inventry
1ry＝2ry
1ry＝2×2ry
31
3.2 Component of cooling system (1. Circulation pump)
3.2.1 Installed location of primary coolant circulation pump
Installed locations of primary coolant
circulation pumps are shown in graph 3.2.1.
Merit and demerit of installed location of
pumps are considered especially for each plant,
but many plants have so-called cold leg pump.
3.2.2 Characteristics of primary circulation pump
Enforced power types of primary circulation pumps are shown in graph 3.2.2.1, and
symbols of type are as follows. Mchanical pumps are mainly used.
1 Hot leg2 Cold leg
1
2
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Loca
tio
n o
f P
um
p
Ｔｈｅｒｍａｌ Ｐｏｗｅｒ MW
Graph 3.2.1 Location of 1ry Pump
（ｂlue・Loop, red・Pool）
1
2
0 100 200 300 400 500 600 700
Typ
e o
f P
um
p
1ry Pump Capacity m3/min
Graph 3.2.2.1 Type of 1ry Pump
0
0.2
0.4
0.6
0.8
1
1.2
0 50 100 150 200 250 300 350 400 450 500 550 600 650
Pu
mp
Hea
dM
Pa
1ry Pump Capacity m3/min
Graph 3.2.2.2 1ry Pump Head
0
200
400
600
800
1000
1200
1400
1600
0 50 100 150 200 250 300 350 400 450 500 550 600 650
Pu
mp
Max
imu
m S
pee
dre
v/m
in
1ry Pump Capacity m3/min
Graph 3.2.2.3 1ry Pump Maximum Speed
1
2
3
4
5
6
7
8
9
10
11
12
0 50 100 150 200 250 300 350 400 450 500 550 600 650
Pri
nci
ple
of
Sp
ee
d C
on
tro
l
1ry Pump Capacity m3/min
Graph 3.2.2.4 Principle of 1ry Pump Speed Control
12 two fixed speed11 constant speed10 variable speed alternator9 fluid coupling8 fluid MG coupling7 revolution regulator6 variable speed alternator5 variable frequency4 Variable voltage3 voltage control2 static scherbius1 ward leonard
2 Electrical1 Mechanicaｌ
32
The relations of the flow rate and pump head are shown in graph 3.2.2.2.
Maximum rotating speeds of pumps are shown in graph 3.2.2.3, and methods of pump
speed control are shown in graph 3.2.2.4. Many kind of methods are considered but the
variable frequency methods are used in many pumps.
3.2.3 Electrical driving power of primary circulation pump
Powers of primary circulation pump are shown in graph 3.2.3.1. The powers are
naturally proportional to flow rates.
Ratios of power for decay heat removal and for full power operation are shown in graph
3.2.3.2. This graph shows the necessary power ratio for decay heat removal is about 1 %
of that for full power operation.
3.2.4 Secondary circulation pump
Pump heads of secondary circulation pumps are shown in graph 3.2.4.
For large coolant flow rate plants, pressures
of primary coolant at pump outlet are about
0.6 MPa but secondary a little bit smaller 0.4
MPa.
3.3 Component of cooling system (2. Intermediate Heat Exchanger)
3.3.1 Structure and number of intermediate heat exchanger
There are two ways for radioactive primary coolant flowing on the shell side or in the
heat transfer tubes of intermediate heat exchanger. The kinds of these ways are shown
in graph 3.3.1.1. As the result, the ways of flowing on the shell side are adopted in
almost all plants.
The reason why these ways are adopted depend mainly on the possibility of perfect
drain of radioactive primary coolant.
Numbers of intermediate heat exchangers per each cooling loop are shown in graph
3.3.1.2.
For small power plant, one intermediate heat exchanger is installed, but many large
plants have two intermediate heat exchangers.
0
0.2
0.4
0.6
0.8
1
1.2
0 50 100 150 200 250 300 350 400 450 500 550 600 650
2ry
Pu
mp
Hea
dM
Pa
2ry Pump Capacity m3/min
Graph 3.2.4 2ry Pump Head
0
1000
2000
3000
4000
5000
6000
7000
0 50 100 150 200 250 300 350 400 450 500 550 600 650
Elec
tric
Po
wer
Inp
ut
kW
1ry Pump Capacity m3/min
Graph 3.2.3.1 1ry Pump Electrical Power Input
0.00
0.01
0.02
0.03
0.04
0.05
0.06
0.07
0.08
0.09
0.10
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Rat
io o
f P
um
p E
lect
ric
Po
wer
Inp
ut
Thermal Power MW
Graph 3.2.3.2 1ry Ratio of Decay Heat/Nominal
Pump Electric Power Input
33
3.3.2 Heat transfer capacity of intermediate heat exchanger
Heat transfer capacities of one intermediate heat exchanger are shown in graph
3.3.2.1 and heat transfer area in graph 3.3.2.2. The capacity reaches the ceiling level
about 600MW except one plant, this means the intention of preventing scale up of
capacity of component.
3.3.3 Heat transfer tube of intermediate heat exchanger
Diameter and thickness of heat transfer tubes are shown in graph 3.3.3.1.
The ratios of thickness and diameter are shown in graph 3.3.3.2, then the ratios are
in the range of about 0.04 ～0.10.
0
1
2
3
4
5
6
0
5
10
15
20
25
30
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Thic
knes
sm
m
Dia
met
er
mm
Thermal Power MW
Graph 3.3.3.1 IHX Heat Transfer Tube Diameter, Thickness
Tube Diameter mm
Tube Thickness mm
0.00
0.02
0.04
0.06
0.08
0.10
0.12
0 200 400 600 800 1000 1200 1400 1600 1800 2000
Thic
knes
s/D
iam
eter
IHX Heat Transfer Capacity per Unit MW
IGraph 3.3.3.2 HX Heat Tube Thickness/Diameter
0
200
400
600
800
1000
1200
1400
1600
1800
2000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
IHX
Hea
t Tr
ansf
er
Cap
acit
y/u
nit
MW
Thermal Power MW
Rgaph 3.3.2.1 Heat Transfer Capacity of IHX per Unit
0
500
1000
1500
2000
2500
3000
3500
4000
4500
5000
0 500 1000 1500 2000
IHX
Hea
t Tr
ansf
er A
rea
m2
IHX Heat Transfer Capacity MW
Graph 3.3.2.2 Heat Transfer Area of IHX per Unit
2 In Tubes 1 on Shell side
1
2
0 500 1000 1500 2000 2500 3000 3500 4000 4500
IHX
1ry
Flo
win
g S
ide
Thermal Power MW
Graph 3.3.1.1 1ry Coolant Flowing Side of Intermediate Heat Exchanger
1
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er o
f IH
X
Thermal Power MW
Graph 3.3.1.2 Number of IHX Units per Loop
34
Moreover the lengths of heat transfer tubes are shown in graph 3.3.3.3, the lengths
are 6～8 m.
And numbers of heat transfer tubes are shown in graph 3.3.3.4, the maximum number
is about 6000 except one special case.
3.3.4 Shell of intermediate heat exchanger
Diameter and thickness of shell are shown in graph 3.3.4.1. Then the relations of them
are shown in graph 3.3.4.2. The ratios are about 1/100 except large plants.
3.4 Component of cooling system (3. Steam Generator)
3.4.1 Configuration of steam generator system
Configurations of steam generator systems are shown in graph 3.4.1. For some small
plants their system include re-heater, but for large plants their system are composed by
evaporator + super-heater, or steam generator, that is one shell with evaporator and
super heater.
0
2,000
4,000
6,000
8,000
10,000
12,000
0 500 1000 1500 2000
IHX
Tub
e Le
ngt
hm
m
IHX Heat Transfer Capacity per Unit MW
IGraph 3.3.3.3 Length of HX Heat Transfer Tube
0
1,000
2,000
3,000
4,000
5,000
6,000
7,000
8,000
9,000
10,000
0 500 1000 1500 2000
IHX
He
at T
ran
sfe
r Tu
be
pe
r U
nit
IHX Heat Transfer Area per Unit MW
Graph 3.3.3.4 Number of Heat Transfer Tubes 0f IHX
0
5
10
15
20
25
30
35
40
45
50
55
60
0
1000
2000
3000
4000
5000
6000
0 20 40 60 80 100 120 140
IHX
She
ll Th
ickn
ess
mm
IHX
She
ll D
iam
ete
rm
m
IHX Heat Transfer Capacity per Unit MW
Graph 3.3.4.1 IHX Shell Diameter-Thickness
IHX Shell Diameter mm
IHX Shell Thickness mm
0
5
10
15
20
25
30
35
40
45
0 1000 2000 3000 4000 5000 6000
IHX
Shel
l Th
ickn
ess
mm
IHX Shell Diameter mm
Graph 3.3.4.2 IHX Shell Diameter-Thickness
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Co
nfi
gura
tio
n o
f SG
Thermal Power MW
Graph 3.4.1 Configuration of SG System
SG
EV+SH
EV+SH+RH
35
3.4.2 Number of steam generator system per cooling loop
Numbers of steam generator system are shown in graph 3.4.2.1 for only steam
generator system, graph 3.4.2.2 for evaporator + super-heater system. Number of steam
generator system has only one for steam generator, and one or two for evaporator +
super-heater in many plants.
3.4.3 Type of heat transfer tubes
Types of heat transfer tubes are shown in graph 3.4.3. Helical coiled or straight tube is
adopted in many plants.
3.4.4 Geometrical size of heat transfer tube
Diameters and thickness of heat transfer tubes of evaporator are shown in graph
3.4.4.1, super-heater in graph 3.4.4.2 and re-heater in graph 3.4.4.3.
Diameters of heat transfer tubes are 15～35 mm but scattered in wide range.
The ratio of thickness and diameter of evaporator is shown in graph 3.4.4.4,
super-heater in graph 3.4.4.5 and re-heater in graph 3.4.4.6.
These ratio spread over the range of 0.05～0.18 for evaporator and super-heater, but
the ratio of re-heater is small according to lower pressure in tube.
1
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er p
er L
oo
p
Thermal Power MW
Graph 3.4.2.1 Number of SG per loop(SG)
1
2
3
4
5
6
7
8
9
10
11
12
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Nu
mb
er
pe
r Lo
op
Thermal Power MW
Graph 3.4.2.2 Number of EV per Loop(EV+SH)
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Typ
e o
f SG
Tu
bes
Thermal Power MW
Graph 3.4.3 Type of SG Tubes
Helical Coiled
Straight Tubes
J、U or S Shaped Tubes
0
1
2
3
4
5
6
7
8
0
5
10
15
20
25
30
35
40
0 500 1000 1500 2000 2500 3000 3500 4000 4500
EVTu
be
Thic
knes
sm
m
EVTu
be
Dia
met
erm
m
Thermal Power MWGraph 3.4.4.1 Diameter and Thickness of EV Heat
Transfer tube
EV Tube Diameter mm
EV Tube Thickness mm
0
1
2
3
4
5
6
7
8
0
5
10
15
20
25
30
35
40
0 500 1000 1500 2000 2500 3000 3500 4000 4500
SH T
ub
e Th
ickn
esm
m
SH T
ub
e D
iam
eter
mm
Thermal Power MW
Graph 3.4.4.2 Diameter and Thickness of SH Heat Transfer Tube
SH Tube Diameter mm
SH Tube Thickness mm
36
Ratios of diameter of heat transfer tubes between mutual components are as follows.
For diameter, the ratio of super-heater and evaporator is shown in graph 3.4.4.7 and
re-heater and evaporator in 3.4.4.8.
Diameter and thickness of evaporator and super-heater are almost equal, but diametr
of re-heater ia a little larger and thiskness is a little thinner than evaporator.
3.4.5 Number of heat transfer tubes and heat transfer areas
Number of heat transfer tubes are shown in graph 3.4.5, but their maximum numbers
are about 2000 tubes except one special case.
0
1
2
3
4
5
6
7
8
0
5
10
15
20
25
30
35
40
45
0 500 1000 1500 2000 2500 3000 3500 4000 4500
RH
Tu
bre
Dia
met
erm
m
RH
Tu
be
Dia
met
erm
m
Thermal Power MW
Graph 3.4.4.3 Diameter and Thickness of RH Heat Transfer Tube
RH Tube Diameter mm
RH Tube Thickness mm
0.00
0.02
0.04
0.06
0.08
0.10
0.12
0.14
0.16
0.18
0.20
0 5 10 15 20 25 30 35 40 45
EVTu
be
Thic
knes
s/D
iam
eter
EV Tube Diameter mm
Graph 3.4.4.4 EV Tube Thickness /Diameter
0.00
0.02
0.04
0.06
0.08
0.10
0.12
0.14
0.16
0.18
0.20
0 5 10 15 20 25 30 35 40 45
SHTu
be
Thic
kne
ss/D
iam
eter
SH Tube Diameter mm
Graph 3.4.4.5 SH Tube Thickness/Diameter
0.00
0.02
0.04
0.06
0.08
0.10
0.12
0.14
0.16
0.18
0.20
0 5 10 15 20 25 30 35 40 45
RH
Tub
e T
hic
kne
ss/D
iam
ete
r
RH Tube Diameter mm
Graph 3.4.4.6 RH Tube Thickness/Diameter
0.4
0.6
0.8
1.0
1.2
1.4
1.6
0 5 10 15 20 25 30 35 40 45
SH/E
VTu
be
Dia
met
er
EV Tube Diameter mm
Grsph 3.4.4.7 SH/EV Tube Diameter
0.4
0.6
0.8
1.0
1.2
1.4
1.6
0 5 10 15 20 25 30 35 40 45
RH
/EV
Tub
e D
iam
eter
EV Tube Diameter mm
Graph 3.4.4.8 RH/EV Tube Diameter
37
3.4.6 Heat transfer capacity of steam generator
Heat transfer capacity is shown in graph
3.4.6. For large plant, heat transfer area
can be assumed from the data of number of
loops, configuration of system and number
of units.
3.5 Steam turbine generator
3.5.1 Number of steam turbine generator
Number of steam turbine generators are shown in graph 3.5.1.
Usually number of steam turbine generator
is one, but for large plant it seems two steam
turbine generators are installed.
3.5.2 Speed of steam turbine generator
Rotating speed of turbine generator is shown in graph 3.5.2. For relating to
commercialized electric cycle in each nation, speed of generator is 1500-3000 or
1800-3600 r.p.m., but many has 3000 r.p.m., 50 cycles per second is overwhelmingly
used.
1
2
3
0 200 400 600 800 1000 1200 1400 1600 1800
Nu
mb
er
of
Turb
ine
Electric Power MW
Graph 3.5.1 Number of Turbine Generator
0
200
400
600
800
1000
1200
1400
1600
1800
2000
0
50
100
150
200
250
300
350
0 500 1000 1500 2000
EVH
eat
Cap
acit
yM
W/U
nit
SH R
HH
eat
Cap
acit
yM
W/U
nit
SG Heat Capacity /Loop MW
Graph 3.4.6 SG Heat Capacity
SH Heat Capacity
RH Heat Capacity
EV Heat Capacity
0
1000
2000
3000
4000
5000
6000
7000
8000
0 500 1000 1500 2000
Nu
mb
er
of
Tub
es
SG Heat Capacity MW
Graph 3.4.5 SG Heat Capacity-Number of EV Tubes
38
3.5.3 Steam condition at turbine inlet
Steam condition at the inlet of steam turbine are shown in graph 3.5.3. The conditions
approach for pressure up to 28MPa and for
temperature up to 530℃.
For reference, critical state of water is
pressure 22.12MPa and temperature
647.3 ℃.
3.5.4 Minimum pressure in steam condenser
Minimum pressure in condenser are shown in graph 3.5.4.
1800
2400
3000
3600
0 500 1000 1500 2000
Spe
ed o
f Tu
rbin
ere
v/m
in
Electric Power MW
Grsph 3.5.2 Speed of Turbine Generator
0
5
10
15
20
25
30
380 400 420 440 460 480 500 520 540
Stea
m P
ress
ure
MP
a
Steam Temperature ℃Graph 3.5.3 Steam Condition at Turbine Inlet under
Full Power
0.000
0.002
0.004
0.006
0.008
0.010
0.012
0 200 400 600 800 1000 1200 1400 1600 1800
Min
imu
m P
ress
ure
MP
a
Electric Power MW
Graph 3.5.4 Minimum Condenser Pressure
39
4 Formation of other systems and components
4.1 Refueling system etc.
4.1.1 Method of refueling
Many kinds of in-core refueling system are examined, so these method are shown in
graph4.1.1.
Triple rotating plugs + 1 Vertical mechanism1 and double rotating plugs + 1 Vertical
mechanism are adopted in many plants.
4.1.2 Method for storing and cooling of spent fuel
Storage places for spent fuel are shown in graph 4.1.2.
For large plant, method of storage outside
primary vessel but inside secondary
containment or method of storage outside
secondary containment seems to be adopted.
4.1.3 Path for transport of fuel
Method for transport of spent fuel are shown in graph 4.1.3.
So called transfer mechanism are used in
many plants, also A-flame system in some
plants
.
3 OSC storage outside secondary containment 2 OPV storage outside primary vessel
but inside secondary containment 1 ORB storage in diagrid positions outside radial blanket
2
3
4
5
6
7
8
9
10
0 200 400 600 800 1000 1200 1400 1600 1800
Met
ho
ds
Electric Power MW
Graph 4.1.1 Refueling Methods
RP PM VM FM10 2 29 3 18 2 27 2 16 2 15 2 14 1 33 1 12 1 1
RP rotating plugPM pantograph mechanismVM vertical mechanismFM fixed-arm mechanism
10 2RP+2PM9 3RP+1VM8 2RP+2VM7 2RP+1PM6 2RP+1FM5 2RP+1VM4 1RP+3VM3 1RP+1PM2 1RP+1FM
1
2
3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Spe
nt
Fue
l Sto
rin
g M
eth
od
s
Thermal Power MW
Graph 4.1.2 Methods used to store Spent Fuel
1
2
3
4
5
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Fue
l Tra
nsf
er
Met
ho
d
Thermal Power MW
Graph 4.1.3 Method used to handle Fuel outside Primary Vessel
5 CC cask car4 TA transfer within an A-frame3 MF mobile transfer flask2 TM transfer mechanism1 MC mobile cask
40
4.2 Secondary containment
4.2.1 Geometry of secondary containment
Geometry of secondary containment are shown in graph 4.2.1.
Cylindrical with dome concept is adopted in many plants, but for some large plant
rectangular building is used for secondary containment.
4.2.2 The ratio of volume of containment and reactor vessel depending on reactor type
The ratio of volume of secondary containment and that of reactor vessel are shown in
graph 4.2.2.
The ratio of them are a few hundred times
for loop type, but several ten times for pool
type reactors.
4.3 Coolant purification system
4.3.1 Number of cold trap
So-called cold trap is used for purification of coolant, and number of cold traps are
shown in graph4.3.1.1.
3 Rectangular Building2 Sphere1 Cylindrical with Dome
0
1
2
3
4
5
6
7
8
0 1 2 3 4 5 6 7 8
2ru
Mes
h V
olu
me
m3
1ry Mesh Volume m3
Graph 4.3.1.2 Volume of Mesh Region in Cold Trap
1ry-2ry
1
2
3
0 50,000 100,000 150,000 200,000 250,000 300,000 350,000
Ge
om
etry
of
Co
nta
inm
ent
Volume of Containment Building m3
Graph 4.2.1 Geometry of Secondary Containment Building
0
10
20
30
40
50
60
70
80
0
100
200
300
400
500
600
700
800
900
1,000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Po
olT
ype
Vo
lum
e R
atio
Loo
pTy
pe
Vo
lum
e R
atio
Electric Power MW
Graph 4.2.2 Ratio of Gontainment/Reactor Vessel Volume
Loop Type Pool Type
0
5
10
15
20
25
0 5 10 15 20 25
2ry
Co
ld T
rap
1ry Cold Trap
Graph 4.3.1.1 Number of Cold Traps/loop 1ry -2ry
41
Number of cold trap in secondary is larger than that of primary, instead of smaller 2ry
inventories. The volumes of mesh region for trapping impurity are shown in graph
4.3.1.2, but the volume of them in primary is nearly equal to secondary for many plants.
Kind of coolant for cold traps are shown in graph 4.3.1.3.
Organic, Nak or gas is used for primary, but air is mainly used for secondary.
6 dowthom5 gas4 NaK3 air2 organic1 nitrogen
0
1
2
3
4
5
6
0 1 2 3 4 5 6
2ry
Co
ld T
rap
Co
ola
nt
1ry Cold Trap Coolant
Grahp 4.3.1.3 Coolant for Cold Trap 1ry-2ry
3
2
4
2 3
42
5 Nuclear characteristics
There are little data of reactor characteristics in the IAEA data book, but some typical
characteristics are shown in this document.
5.1 Volume ratio in the core
5.1.1 Core volume fraction of fuel, coolant, steel and void
Core volume fractions of fuel, coolant, steel and void in each reactor are shown in
graph 5.1.1.1.
Graph 5.1.1.2 shows these ratios for depending on power.
These volume ratios are scattered in wide range, but generaly, their averaged value of
fuel is about 35%, coolant about 35% and steel 25% respectively. But some large plants
have the ratio of fuel larger than 40%.
The ratios of fuel and coolant, most effective factor for reactor physics, are shown in
graph 5.1.1.3, so these ratios are around 1.0.
5.2 Neutron flux
5.2.1 Maximum neutron flux
Maximum neutron fluxes in the core are shown in graph 5.2.1.1.
The unit of neutron flux is neutron/cm2/sec and their values are in the range of the
15th power of ten, but scattered widely.
The ratios of maximum and average neutron flux are shown in graph 5.2.1.2, them
their value are around 1.6.
0
10
20
30
40
50
60
70
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Vo
lum
e Fr
acti
on
%
Thermal Power MW
Graph 5.1.1.2 Volume Fraction of Fuel,Coolant,Steel
Fuel
Coolant
Steel
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
1.8
2.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Fuel
/Co
ola
nt
Vo
lum
e Fr
acti
on
Thermal Power MW
Graph 5.1.1.3 Ratio of Fuel/Coolant Volume Fraction
0%
10%
20%
30%
40%
50%
60%
70%
80%
90%
100%
Vo
lum
e Fr
acti
on
Thermal Power MW
Graph 5.1.1.1 Core Volume Fraction of Fuel,Coolant,Steel,Void or Fission Gas Volume
Void
Steel
Coolant
Fuel
43
5.3 Linear power
5.3.1 Maximum linear power
The maximum linear powers are shown in graph 5.3.1.1, and the ratios of their
maximum and average in graph 5.3.1.2.
The linear powers have values in the range of 40 to 50 under the unit of kW/m except
small reactors. And the ratios of maximum and average of them are about 1.6.
5.3.2 Neutron flux and linear power
The relations between maximum neutron flux and linear power are shown in graph
5.3.2, the value of linear power are almost
constant instead of the value of neutron flux.
In other words, reactor characteristics are
particularly specified by linear power than
maximum neutron flux.
This means that the linear power are little
influenced by the value of detail specifications
of reactor physics, and from this point the
linear power is typical specification for heat
transfer to cooling system.
5.4 Power density
5.4.1 Power density
Average neutron flux, linear power and power density are shown in graph 5.4.1.1.
Power density spread widely compared with linear power and neutron flux.
The ratios of maximum and average of power density are shown in graph 5.4.1.2, the
0
2
4
6
8
10
12
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
.N
eu
tro
n F
lux
×1
0＾
15
n/c
m2
sec
Thermal Power MW
Graph 5.2.1.1 Maximum Neutron Flux
1.0
1.2
1.4
1.6
1.8
2.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
/Mea
n N
eu
tro
n F
lux
Thermal Power MW
Graph 5.2.1.2 Ratio of Max/Mean Neutron Flux
0
10
20
30
40
50
60
0 2 4 6 8 10 12
Max
. Lin
ear
Po
we
rkW
/m
Max Neutron Flux ×10＾15 n/cm2sec
Graph 5.3.2 Maximum Neutron Flux-Linear Power
0
10
20
30
40
50
60
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
Lin
ear
Po
wer
kW/m
Thermal Power MW
Graph 5.3.1.1 Maximum Linear Power
1.0
1.2
1.4
1.6
1.8
2.0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
/Mea
n L
ine
ar P
ow
er
Thermal Power MW
Graph 5.3.1.2 Max/Mean Linear Power
44
ratios are also about 1.6.
5.5 Enrichment of plutonium
5.5.1 Enrichment of plutonium(in two zone core)
Values of plutonium enrichments in two zone core are shown in graph 5.5.1.1.
Their values are about 20% except these of small reactors.
The ratios of plutonium enrichment in core zone 1 and 2 are shown in graph 5.5.1.2,
their values are considerably scattered in the range of 1.1～1.5.
This means the design is aiming at equalizing maximum neutron flux in zone 1 and 2,
maximum heat generation rate, maximum cladding temperature or other
characteristics of the core.
5.6 Total breeding gain
Total breeding gain is defined as
(rate of creation of atoms₋rate of destruction of atoms)/(rate of destruction atoms) in the
reactor.
From this definition, so called breeding ratio is ‘total breeding gain +1’ in the case of over
the value 1.0.
5.6.1 Total breeding gain 1 (by fuel geometry)
Total breeding gains, the typical characteristic specification for FR, are shown in graph
5.6.1.1.
For small reactors their values are rather scattered, but for large reactors the target
value of them seems to be around 0.1.
0
500
1000
1500
2000
2500
0
5
10
15
20
25
30
35
40
45
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Po
we
r D
en
sity
Ne
utr
on
Flu
x, L
ine
ar P
ow
er
Thermal Power MW
Graph 5.4.1.1 Average Neutron Flux, Linear Power, Power Density
Average Neutron Flux ×10＾15 n/cm2sec
Average Linear Power kW/m
Average Power Density kW/l
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Plu
ton
ium
En
rich
me
nt
%
Thermal Power MW
Graph 5.5.1.1 Fuel Plutonium Enrichment ofPlant Classification
Inner Core %
Inner Core %
Inner Core %
Outer Core %
Outer Core %
Outer Core %
1.0
1.1
1.2
1.3
1.4
1.5
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Rat
io o
f O
ute
r/In
ne
r P
u E
nri
chm
ent
Thermal Power MW
Graph 5.5.1.2 Ratio of Outer/Inner Plutonium Enrichment
1.0
1.2
1.4
1.6
1.8
2.0
2.2
2.4
2.6
2.8
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
/Ave
rage
Po
wer
de
nsi
ty
Thermal Power MW
Graph 5.4.1.2 Maximum/Average Power Density
45
For investing the special specifications related to total breeding gain, the relations
between diameters of fuel pellet to total breeding gain are shown in graph 5.6.1.2,
diameters of fuel cladding to total breeding gain in graph 5.6.1.3. Both specifications are
not directly related to breeding gain.
It seems that breeding gain is not simple specification but is related to many
specifications.
5.6.2 Total breeding gain 2 (by radial blanket)
The relations between number of radial blanket row and total breeding gain are shown
in graph 5.6.2.
In the case of number of blanket layer larger
than 2, breeding gains increase and some
reactor have the breeding gain reaching the
value of larger than 0.20.
5.6.3 Breeding gain 3 (by upper and lower blanket)
For investigating the length of upper and lower blanket in this document, the ratio of
blanket length and neutron mean free path is conveniently used just like the row
numbers of radial blanket.
Normalized lengths, the ratios of length and neutron mean free path, are shown in
graph 5.6.3.1 for upper blanket, in graph 5.6.3.2 for lower blanket. Both normalized
lengths are 2～3 for almost all reactors, so it seems that this fact means the same
-0.2
-0.1
0.0
0.1
0.2
0.3
0 500 1000 1500 2000 2500 3000 3500 4000 4500Toat
l Bre
edin
g G
ain
Thermal Power MW
Graph 5.6.1.1 Total Breeding Gain(thermal power)
-0.2
-0.1
0.0
0.1
0.2
0.3
3 4 5 6 7 8Tota
l Bre
ed
ing
Gai
n
Fuel Pellet Diameter mm
Graph 5.6.1.2 Total Breeding Gain(calculated fuel pellet diameter)
-0.2
-0.1
0.0
0.1
0.2
0.3
3 4 5 6 7 8 9 10 11
Tota
l Bre
ed
ing
Gai
n
Outer Diameter of Core Fuel Pin mm
Graph 5.6.1.3 Total Breeding Gain(diameter of fuel cladding)
-0.20
-0.15
-0.10
-0.05
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0 1 2 3
Tota
l Bre
ed
ing
Gai
n
Number of Radial Blanket Row
Graph 5.6.2 Number of Radial Blanket -Breeding Gain
46
method of decision on radial blanket.
5.7 Fuel burn up
5.7.１ Fuel burn up
Values of averaged fuel burn up of each plant are shown in graph 5.7.1.1. For large
reactors, the fuel burn up seems higher than 100 thousand MWd/t of heavy metal. And,
the ratios of maximum and average burn up are shown in graph 5.7.1.2, but these
values are about 1.5, which are lower than other maximum/average ratios.
5.8 Reactivity and Doppler coefficient
5.8.1 Reactivity
Reactivity of the core is the most important specification on reactor physical
-20
-18
-16
-14
-12
-10
-8
-6
-4
-2
0
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Rea
ctiv
ity
Co
effi
cien
ts
Thermal Power MW
Graph 5.8.1.1 Temperature and Power Reactivity Coefficient
Isothermal Temperature Coefficientsat Full Power, 20 ℃ pcm/℃
Total Power Coefficient at Full Power pcm/MWth
-25
-20
-15
-10
-5
0
5
10
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
. C
oo
lan
t V
oid
Co
effi
cien
t
Thermal Power MW
Graph 5.8.1.2 Maximum Coolant Void Effect, including only regions with a Positive Coolant
Reactivity Worth dollars
0
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Equ
ival
en
t R
ow
Nu
mb
er
of
Up
pe
r A
xal
Bla
nke
t
Thermal Power MW
Graph 5.6.3.1 Equiv alent Row Number of Thickness of Upper Axial Blanket
0
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500Equ
ival
en
t N
um
be
r o
f Lo
we
r A
xial
B
lan
ket
Thermal Power MW
Graph 5.6.3.2 Equivalent Row Number of Thickness of Lower Axial Blanket
0
20,000
40,000
60,000
80,000
100,000
120,000
140,000
160,000
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Ave
rage
d B
urn
up
MW
d/t
Thermal Power MW
Graph 5.7.1.1 Average Achieved Burnup
1.0
1.2
1.4
1.6
1.8
2.0
2.2
2.4
2.6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
/Ave
rage
Bu
rnu
p
Thermal Power MW
Graph 5.7.1.2 Ratio of Maximum/Average Achieved Burnup
47
characteristics. These reactivity coefficients depending on temperature and on total
power are shown in graph 5.8.1.1. Naturally both reactivity coefficients have the minus
value.
And maximum void coefficients are shown in graph 5.8.1.2. These coefficients scatter
in wide range, but some reactors have positive value of slightly larger than 0.0.
5.8.2 Doppler coefficient
Doppler coefficients of containing void and without void in the core are shown in graph
5.8.2.
In general, the larger reactors have larger
minus value of Doppler coefficient.
-0.010
-0.008
-0.006
-0.004
-0.002
0.000
0.002
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Do
pp
ler
Co
effi
cien
t
Thermal Power MW
Graph 5.8.2 Doppler Coefficient
Voided Core Tdk/dt
Unvoided Core Tdk/dt
48
6 Cooling characteristics in the core
6.1 Hydrodynamics of coolant in the core
6.1.1 Coolant velocity in the core
Cooling capability of core depends on coolant velocity, so the maximum velocities of
coolant are shown in graph 6.1.1.
Being independence on the reactor scale,
coolant velocity seems to have the limited
value. Because the higher velocity of sodium
flowing in narrow clearance among fuel
elements makes unsuitable vibration of fuel
elements and larger flow resistance by
hydrodynamic characteristics.
6.1.2 Coolant pressure at core inlet and outlet
Values of coolant pressure at core inlet of each plant are shown in graph 6.1.2.1, at
core outlet in graph 6.1.2.2.
6.1.3 Pressure drop by coolant flow across the core
Values of pressure drop by coolant flow across the core are shown in graph 6.1.3.
The values are equal to about 0.5 MPa.
0
0.2
0.4
0.6
0.8
1
1.2
1.4
1.6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Co
re In
let
Co
ola
nt
Pre
ssu
reM
Pa
Thermal Power MW
Graph 6.1.2.1 Coolant Pressure at Core Inlet
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Co
re O
utl
et C
oo
lan
t P
ress
ure
MP
a
Thermal Power MW
Graph 6.1.2.2 Coolant Pressure at Core Outlet
0
2
4
6
8
10
12
14
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
Co
ola
nt
Vel
oci
tym
/s
Thermal Power MW
Graph 6.1.1 Maximum Coolant Velocity in Core
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Pre
ssu
re D
rop
MP
a
Thermal Power MW
Graph 6.1.3 Pressure Drop across Core
49
6.2 Maximum coolant temperature
6.2.1 Coolant temperature at core outlet
One of the most important specification of FR characteristics is the1ry coolant
temperature at core outlet, then the values of their temperature are shown in graph
6.2.1.
Except small reactors, their temperatures
have the value of about 540℃. These values
are seemed to be decided from the design
limitations of maximum temperature of fuel,
cladding and coolant.
6.2.2 Effect on maximum linear power
The relations between maximum linear power and primary coolant maximum
temperature, namely the temperature at core outlet, are shown in graph 6.2.2.
The correlation between them seems to be
week.
6.3 Maximum surface temperature of fuel cladding
6.3.1 Maximum surface temperature of fuel cladding
The fuel cladding material is exposed to the most severe heated and irradiated
conditions. As the result of this phenomena, the maximum cladding surface
temperature has the higher possibility of being above its design limitations.
So, in this section, the investigations are made about the effect by various
specifications.
6.3.2 Effect of coolant velocity
Effects of coolant velocity for maximum cladding surface temperature are shown in
graph 6.3.2.
The values of maximum cladding surface temperature are almost constant in spite of
the coolant velocity.
25
30
35
40
45
50
55
400 450 500 550 600
Max
. Lin
ear
Po
wer
kW/m
Max. 1ry Coolant Temperature ℃
Graph 6.2.2 Max Linear Power-Max 1ry Temperature
300
350
400
450
500
550
600
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Max
. 1ry
Co
ola
nt
Tem
pe
ratu
re℃
Thermal Power MW
Graph 6.2.1 Maximum Coolant Temperature in 1ry Circuit at outlet of core
50
6.3.3 Effect of maximum linear power
The relations between maximum linear power and maximum cladding surface
temperature are shown in graph 6.3.3. Their values scatter in wide range, this is
because the maximum cladding surface
temperature is related to both heat
generation and cooling capability.
The maximum cladding surface
temperatures are limited under about 700℃
by the properties of used cladding material.
6.3.4 Effect of fuel pellet diameter
Effects of pellet diameter to the maximum cladding surface temperature are shown in
graph 6.3.4.
The maximum cladding surface
temperatures also scatter, this means their
temperatures are effected many specifications
including geometry of fuel.
6.3.5 Effect of fuel cladding diameter
The relations between the maximum cladding surface temperature and fuel cladding
diameter are shown in graph 6.3.5.
There are no special relations like fuel pellet diameter.
400
450
500
550
600
650
700
750
0 2 4 6 8 10 12 14
Max
. C
lad
din
g Te
mp
erat
ure
℃
Max. Coolant Velocity m/s
Graph 6.3.2 Maximum Cladding Surface
temperature of Core Fuel Pin－Coolant Veloｃity
550
600
650
700
750
25 30 35 40 45 50 55
Max
Cla
dd
ing
Tem
pe
ratu
re℃
Max Linear Power kW/m
Graph 6.3.3 Maximum Cladding Surface Temperature of Core Fuel Pin-Linear Power
550
600
650
700
750
3 4 5 6 7 8
Max
.C
lad
din
g Te
mp
era
ture
℃
Pellet Diameter mm
Graph 6.3.4 Maximum Cladding Surface Temperature of Core Fuel Pin -Diameter of Pellet
51
550
600
650
700
750
3 4 5 6 7 8 9 10 11
Max
. C
lad
din
g Te
mp
era
ture
℃
Cladding Diameter mm
Graph 6.3.5 Maximum Cladding Surface Yemperature of Core Fuel Pin-Diameter of Cladding
52
7 Plant cooling characteristics
7.1 Temperature of cooling system
7.1.1 Temperature of cooling system
As the typical specifications of cooling system, the primary coolant temperatures of at
core outlet, secondary coolant temperature at intermediate heat exchanger outlet and
steam temperature at turbine inlet are shown in graph 7.1.1.
In natural, these temperatures are lined
with the order of their values, and the
temperature differences of them are a little
larger value than 20 ℃ respectively.
7.1.2 Temperature of hot leg of cooling system
Coolant temperatures at core outlet are shown in graph 7.1.2.1, steam temperature at
turbine inlet in graph 7.1.2.2.
7.1.3 Relation between primary coolant flow rate and temperature difference of hot/cold leg
Relation between primary coolant flow rates and temperature differences of hot/cold
leg are shown in graph 7.1.3.
Their temperature differences are about
150℃ for large plants, this phenomena
show the flow rates are proportional to
thermal powers of plants.
7.1.4 Temperature difference of secondary/primary cooling system
The relations between temperature differences of hot/cold leg of primary and that of
400
420
440
460
480
500
520
540
560
580
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Co
ola
nt
Tem
per
atu
re℃
Thermal Power MW
Graph 7.1.1 Temperature odf Coolant System
Reactor Outlet ℃
IHX 2ry Outlet ℃
Turbine Inlet ℃
200
250
300
350
400
450
500
550
600
0 500 1000 1500 2000 2500 3000 3500 4000 45001ry
co
ola
nt
Tem
pe
ratu
re a
t IH
X In
let ℃
Thermal Power MWGraph 7.1.2.1 Mixed Coolant Temperature in 1ry
Circuit at IHX Inlet
200
250
300
350
400
450
500
550
600
0 200 400 600 800 1000 1200 1400 1600 1800
Stea
m T
emp
erat
ure
at
Turb
ine
Inle
t℃
Thermal Power MW
Graph 7.1.2.2 Steam Temperature at Turbine Inlet
100
120
140
160
180
200
220
0 2,500 5,000 7,500 10,000 12,500 15,000 17,500 20,0001ry
Ho
t-C
old
Tem
per
atu
re D
iffe
ren
ce℃
1ry Coolant Flow Rate kg/2Graph 7.1.3 1ry Temperature Difference between
Hot and Cold Leg
53
secondary cooling system are shown in graph 7.1.4.
Temperature differences of secondary are a
little higher than these of primary cooling
system.
7.1.5 Effects of various temperature difference
One of the typical temperature difference for effecting the plant characteristics is the
temperature difference between core outlet and turbine inlet.
The temperature differences between core outlet coolant and turbine inlet steam are
shown in graph 7.1.5.1. This graph indicates these temperature differences are about
50℃.
Rewriting this graph with using plant scale, these temperature differences are shown
in graph 7.1.5.2, and their temperature differences are constant about 50℃.
Smaller their temperature difference makes better for heat availability.
7.2 Temperature of intermediate heat exchanger
7.2.1 Coolant temperature of intermediate heat exchanger
Temperatures of secondary coolant at inlet and outlet of intermediate heat exchanger
are shown in graph 7.2.1.
This graph indicates a little higher than
500℃ at outlet、and about 350℃ at outlet.
400
420
440
460
480
500
520
540
560
580
400 420 440 460 480 500 520 540 560 580
Turb
ine
Inle
t Ta
mp
erat
ure
℃
Reactor Outlet Temperature ℃
Graph 7.1.5.1 Reactor Outlet -Turbine InletTemperature
ΔT=50℃
ΔT=0℃
ΔT=100℃
0
20
40
60
80
100
120
140
160
180
200
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Rea
cto
r O
utl
et -
Turb
ine
Inle
t
Tem
per
atu
re℃
Thermal Power MW
Grsph 7.1.5.2 Temperatu Difference between Reactor Outlet and Turbine Inlet
100
120
140
160
180
200
220
240
260
280
100 120 140 160 180 200 220
2ry
Co
ola
nt
Tem
pe
ratu
re D
iffe
ren
ce℃
1ry Coolant Temperature Difference ℃
Graph 7.1.4 Temperature Difference of 1y and 2ry hot/cold Coolant
100
200
300
400
500
600
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Tem
pe
ratu
re o
f IH
X℃
Thermal Power MW
Graph 7.2.1 2ry Coolant Temperature of IHX
IHX 2ry Inlet Temperature ℃
IHX 2ry Outlet Temperature ℃
54
7.2.2 Temperature difference of primary and secondary coolant of intermediate heat
exchanger
Temperature differences of primary and secondary coolant of intermediate heat
exchanger are shown in graph 7.2.2.
7.3 Temperature of steam generator
7.3.1 Steam temperature and pressure at outlet of steam generator
Steam temperatures at outlet of steam generator are shown in graph 7.3.1.1, pressures
in graph 7.3.1.2.
The temperatures are almost constant 500℃、the pressures scatter in wide range.
Also the relations between the temperature and pressure are shown in graph 7.3.1.3.
7.3.2 Temperature difference between water at inlet and steam at outlet
Temperature differences between water at inlet and steam at outletare are shown in
graph 7.3.2.
200
250
300
350
400
450
500
550
0 500 1000 1500 2000 2500 3000 3500 4000 4500
SHO
utl
et T
em
pe
ratu
re℃
Thermal Power MW
Graph 7.3.1.1 Steam Temperature at SH Outlet
0
5
10
15
20
25
30
0 500 1000 1500 2000 2500 3000 3500 4000 4500
SH O
utl
et P
ress
ure
MP
a
Thermal Power MW
Graph 7.3.1.2 Steam Pressure at SH Outlet
100
120
140
160
180
200
220
240
260
100 120 140 160 180 200 220
IHX
2ry
Co
ola
nt
Tem
per
atu
re
Dif
fere
nce
℃
IHX 1ry Coolant Temperature Difference ℃
Graph 7.2.2 IHX 1ry and 2ry Coolant Temperature Difference
1ry＝2ry
1.2×1ry＝2ry
0
5
10
15
20
25
30
400 420 440 460 480 500 520 540
SH O
utl
et S
team
Pre
ssu
re M
Pa
SH Outlet Steam Temperature ℃
Graph 7.3.1.3 Steam Temperature and Pressure at SH Outlet
55
Valiation of specific enthalpy of water-steam
system is remarkable because of their phase
change, so, these temperature difference has
not special meaning but capability for easy
understanding.
7.4 Plant thermal efficiency
Plant thermal efficiency is affected by various specifications, but in this section some of
them are investigated.
7.4.1 Plant thermal efficiency 1 (influence of plant scale)
First, the relations between plant scale and thermal efficiency are shown in graph
7.4.1.
For fast reactors, their thermal efficiencies are realized the value of higher than 40%
by high temperature intention.
7.4.2 Plant thermal efficiency 2 (influence of temperature)
The effects of turbine inlet temperature for thermal efficiency are shown in graph
7.4.2.1, reactor outlet temperature in graph 7.4.2.2.
Generally speaking, the thermal efficiency and turbine inlet temperature has mutual
correlation, but little correlation between thermal efficiency and core outlet
temperature.
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0.35
0.40
0.45
0.50
200 250 300 350 400 450 500 550
The
rmal
eff
icie
ncy
%
Turbine Inlet Steam Temperature ℃
Graph 7.4.2.1 Thermal Efficiency (Turbine Temperature)
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0.35
0.40
0.45
0.50
400 420 440 460 480 500 520 540 560 580
The
rmal
Eff
icie
ncy
%
Reactor Outlet Temperature ℃
Graph 7.4.2.2 Thermal Efficiency (Reactor Outlet Temperature)
0
50
100
150
200
250
300
350
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Tem
per
atu
re D
iffe
ren
ce o
f St
eam
an
d
Wat
er℃
Thermal Power MW
Graph7.3.2 STemperature difference of SG Outlet and Inlet
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0.35
0.40
0.45
0.50
0 200 400 600 800 1000 1200 1400 1600 1800
The
rmal
eff
icie
ncy
%
Electric Power MW
Graph 7.4.1 Thermal Efficiency (Electric Power)
56
7.4.3 Plant thermal efficiency 3 (influence of temperature difference)
As the candidate specifications affecting to the thermal efficiency, temperature
differences between reactor outlet and turbine inlet are shown in graph 7.4.3.1, and
between turbine inlet and feed water
temperature in graph 7.4.3.2. Both specifications are given little effects.
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0.35
0.40
0.45
0.50
0 50 100 150 200
The
rmal
Eff
icie
ncy
%
Reactor Outlet -Turbine Inlet Temperature Graph 7.4.3.1 Heat Efficiency (Temperature
Difference 1)
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0.35
0.40
0.45
0.50
0 50 100 150 200 250 300
The
rmal
Eff
icie
ncy
%
Turbine Inlet - Feedwater Temperature Graph 7.4.3.2 Heat Efficiency (Temperature
Difference 2)
57
8 Structural integrity (including material properties)
8.1 Fuel cladding tube
8.1.1 Cladding material
Kinds of material of fuel cladding tubes are shown in graph 8.1.1.1.
Many kinds of materials are adopted, but 316 SS are used in wide temperature range.
The cladding materials used for already constructed plants are shown in graph 8.1.1.2.
This graph shows chromium steels are used
in recent years.
8.1.2 Gas pressure produced by nuclear fission
The relations between FP gas pressure and average fuel burn up are shown in graph
8.1.2.1. It seems their pressures are up to several MPa.
Here, for presenting evaluation of stress in cladding tubes, the relations between fuel
burn up and cladding thickness are shown in graph 8.1.2.2.
8.1.3 Circumferential stress in cladding by FP gas pressure (including safety factor)
Here, design stress happened in cladding material, which is the basic specification for
13. ODS12. HT-911. PE-1610. EP-8239. 1.4970 SS 8. niobium7. Zr6. Cr17Ni135. Cr16Ni154. Cr16Ni113. Cr15Ni152. 12Crsteel1. 316SS
1
2
3
4
5
6
7
8
9
10
11
12
13
560 580 600 620 640 660 680 700 720 740
Cla
dd
ing
Mat
eri
al
Max. Temperature of Cladding ℃
Graph 8.1.1.1 Cladding Material
1
2
3
4
5
6
7
8
9
10
11
12
13
1950 1960 1970 1980 1990 2000 2010
Cla
dd
ing
Mat
eria
l
Dates of Start of Construction CY
Graph 8.1.1.2 Cladding Material at start of Construction
0
200
400
600
800
1,000
1,200
1,400
0 500 1,000 1,500 2,000 2,500 3,000 3,500 4,000
Gas
Ple
nu
mLe
ngt
hm
m
Fuel Element Length mm
Graph 2.1.8.2 Length of FP Gas Plenum
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0 50,000 100,000 150,000 200,000 250,000 300,000
Thic
knes
s o
f C
lad
din
g m
m
Maximum Burnup MWd/t
Graph 8.1.2.2 Thickness of Cladding
58
deciding thickness of cladding tube, is investigated in this section.
First, calculated values of circumferential direction stress are shown in graph 8.1.3.1
for thermal power, in graph 8.1.3.2 for maximum cladding temperature, in graph 8.1.3.3
for fuel storage time in the core.
These stress scatter in wide range, but their value are about 40 MPa
8.1.4 Circumferential stress in cladding by kinds of material (including safety factor)
The relations between circumferential stress and kinds of cladding materials are
shown in graph 8.1.4.
In this graph, two data, having the highest stress value, are belonging to experimental
reactors.
Except these data, the ODS design, having stress about 80MPa which is much higher
than others, is remarkable for future design. If the relations between cladding
materials and starting time of the plants are published, it is possible to understand the
development of core design. But some documents have detail data of cladding materials,
following these documents and getting the history of cladding development are very
13. ODS12. HT-911. PE-1610. EP-8239. 1.4970 SS 8. niobium7. Zr6. Cr17Ni135. Cr16Ni154. Cr16Ni113. Cr15Ni152. 12Crsteel1. 316SS
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500 4000 4500
,Cir
cum
fere
nti
al S
tres
sM
Pa
Thermal Power MW
Graph 8.1.3.1 Circumferential Stress of Cladding
0
10
20
30
40
50
60
70
80
90
100
500 550 600 650 700 750 800
Cir
cum
fere
nti
al S
tres
sM
Pa
Max. Cladding Temperature ℃
Graph 8.1.3.2 Circumferential Stress of Cladding
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500
Cir
cum
fere
nti
al S
tre
ssM
Pa
Mean Residence Time in Core days
Graph 8.1.3.3 Circumferential Stress of Cladding
1
2
3
4
5
6
7
8
9
10
11
12
13
20 30 40 50 60 70 80 90 100
Mat
eria
l of
Cla
dd
ing
Circumferential Stress of Cladding MPa
Grsph 8.1.4 Material and Stress of Cladding
59
interesting. For example, the document on ODS design was published in 2003.
8.2 Fuel element spacer, cladding of blanket, wrapper tube
8.2.1 Material of fuel element spacer
Materials of fuel element spacer, keeping the clearance between the fuel elements, are
shown in graph 8.2.1.
Comparing with fuel cladding, less various materials are used, but many materials are
tried to use for spacers.
8.2.2 Material of blanket cladding
Materials of blanket cladding are shown in graph 8.2.2.
Many kinds of materials are used like fuel claddings.
8.2.3 Material of wrapper tube of subassemblies
Materials of wrapper tubes are shown in graph 8.2.3.
Their materials are same as those of fuel
cladding and more other materials.
8.3 Neutron absorber
1
2
3
4
5
6
7
8
9
10
560 580 600 620 640 660 680 700 720
Spac
er
Mat
eri
al
Max. Temperature of Cladding ℃
Graph 8.2.1 Material of Mechanical Separation of Pins
10 EM109 PNC-FMS8 advanced SS7 HT-96 PE165 Cr12NiMo4 Cr13NiMo3 Cr15NiMo2 Cr16NiMo1 316SS
1
2
3
4
5
6
7
8
9
10
11
12
13
400 450 500 550 600
Mat
eria
l of
Bla
nke
t C
lad
din
g
Temperature of 1ry Hot Leg Coolant ℃
Grsph 8.2.2 Material of Blanket Cladding
13. ODS12. HT-911. PE-1610. EP-8239. 1.4970 SS 8. niobium7. Zr6. Cr17Ni135. Cr16Ni154. Cr16Ni113. Cr15Ni152. 12Crsteel1. 316SS
123456789
1011121314151617181920
400 450 500 550 600
Mat
eria
l of
Wra
pp
er
Temperature of 1ry Hot Leg Coolant ℃
Graph 8.2.3 Material of Wrapper
20 PNC-FMS 10. EP-82319 Cr13MnNb 9. 1.4970 SS 18 Advanced SS 8. niobium17 PE10 7. Zr16 14981SS 6. Cr17Ni1315 304 SS 5. Cr16Ni1514 18/8/1 SS 4. Cr16Ni1113. ODS 3. Cr15Ni1512. HT-9 2. 12Crsteel11. PE16 1. 316SS
60
8.3.1 Material of neutron absorber
Materials of neutron absorber, used for controlling the numbers of neutrons in the core,
are shown in graph 8.3.1.
B4C are used almost of all the plants for
neutron absorber.
8.3.2 Enrichment of B10 in neutron absorber B4s
Enrichments of B10, one of the isotope of Boron and having large neutron absorption
cross section, are shown in graph 8.3.2.
Their values scatter in wide range, but for large plants their value are in the range of
80～90%.
8.4 Reactor vessel
8.4.1 Material of reactor vessel
The relations between materials of reactor vessels and core outlet coolant
temperatures are shown in graph 8.4.1. For the materials of reactor vessel, SUS304,
SUS316 and 18Cr9NiSS are mainly used, and
moreover 1.6770SS.
3．Fuel Removal2. B B2O Er2O3 1．B4C
1
2
3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Mat
eri
al o
f A
bso
rbe
r
Thermal Power MW
Graph 8.3.1 Material of Neutron Absorber
1
2
3
4
5
6
7
8
340 360 380 400 420 440 460 480 500 520 540 560 580 600
Rea
cto
r V
esse
l Mat
eria
l
Temperature of Reactor Outlet Coolant ℃
Graph 8.4.1 Material of Reactor Vessel
8. 1.6770SS 7. 18Cr13Ni SS6. 18Cr12Ni SS5. 18Cr9Ni SS4. 18Ni8Cr SS3. 316SS2. 321SS1. 304SS
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500 4000 4500
B1
0En
rich
men
t%
Thermal Power MW
Graph 8.3.2 Enrichment oｆ B10
61
8.5 Primary piping
8.5.1 Material of primary hot leg piping
Materials of primary hot leg piping are shown in graph 8.5.1.
For the materials of primary hot leg piping,
SUS304, SUS316 and 18Cr9NiSS are mainly
used, and moreover 1.6770SS, just like these
of reactor vessels.
8.5.2 Connection between reactor vessel and primary piping
The relations between materials of reactor vessel and these of primary piping are
shown in graph 8.5.2. In this graph, numbers indicate the numbers adopted in plants.
Same materials are used in many plants, and they are 18Cr9NiSS, SUS304 and
1.6770SS.
8.5.3 Thickness of reactor vessel and hot leg primary piping
The relations between thickness of reactor vessel and that of primary hot leg piping
are shown in graph 8.5.3.
The ratios of them are about a few times, so
the stress investigations were held for the
each case.
8.6 Primary main pump
8. 1.6770SS 12Crsteel EP3127. 18Cr13Ni SS6. 18Cr12Ni SS5. 18Cr9Ni SS4. 18Ni8Cr SS3. 316SS2. 321SS1. 304SS
1
2
3
4
5
6
7
8
340 360 380 400 420 440 460 480 500 520 540 560 580 600
1ry
Pip
ing
Mat
eria
l
Temperature of Reactor Outlet ℃
Graph 8.5.1 Material of 1ry Piping
8. 1.6770SS 7. 18Cr13Ni SS6. 18Cr12Ni SS5. 18Cr9Ni SS4. 18Ni8Cr SS3. 316SS2. 321SS1. 304SS
1
2
3
4
5
6
7
8
1 2 3 4 5 6 7 8
1ry
Pip
ing
Mat
eri
al
Reactor Vessel Material
Graph 8.5.2 Reactor Vessel - 1ry Piping Material
5
6
3
4
1
1
1
1
1 1
1
0
2
4
6
8
10
12
14
16
18
0 10 20 30 40 50 60 70 80
1ry
Pip
ing
Thic
knes
sm
m
Reactor Vessel Thickness mm
Graph 8.5.3 Thickness of Reactor Vessel and 1ry Piping
62
8.6.1 Material of primary main pump
As the materials of parts in primary pump, shafts and hard facings of hydrostatic
bearing are shown in graph 8.6.1.1, impeller in graph 8.6.1.2 and the relations between
impeller and diffuser in graph 8.6.1.3.
8.7 Intermediate heat exchanger
8.7.1 Material of intermediate heat exchanger
Same materials are used for shell and heat
transfer tube of intermediate heat exchangers,
and they are shown in graph 8.7.1.
Many kinds of materials are used for them.
1
2
3
4
5
6
7
0 1 2 3 4 5 6 7 8 9 10 11
Mat
eria
l of
Bea
rin
g
Material of Shaft
Graph 8.6.1.1 Material of Main Pump Shaft and Hydrostatic Bearing
Shaft11 Cr12-steel10 Cr22Ni129 Cr16Ni108 special steel7 316 SS6 Em5 304 SS4 SS3 SCS132 Z15CNW22-121 Cr21Ni2W2.7
Hydrostatic bearing7 Cr12-steel6 SiC5 18/8/21SS4 stellite-123 stellite-62 stellite1 colomonoy
1
2
3
4
5
6
7
8
9
250 300 350 400 450 500 550
Mat
eri
al o
f Im
pe
ller
Temperature of Coolant at Pump ℃
Graph 8.6.1.2 Material of Pump Impeller
1
2
3
4
5
6
7
8
9
1 2 3 4 5 6 7 8 9
Mat
eria
l of
Pu
mp
Dif
fuse
r
Material of Pump Impeller
Graph 8.6.1.3 Material of Main Pump Impeller and Diffuser
Impeller/diffuser9 Cr12-steel8 Cr22Ni10MnSi7 Cr16Ni106 CF 35 304 SS4 SS3 SCS132 Z6CND 19-101 316 SS
10. Cr12Steel9. Cr16Ni118. 1.6770SS 12Crsteel EP3127. 18Cr13Ni SS6. 18Cr12Ni SS5. 18Cr9Ni SS4. 18Ni8Cr SS3. 316SS2. 321SS1. 304SS
63
8.8 Steam generator
8.8.1 Material of heat transfer tube of steam generator
Materials of heat transfer tubes of steam generators are shown in graph 8.8.1.1.
Those of evaporator, super-heater and/or re-heater are shown respectively in graph
8.8.1.2 and 8.8.1.3.
2 1/4Cr 1Mo steel are used both evaporator and super-heater of many steam
generators.
13. Cr21Ni3212. Cr18Ni911. Cr12 or 2 1/4Cr10. Cr12-steel9. Cr10Mo28. ９Cr 1Mo7. 2 1/4Cr 1Mo6. Ni33Cr21TiAlMn5. 1.6770SS4. Incoloy 8003. austenic SS2. 321 H SS1. 18/8/1 SS
1
2
3
4
5
6
7
8
9
10
11
12
13
280 300 320 340 360 380 400 420 440 460 480 500 520 540 560
SH a
nd
/or
RH
Mat
eria
l
Max.Temperature ℃
Graph 8.8.1.3 SH and/or RH
1
2
3
4
5
6
7
8
9
10
11
12
13
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Mat
eri
al o
f Tu
be
s
Thermal Power MW
Graph 8.8.1.1 Material of SG Tubes
EV
SH
RH
1
2
3
4
5
6
7
8
9
10
11
12
13
280 300 320 340 360 380 400 420 440 460 480 500 520 540 560
EVM
ater
ial
EV Max Temperature ℃
Graph 8.8.1.2 EV Temperature -Material of Tubes
1
2
3
4
5
6
7
8
9
10
300 350 400 450 500 550 600
Mat
eri
al
Temperature of Coolant at Pump ℃
IGraph 8.7.1 Material of IHX Shell and Tubes
64
8.8.2 Stress in heat transfer tube (super-heater only)
For super-heaters, the relations between the circumferential stress and kinds of
materials of heat transfer tubes are shown
in graph 8.8.2.
8.9 Secondary containment building
8.9.1 Material of secondary containment building
The relations between kinds of material and the volume of secondary containment
building are shown in graph8.9.1. But for three extra larger plants, there are no data of
volume, so, those data are not shown in the graph.
8.9.2 Design pressure of containment building
Design pressures of containment building are shown in graph 8.9.2. The data scatter
in wide range for small reactors, but those of large reactors are around 0.1 MPa.
8.9.3 Design of seismic acceleration of containment
Design values of seismic accelerations of containments are shown in graph 8.9.3.
0.00
0.05
0.10
0.15
0.20
0.25
0.30
0.35
0 50,000 100,000 150,000 200,000 250,000 300,000 350,000
Des
ign
Pre
ssu
re O
f C
on
tain
men
tM
Pa
Volume of Containment m3
Graph 8.9.2 Maximum Design Pressure of Containment
1 Steel2. Concrete
1
2
3
4
5
6
7
8
9
10
11
12
13
0 50 100 150 200 250 300
Mat
eria
l of
SHTu
bes
l
Stress of SH Tubes MPa
Graph 8.8.2 Material - Circumferencial Stress of SHTubes
1
2
0 50,000 100,000 150,000 200,000 250,000 300,000 350,000
Mat
eria
l of
Co
nta
inm
ent
Volume of Containment Building m3
Graph 8.9.1 Material of Secondary Containment Building
13. Cr21Ni3212. Cr18Ni911. Cr12 or 2 1/4Cr10. Cr12-steel9. Cr10Mo28. ９Cr 1Mo7. 2 1/4Cr 1Mo6. Ni33Cr21TiAlMn5. 1.6770SS4. Incoloy 8003. austenic SS2. 321 H SS1. 18/8/1 SS
65
These data seem to depend on the local
seismic conditions.
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0 50,000 100,000 150,000 200,000 250,000 300,000 350,000
Seis
mic
acc
eler
atio
ng
Volume of Containment m3
Graph 8.9.3 Seismic Acceleration of Containment
Vertical Seismic Acceleration g
Horizontal Seismic Acceleration g
66
9 Safety
9.1 Plant shutdown
9.1.1 Main Criteria for initiating automatic shutdown
Main criteria for initiating automatic shutdown are shown in graph 9.1.1.1.
Numbers of plants adopted for main
criteria are shown in graph 9.1.1.2.
Higher coolant temperature at reactor
outlet, higher reactivity, lower primary
coolant flow rate and higher neutron flux are
adopted in many plants.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Mai
n C
rite
ria
for
Sh
utd
ow
n
Thermal Power MW
Graph 9.1.1.1 Main Criteria for Initiating Automatic Shutdown
25 LSG Leakage of SG24 HR High radiation in containment23 EQ Earthquake22 DND Delayed neutron detection signal21 ABNS Acoustic boiling noise signal20 HRFI High rate-of-change of flow rate19 LSF Low 2ry coolant flow18 LEP Loss of electrical power17 LCL Low coolant level in reactor vessel16 LNF Low neutorn flux indication15 FiL Failure of I loops14 HF High neutron flux13 HD hydrogen detection12 TT Turbine trip11 CI Containment isolation demand10 LCI Low coolant level in IHX9 HCE High coolant level in pipe enclosure8 HPSS High pressure in 2ry coolant system7 HIT High 1ry coolant inlet temoerature6 CFI 1ry/2ry coolant flow imbalance5 HPF High ratio of 1ry coolant flow to core flux4 LPF Low ratio of 1ry coolant flow rate to core flux3 HRT High rate of coolant temperature change2 HRE High rate of flux chsnge(reactivity)1 HT High 1ry coolant outlet temperature
0
5
10
15
20
25
30
35
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25
Nu
mb
er o
f A
do
pte
d P
lan
ts
Main Criteria for ShutdownGraph 9.1.1.2 Main Criteria for Shutdown
(Total 35 Plants)
67
9.1.2 Principal plant shutdown system
Reactor shutdown systems are shown in graph 9.1.2.
Two redundant systems are adopted in many
reactors.
9.2 Reactor scram
9.2.1 Control rod
The values of worth of control rods are shown in graph 9.2.1.
9.2.2 Inserting stroke and rod-drop time of control rod
Inserting rod strokes are shown in graph 9.2.2.1, rod-drop times in graph 9.2.2.2.
Some of inserting rod strokes are higher than 1 meter, and these relate to the height of
core. The rod-drop times are almost shorter than 1 second.
The relations between inserting rod strokes and rod-drop times are shown in graph
9.2.2.3.
1
2
3
4
5
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Shu
tdo
wn
Sys
tem
Electric Power MW
Graph 9.1.2 Principal Shutdown System
0
2
4
6
8
10
12
14
16
0 2 4 6 8 10 12 14 16
Rea
ctiv
ity
Wo
rth
%△
K/K
Total Reactivity Worth %△K/K
Graph 9.2.1 Worth of Control Rods
Fine ControlRod Worth
Course Control Rod Worth
0
200
400
600
800
1000
1200
1400
0 200 400 600 800 1000 1200 1400
Stro
ke o
f F
ine
-Co
arse
Ro
ds
mm
Stroke of Safety Rods mm
Graph 9.2.2.1 Stroke of Control Rods
Stroke of Fine Rods mm
Stroke of coarse Rods mm
0
0.2
0.4
0.6
0.8
1
1.2
1.4
1.6
1.8
2
2.2
2.4
2.6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Ro
d D
rop
Tim
e se
c
Thermal Power MW
Graph 9.2.2.2 Rod Drop Time
68
9.2.3 Kinds and numbers of additional shutdown system
The reactor shutdown systems except inserting control rods are shown in graph
9.2.3. These systems are additionally used in only small numbers of reactors.
9.3 Decay heat removal
9.3.1 Safety features for coolant leakage
Keeping the coolant level for cooling the core in coolant leakage accident is particularly
excellent characteristics of FR. Here, these safety features are shown in graph 9.3.1.
In the plants, some of these safety features are actually used together.
1
2
3
4
5
6
7
8
9
10
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Safe
ty F
eat
ure
s
Thermal Power MW
Graph 9.3.1 Safety Features for Coolant Leakage
10 LG leak jacket9 IS isolation system with containment building8 PC collecting and cooling core meltdown debris7 NP no penetrations below sodium level6 EC elevation for natural cooling5 TOGV tank and piping guard vessel4 TGV reactor tank guard vessel3 CI containment isolation2 DW double wall1 EP elevated piping guard vessel
0
0.2
0.4
0.6
0.8
1
1.2
1.4
1.6
1.8
2
2.2
2.4
2.6
0 200 400 600 800 1000 1200 1400
Ro
d D
rop
Tim
ese
c
Stroke mm
Graph 9.2.2.3 Stroke - Rod Drop Time
0
2
4
6
8
10
12
14
16
18
840 392 700 2100 840 4000
Nu
mb
er o
f A
dd
itio
nal
Sh
utd
ow
n R
od
s
Thermal Power MW
Graph 9.2.3 Additional Shutdown Rods
HSR
USS
GEM
69
9.3.2 Emergency cooling system
Emergency cooling systems are adopted for core cooling in the case of emergency.
Their types of cooling system are shown in graph 9.3.2.1, their cooling power in graph
9.3.2.2, the ratio of their cooling power(except an example having 85% power) and
normal operation power in graph 9.3.2.3, the starting time for initiating operation in
graph 9.3.2.4, and the relation between their power and starting time in graph 9.3.2.5.
Actually, main cooling system with initiating the pump operation by pony motor
method, special heat removal exchanger method and natural circulation method are
adopted. Their power capacities are about 2 % of nominal power.
The times required initiating start of emergency cooling system are 30 minutes or
prompt starting just after accident in many plants.
9.3.3 Process after stopping primary pump
Processes after stopping primary pump are shown in graph 9.3.3.
4 Thermal Syphon Loops3 Natural Convection2 Special Heat Removal Loops1 Pony motors
1
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
De
cay
He
at R
emo
val S
yste
m
Thermal PowerMWGraph 9.3.2.1 Decay Heat Removal System
0
50
100
150
200
250
300
350
400
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Cap
acit
y fo
r D
ecay
Hea
t R
emo
val
MW
Thermal Power MW
Graph 9.3.2.2 Capacity for Emergency Removal of Decay Heat
0.00
0.02
0.04
0.06
0.08
0.10
0.12
0.14
0.16
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Dec
ay H
eat/
Full
Pow
er
Thermal Power MW
Graph 9.3.2.3 Ratio of Decay Heat/Nominal Power
0
10
20
30
40
50
60
70
80
90
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Del
ay b
efo
re O
per
atio
nm
in
Thermal Power MW
Graph 9.3.2.4 Delay before Operation in an Emergency Situation
0
10
20
30
40
50
60
70
80
90
0 50 100 150 200 250 300 350 400
De
lay
be
rore
Op
era
tio
nm
in
Capacity for Decay Heat Removal MW
Graph 9.3.2.5 Capacity - Delay berore Operation of Decay Heat Removal System
70
In many plants, automatic reactor scram
happen just after primary pump stopping,
and at the same time, pump driving power
change from the main motor to the pony
motor.
9.4 Detection of fuel failure and location of failed fuel
9.4.1 Detection of failed fuel
In case of an emergency of fuel failure, the detection methods are shown in graph 9.4.1.
9.4.2 Gas tagging method for detecting the location of failed fuel
One of the methods for detecting the location of failed fuel is gas tagging method.
The plants adopted the gas tagging method are shown in graph 9.4.2.1 for plant scale,
and in graph 9.4.2.2 for numbers of fuel elements (its maximum value is 212,502). In
these graphs, 1 indicates the plant having gas tagging method but 0 no method. But for
large scale plants this method does not adopted. For confirmation of this reason, the
relations between the gas tagging method and the starting time of plant construction
are shown in graph 9.4.2.3. The plant, its starting time is 1985, is the prototype reactor,
has 33,464 fuel elements and 714 MW thermal power. For larger plants than these
conditions, accuracy of gas tagging method and the fabrication of equipment is surmised
having the technical limitations.
7 DHR by natural circulation 6 power reduction and shut down5 power set back4 auxiliary power supply3 DHR by auxiliary EM pump 2 DHR by pony motor pump1 automatic scram
4 Wet Shipping Method3 Dry Shipping Method2 Gas Tagging1 Sodium Sampling
1
2
3
4
5
6
7
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Safe
ty M
eth
od
s
Thermal Power MWGraph 9.3.3 Plant Response with Seizureor Stopping
of 1ry Pump
1
2
3
4
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Met
ho
d o
f D
etec
tin
g Fa
iled
Pin
s
Thermal Power MW
Graph 9.4.1 Method of Detecting Failed Pins
0
1
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Gas
Tag
iing
Thermal Power MW
Graph 9.4.2.1 Gas Tagging - Thermal Power
0
1
0 50,000 100,000 150,000 200,000 250,000
Gas
Tag
gin
g
Number of Fuel Pins
Grsph 9.4.2.2 Gas Tagging - Number of Fuel Pins
71
9.5 Detection of coolant leakage
9.5.1 Detection method of sodium leakage
Detection methods of coolant sodium leakage are shown in graph 9.5.1.
Both electric contact method and aerosol detecting method are used in many plants.
For primary cooling leakage, radiation
detecting methods are adopted.
9.6 Water leakage in steam generator
9.6.1 Impurity in secondary coolant (under normal operation condition)
Concentrations of hydrogen and carbon in secondary coolant are shown in graph 9.6.1.
The values of hydrogen are lower than 0.5 ppm, and carbon scatter in range of 10～50
ppm.
9.6.2 Detective sensitivity and time loss of hydrogen concentration at water leakage
12 gas sampling11concrete temperature10 thermocouple9 radiation8 sodium-ion7 sodium fire6 sodium level5 H2 detector4 smoke3 aerosol2 electric contact1 conductivity
0
10
20
30
40
50
60
0
1
2
3
4
5
6
7
0 2 4 6 8 10 12
Car
bo
n Im
pu
rity
pp
m
Hyd
roge
n Im
pu
rity
pp
m
Oxygen Impurity ppm
Graph 9.6.1 Maximum Permissible Impurity Concentration in 2ry Coolant
Hydrogen 2ry
Carbon 2ry
1
2
3
4
5
6
7
8
9
10
11
12
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Leak
Det
ecti
on
Met
ho
d
Thermal Power MW
Garph 9.5.1 Method of Detection of Coolant Leaks
0
1
1950 1960 1970 1980 1990 2000 2010
Gas
Tag
gin
g
Start of Construction CY
Graph 9.4.2.3 Gas Tagging - Start of Construction
1 Gas Tagging0 none
72
The relations between detective sensitivity and time loss of hydrogen concentration at
water leakage in steam generator are shown in graph 9.6.2.
Detection sensitivities of hydrogen are very
low value about 0.01ppm, so, their detecting
time loss scatter in the range of longer 20 ～
120 seconds.
0
0.01
0.02
0.03
0.04
0.05
0.06
0 20 40 60 80 100 120 140
Hyd
roge
n Im
pu
rity
pp
m
Delay Time sec
Graph 9.6.2 Delay Time - Impurity of Hydrogen at Coolant leakage of SG
73
10 Management of operation
10.1 Operation Method
10.1.1 Plant operation method
Philosophies for plant operation are shown in following table.
These philosophies are composed from three elements, predicted as follows
Plant operation methods are shown in graph 10.1.1.1, power control specifications in
graph 10.1.1.2, and electric power supply method in graph 10.1.1.3.
Almost all the plants adopt automatic control operation method.
Targets of control are mainly temperature difference or temperature, but also flow rate
and power level remarkably.
Some plants have keeping constant power method, but grid following method is
attractive for large plants.
6 reactivitｙ5 power4 level3 flow2 temperature difference1 temperature
3 automatic2 manual and/or automatic1 manual
Experimental FR
Thermal Power40 58 40 120 140 60 55 62.5 200 400 8 65
Power Control 1 1 3 1 1 1 3 2 2 3 2 2
Control Specification 2 1,3 1,3 3 2 2
Electric Power Supply 2 1 1
Demonstration or Prototype FR
Thermal Power65 563 762 1250 714 670 975 750 1470 840 392 265 700
Power Control 2 1 2 1 3 2 3 3 3 2 2 2 2
Control Specification 2 2 2 5 2 5 3 1,3 1 4,5
Electric Power Supply 3 2 3
Commercial Sized FR
Thermal Power2990 3600 3420 1600 3800 4200 2100 3600 840 280 4000 2800 3530
Power Control 3 3 3 3 3 3 3 3 2 2 3 2 3
Control Specification 1 2 1,3 1,2,5 1,5 2 2 1 1 4 2 1 1,3
Electric Power Supply 1 3 3 1 3 3 3
1
2
3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Po
we
r C
on
tro
l
Thermal Power MW
Graph 10.1.1.1 Power Control
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Co
ntr
ol S
pe
cifi
cati
on
Thermal Power MW
Graph 10.1.1.2 Control Specification
74
10.1.2 Range for automatic power operation
For the range of automatic power operation, the upper of control range is naturally
100% of normal power and the lower are
shown in graph 10.1.2.
The lower range scatter 15～30%, but 25%
in many plants.
10.2 Mean length of reactor run (Operation period)
10.2.1 Mean length of routine shutdown for refueling
Mean length of reactor run are shown in graph 10.2.1.1 on days and in graph 10.2.1.2
on years, except one example having the period 2200 days (equal to 6 years) for visual
understanding of the graphs.
And mean length of reactor run, having the experience of construction, are shown in
graph 10.2.1.3. In this graph, one plant (SPX-1) has 1.753 years operation length, but all
other plants have operation length shorter than one year.
3 grid following2 load following1 constant
0
10
20
30
40
50
60
0 500 1000 1500 2000 2500 3000 3500 4000 4500Min
imu
m 1
ry P
um
p S
pe
ed
Co
ntr
ol
%
Thermal Power MW
Graph 10.1.2 Minimum Operating Range of 1ry Pump Speed Control
1
2
3
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Elec
tric
Po
wer
Su
pp
ly
Thermal Power MW
Graph 10.1.1.3 Electric Power Supply
0
100
200
300
400
500
600
700
800
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Re
acto
r R
un
Tim
e
day
s
Thermal Power MW
Graph 10.2.1.1 Mean Length of Reactor Run(days)
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
1.8
2.0
2.2
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Rea
cto
r R
un
Tim
e y
ears
Thermal Power MW
Graph 10.2.1.2 Mean Length of Reactor Run(years)
75
10.2.2 Rate of operation time (Net rate for working time)
Design rates of operation time, namely the ratio of operation time and total sum
operation time plus operation pausing time, are shown in graph 10.2.2.
Design aims the rate is much higher value
than about 85 %.
10.3 Exchange of fuel and others
10.3.1 Fuel residence time in core and refueling ratio
The relations of the residence time of fuel in inner core and in outer core are shown in
graph 10.3.1.1. It seems that their residence times are equivalent in almost all reactors.
So, the refueling ratios of fuel in inner core are shown in graph 10.3.1.2, in outer core
in graph 10.1.3. It seems that both exchange ratios are about 25 %.
The relations of refueling ratio between inner and outer fuel are shown in graph
10.3.1.4, and their ratio also seems to be equal.
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
1.8
1950 1960 1970 1980 1990 2000 2010
Act
ual
Re
acto
r R
un
Tim
e y
ear
s
Start of Construction CY
Graph 10.2.1.3 Mean Length of Reactor Run(Actual years)
0.60
0.65
0.70
0.75
0.80
0.85
0.90
0.95
1.00
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Ava
ilab
ility
of
Pla
nt
Op
erat
ion
Thermal Power MW
Graph 10.2.2 Ratio of (Reactor Pun)/(Reactor Run+Routine Shutdown)
0
500
1000
1500
2000
2500
3000
3500
0 500 1000 1500 2000 2500 3000 3500Res
iden
ce T
ine
of
Ou
ter
Co
re F
uel
d
ays
Residence Time of Inner core Fuel days
Graph 10.3.1.1 Mean Residence Time for Fuel
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500
Ref
ue
ling
Rat
io %
Residence Time days
Graph 10.3.1.2 Refueling Ratio of Inner Core Fuel
76
10.3.2 Exchanging ratio of control rod and blanket
Exchanging ratios of blankets and control rods are shown in graph 10.3.2.1 a nd
10.3.2.2.
10.4 Preheating
10.4.1 Method of preheating
Methods of preheating of primary system of loop type reactor are shown in graph
10.4.1.1, and of pool type in graph 10.4.1.2.
Electric heating methods are used in many loop type reactors, but gas heating and/or
electric heating methods in pool type reactors.
Preheating methods for secondary system are shown in graph 10.4.1.3, so electric
heating methods are used in all reactors.
6. Nitrogen +electric5 .steam4 .Argon3. Nitrogen2 .gas1. electric
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Pre
he
atin
g M
eth
od
Thermal Power MW
Graph 10.4.1.1 Preheating of 1ry of Loop type
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500
Ref
uel
ing
Rat
io%
Residence Time days
Graph 10.3.1.3 Refueling Ratio of Outer Core Fuel
0
10
20
30
40
50
60
70
80
90
100
0 10 20 30 40 50 60 70 80 90 100Ref
uel
ing
Rat
io o
f O
ute
r C
ore
Fu
el
%
Refueling Ratio of Inner Core Fuel %
Graph 10.3.1.4 Refueling Ratio ofInner - Outer Core Fuel
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500
Exch
ange
Rat
io o
f B
lan
ket
%
Residence Time of Inner Core Fuel days
Graph 10.3.2.1 Exchange Ratio of Blanket
0
10
20
30
40
50
60
70
80
90
100
0 500 1000 1500 2000 2500 3000 3500
Exch
ange
Rat
io o
f C
on
tro
l Ro
d%
Residence Time of Inner Core Fuel days
Graph 10.3.2.2 Exchange Ratio of Control Rod
77
10.4.2 Preheating temperature
Preheating temperatures of primary system are shown in graph 10.4.2.1, and it
seems their temperatures are about 150～200℃。
The ratios of preheating temperature of primary and secondary system are shown in
graph 10.4.2.2. Both temperatures are equal in small reactors, but those of secondary
are higher than primary in large reactors.
10.5 Purity of coolant
10.5.1 Maximum permissible impurity concentration of primary coolant
Maximum permissible impurity concentration of oxygen in primary coolant are shown
in graph 10.5.1.1. Their values are generally lower than 10 ppm.
Using the plugging temperature for indicating impurity is convenient for plant
0
2
4
6
8
10
12
14
0 500 1000 1500 2000 2500 3000 3500 4000
Imp
uri
ty o
f 1
ry O
xyge
n
pp
m
1ry Coolant Inventry ton
Graph 10.5.1.1 Maximum Permissible Impurity Concentration of 1ry Oxygen
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Pre
he
atin
g M
eth
od
Thermal Power MW
Graph 10.4.1.2 Preheating of 1ry of Pool Type
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
Pre
he
atin
g M
eth
od
Thermal Power MW
Graph 10.4.1.3 Preheating of 2ry System
0
50
100
150
200
250
300
350
400
450
0 500 1000 1500 2000 2500 3000 3500 4000 4500
1ry
Pre
he
atin
g Te
mp
era
ture
℃
Thermal Power MW
Graph 10.4.2.1 Preheating Temperature of 1ry System
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
1.6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
2ry
/1ry
Pre
he
atin
g Te
mp
era
ture
Thermal Power MW
Graph 10.4.2.2 Ratio of 2ry/1ry Preheating Temperature
130
135
140
145
150
155
160
0 500 1000 1500 2000 2500 3000 3500 4000
Plu
ggin
g Te
mp
era
ture
℃
1ry Coolant Inventry ton
Graph 10.5.1.2 Maximum Permissible Plugging Temperature of 1ry Coolant
78
operation. So, the plugging temperatures of primary coolant are shown in graph
10.5.1.2.
Eichelbarger’s equation is famous for the relation between the concentration of oxygen
and the plugging temperature, and its equation express that 10 ppm oxygen
concentration is equitable to 150 ℃ plugging temperature.
10.5.2 Maximum permissible impurity concentration of secondary coolant
Maximum permissible impurity concentrations of oxygen in secondary coolant are
shown in graph 10.5.2.1, and plugging temperatures in graph 10.5.2.2.
10.5.3 Maximum permissible impurity concentration of carbon and hydrogen in coolant
Maximum permissible impurity concentrations of hydrogen, and carbon in primary
coolant are shown in graph 10.5.3.
For those values in secondary coolant are
already shown in 9.6.1.
10.6 In-service inspection
10.6.1 Reactor vessel
Methods of in-service inspection of inner surface of reactor vessel and in-vessel
structures are shown in graph 10.6.1.1, and outer surface of reactor vessel in graph
10.6.1.2.
For inner surface of reactor vessel and in-vessel structures, under sodium and in gas
viewing methods and displacement monitoring methods are used. On the contrast with
those, it is possible to access to the outer surface of reactor vessel and little limitations
for measurements, many methods are adopted.
0
2
4
6
8
10
12
14
16
18
20
0 500 1000 1500 2000 2500 3000
Imp
uri
ty o
f 2
ry O
xyge
n p
pm
2ry Coolant Inventry ton
Graph 10.5.2.1 Maximum Permissible Impurity Concentration Of 2ry Oxygen
130
135
140
145
150
155
160
165
170
175
180
0 500 1000 1500 2000 2500 3000
plu
ggin
g Te
mp
era
ture
℃
2ry Coolant Inventry ton
Graph 10.5.2.2 Maximum Permissible Plugging Temperature of 2ry Coolant
0
1
2
3
4
5
6
7
0
10
20
30
40
50
60
0 2 4 6 8 10 12
Hyd
roge
n
pp
m
Car
bo
n
pp
m
Oxygen ppm
Graph 10.5.3 Maximum Impurity of Oxygen,Carcon,Hydrogen in 1ry Coolant
1ry Carbon
1ry Hydrogen
79
10.6.2 Circuit pipe of cooling system
Methods of in-service inspection of primary circuit pipes and secondary circuit pipes
are shown in graph 10.6.2.1 and 10.6.2.2.
The number indicated the method show as follows.
For primary circuit pipes of large plant, methods by visual inspection and ultrasonic
measurement are used, but for secondary circuit pipes, sodium leak monitoring are
adopted for inspection.
10 US ultrasonic measurements9 Elcon electrical contact8 EC eddy current measurements7 VI visual inspection by optical equipment6 AD aerosol detection of 1ry vessel leak5 FV free-moving vehicle4 TV tracked vehicle3 DM displacement monitoring by ultrasonic detection2 UGV under gas viewing(optical periscope)1 USV under-sodium viewing
6 VE volumetric examination5 LM sodium leak monitoring4 DM displacement measurements3 US ultrasonic measurements2 EC eddy current measurements1VI visual inspection
1
2
3
4
5
6
7
8
9
10
0 500 1000 1500 2000 2500 3000 3500 4000 4500
ISI -
Rea
cto
r V
esse
l an
d In
tern
al
Thermal PowerMW
Graph 10.6.1.1 In-Service Inspection Provisions for Reactor Vessel Inside and Internal Structure
1
2
3
4
5
6
7
8
9
10
0 500 1000 1500 2000 2500 3000 3500 4000 4500
ISI -
Re
acto
r V
ess
el O
ute
r Su
rfac
e
Thernal Power MW
Graph 10.6.1.2 In-Service Inspection Provisions for Reactor Vessel Outer Surface
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
ISI -
1ry
Pip
ing
Thermal Power MW
Graph 10.6.2.1 Inservice Inspection Provisions for 1ry Pipinng
1
2
3
4
5
6
7
0 500 1000 1500 2000 2500 3000 3500 4000 4500
ISI -
2ry
Pip
ing
Thermal Power MW
Graph 10.6.2.2 Inservice Inspection Provisionsfor 2ry Piping
7 XR X-rays6 MA manned access5 LM sodium leak monitoring4 SM smoke detector3 US ultrasonic measurements2 EC eddy current measurements1 VI visual inspection
80
10.6.3 Intermediate heat exchanger
Methods of in-service inspection of intermadiate heat exchanger are shown in graph
10.6.3.
Coolant leakage detection methods are used in
many plants, and also sodium level monitoring
and pressure change of argon gas method are
adopted.
10.6.4 Steam generator
Methods of in-service inspection of steam generator units are shown in graph 10.6.4.
Many methods are really used and more than
two methods are adopted in some plants.
12 AT accessibility to all tubes11 RVI regular visual inspection of tube-bores and structures10 LL lead level9 HD hydrogen detectors8 VM volumetric measurement7 EC eddy current measurements6 US ultrasonic measurements5 AP Argon pressure4 SM sodium leak monitoring3 SL sodium level2 LD leak detection1 VI visual inspection by optical equipment
1
2
3
4
5
6
0 500 1000 1500 2000 2500 3000 3500 4000 4500
ISI -
IHX
Pip
ing
Thermal Power MW
Graph 10.6.3 Inservise Inspection Provisions for IHX Piping
6 TR tube bundle can be removed for inspection 5 AP Argon pressure4 SM sodium leak monitoring3 SL sodium level2 LD leak detection1 VI visual inspection by optical equipment
1
2
3
4
5
6
7
8
9
10
11
12
0 500 1000 1500 2000 2500 3000 3500 4000 4500
ISI -
SG P
ipin
g
Thermal Power MW
Graph 10.6.4 Inservice Inspectio Provisions for SG Piping
81
Literature
In this document, all graphs are made by using the data of the following IAEA
technical document.
IAEA-TECDOC-1531 ‘Fast Reactor Database 2006 Update’
IAEA 2006
82
Index
For searching your required graph, find the keyword arranged by the order of
alphabet, and check the items of X-axis and Y-axis, then look for the graph by the its
graph number.
No. Key Word Graph
No.
Horizontal X Axis Vertical Y Axis
1 Absorber Enrichment 8.3.2 Thermal Power B10 Enrichment
2 Absorber Material 8.3.1 Thermal Power Material of Absorber
3 Blanket 2.3.7 Thermal Power Blanket Diameter
4 Blanket Cladding Material 8.2.2 Temperature of 1ry Hot Leg
Coolant
Material of Blanket Cladding
5 Breeding Gain 1 5.6.1.1 Thermal Power Total Breeding Gain
6 Breeding Gain 2 5.6.1.2 Fuel Pellet Diameter Total Breeding Gain
7 Breeding Gain 3 5.6.1.3 Outer Diameter of Core Fuel
Pin
Total Breeding Gain
8 Breeding Gain 4 5.6.2 Total Breeding Gain Number of Axial Blanket
9 Breeding Gain 5 5.6.3.1 Thermal Power Equivalent Row Number of
Upper Axial Blanket
10 Breeding Gain 6 5.6.3.2 Thermal Power Equivalent Row Number of
Lower Axial Blanket
11 Burnup 1 5.7.1.1 Thermal Power Averaged Burnup
12 Burnup 2 5.7.1.2 Thermal Power Max./Average Burnup
13 Cladding Diameter 1 2.1.2.1 Thermal Power Cladding Diameter
14 Cladding Diameter 2 2.1.2.2 Electric Power Cladding Diameter
15 Cladding Material 3 8.1.4 Circumferential Stress Material of Cladding
16 Cladding Material 1 8.1.1.1 Maximum Temperature of
Cladding
Cladding Material
17 Cladding Material 2 8.1.1.2 Date of Start of Construction Cladding Material
18 Cladding Temperature 1 6.3.2 Maximum Coolant Velocity Maximum Cladding
Temperature
19 Cladding Temperature 2 6.3.3 Maximum Linear Power Maximum Cladding
Temperature
20 Cladding Temperature 3 6.3.4 Pellet Diameter Maximum Cladding
Temperature
21 Cladding Temperature 4 6.3.5 Cladding Diameter Maximum Cladding
Temperature
22 Cladding Thickness 1 2.1.2.3 Thermal Power Cladding Thickness
23 Cladding Thickness 2 2.1.2.4 Electric Power Cladding Thickness
24 Cladding Thickness 3 2.1.2.5 Cladding Diameter Cladding Thickness
25 Cladding Thickness/Diameter 2.1.2.6 Cladding Diameter Ratio of Cladding
Thickness/diameter
26 Cold Trap Coolant 4.3.1.3 1ry Cold Trap Coolant 2ry Cold Trap Coolant
83
27 Cold Trap Number 4.3.1.1 1ry Cold Trap 2ry Cold Trap
28 Cold Trap Volume 4.3.1.2 1ry Mesh Volume 2ry Mesh Volume
29 Condenser Pressure 3.5.4 Electric Power Minimum Pressure
30 Containment Material 8.9.1 Volume of Containment
Building
Material of Containment
31 Containment Pressure 8.9.2 Volume of Containment Operation Pressure of
Containment
32 Containment Secondary 4.2.1 Volume of Containment
Building
Geometry of Containment
33 Containment Seismic Acceleration 8.9.3 Volume of Containment Seismic Acceleration
34 Containment/Reactor Vessel Volume
Ratio
4.2.2 Electric Power Loop and Pool Type Ratio
35 Control Rod Element Number 1 2.3.11.1 Number of Coarse Rod
Element
Diameter of Coarse rod
36 Control Rod Element Number 2 2.3.11.2 Number of Fine Rod Element Diameter of Fine rod
37 Control Rod Element Number 3 2.3.11.3 Number of Safety Rod
Element
Diameter of Safety rod
38 Control Rod Kind Number 2.3.13 Thermal Power Nuｍber of Control Rod
39 Control Rod Subassembly Number 1 2.3.12.1 Thermal Power Number of Control Rod
40 Control Rod Subassembly Number 2 2.3.12.2 Thermal Power Number of Rod
41 Coolant Kind 1 1.2.2.1 Construction Start 1ry Coolant Kind
42 Coolant Kind 2 1.2.2.2 Thermal Power 1ry Coolant Kind
43 Coolant Pressure 1 6.1.2.1 Thermal Power Core Inlet Coolant Pressure
44 Coolant Pressure 2 6.1.2.2 Thermal Power Core Outlet Coolant Pressure
45 Coolant Pressure 3 6.1.3 Thermal Power Pressure Drop in Core
46 Coolant Pressure 4 7.3.1.2 Thermal Power SH Outlet Pressure
47 Coolant Pressure-temperature 7.3.1.3 SH Outlet Steam temperature SH Outlet Steam Pressure
48 Coolant Purity 1 10.5.1.1 1ry Coolant Inventory Impurity of 1ry Oxygen
49 Coolant Purity 2 10.5.1.2 1ry Coolant Inventory Plugging Temperature
50 Coolant Purity 3 10.5.2.1 2ry Coolant Inventory Impurity of 2ry Oxygen
51 Coolant Purity 4 10.5.2.2 2ry Coolant Inventory Plugging Temperature
52 Coolant Purity 5 10.5.3 Oxygen Impurity Hydrogen, Carbon Impurity
53 Coolant Temperature 1 6.2.1 Thermal Power Maximum 1ry Coolant
Temperature
54 Coolant Temperature 1 7.1.1 Thermal Power Coolant Temperature
55 Coolant Temperature 10 7.3.1.1 Thermal Power SH Outlet Temperature
56 Coolant Temperature 11 7.3.2 Thermal Power Temperature Difference of
Steam and Water
57 Coolant Temperature 2 6.2.2 Maximum 1ry Coolant
Temperature
Maximum Linear Power
58 Coolant Temperature 2 7.1.2.1 Thermal Power 1ry Coolant Temperature at
IHX Inlet
59 Coolant Temperature 3 7.1.2.2 Thermal Power Steam Temperature at Turbine
Inlet
60 Coolant Temperature 4 7.1.3 1ry Coolant Flow Rare 1ry Hot-Cold Temperature
84
Difference
61 Coolant Temperature 5 7.1.4 1ry Coolant Temperature
Difference
2ry Coolant Temperature
Difference
62 Coolant Temperature 6 7.1.5.1 Reactor Outlet Temperature Turbine Inlet Temperature
63 Coolant Temperature 7 7.1.5.2 Thermal Power Reactor Outlet-Turbine Inlet
Temperature
64 Coolant Temperature 8 7.2.1 Thermal Power Coolant Temperature of IHX
65 Coolant Temperature 9 7.2.2 IHX 1ry Coolant Temperature IHX 2ry Coolant Temperature
66 Coolant Velocity 6.1.1 Thermal Power Maximum Coolant Velocity
67 Cooling Flow Rate 1 3.1.2.1 Thermal Power Flow rate of 1ry Coolant
68 Cooling Flow Rate 2 3.1.2.2 Thermal Power Flow rate of 2ry Coolant
69 Cooling Flow Rate 3 3.1.2.3 Thermal Power 1ry/2ry Coolant Flow rate
70 Cooling Piping 1 3.1.3.1 Thermal Power 1ry Piping Diameter
71 Cooling Piping 10 3.1.4.6 Thermal Powert 3rd Piping Thickness/Diameter
72 Cooling Piping 11 3.1.5.1 1ry Piping Diameter 1ry Piping Structure
73 Cooling Piping 12 3.1.5.2 1ry Piping Diameter 1ry Piping Sodium Leak
Protection
74 Cooling Piping 13 3.1.5.3 2ry Piping Diameter 2ry Piping Sodium Leak
Protection
75 Cooling Piping 2 3.1.3.2 Thermal Power 2ry Piping Diameter
76 Cooling Piping 3 3.1.3.3 Thermal Power 3rd Piping Diameter
77 Cooling Piping 4 3.1.3.4 Thermal Power Diameter of Piping
78 Cooling Piping 5 3.1.4.1 1ry Piping Diameter Hot Leg Piping Thickness
79 Cooling Piping 6 3.1.4.2 2ry Piping Diameter Hot Leg Piping Thickness
80 Cooling Piping 7 3.1.4.3 3ry Piping Diameter Hot Leg PipingT hickness
81 Cooling Piping 8 3.1.4.4 Thermal Power 1ry Piping Thickness/Diameter
82 Cooling Piping 9 3.1.4.5 Thermal Power 2ry Piping Thickness/Diameter
83 Cooling System Number 1 3.1.1.1 Thermal power 1ry cooling Loop Number
84 Cooling System Number 2 3.1.1.2 Thermal power 2ry cooling Loop Number
85 Cooling System Number 3 3.1.1.3 1ry Cooling System Number 2ry cooling Loop Number
86 Cooling Valve 3.1.6 Thermal Power Valve
87 Core Height 2.1.6.2 Core Height Length of Fuel Element
88 Core Size 1 2.3.4.1 Thermal Power Diameter of Core
89 Core Size 2 2.3.4.2 Thermal Power Height of Core
90 Core Size 3 2.3.4.3 Thermal Power Height/Diameter
91 Core Size 4 2.3.4.4 Thermal Power Core Volume
92 Core Volume Fraction 1 5.1.1.1 Thermal Power(Plant) Core Volume Fraction
93 Core Volume Fraction 2 5.1.1.2 Thermal Power Core Volume Fraction
94 Core Volume Fraction 3 5.1.1.3 Thermal Power Fuel/Coolant Volume Fraction
95 Core Zone 1 2.3.5.1 Thermal Power Diameter of Inner Core
96 Core Zone 2 2.3.5.2 Thermal Power Ratio of Diameter inner/outer
Core
97 Core Zone Number 2.3.3 Thermal Power Number of Zone
98 Cover Gas 1 2.3.15.1 Thermal Power Kind of 1ry Cover Gas
85
99 Cover Gas 2 2.3.15.2 Thermal Power Kind of 2ry Cover Gas
100 Cover Gas 3 2.3.16.1 Thermal Power 1ry Cover Gas Pressure
101 Cover Gas 4 2.3.16.2 1ry Cover Gas Pressure 2ry Cover Gas Pressure
102 Enrichment Plutonium 1 5.5.1.1 Thermal Power Pu Enrichment
103 Enrichment Plutonium 2 5.5.1.2 Thermal Power Ratio of Outer/Inner Pu
Enrichment
104 First Criticality 1 1.4.1 First Criticality 1ry Circuit Configuration
105 First Criticality 2 1.4.2 First Criticality Thermal Power
106 First Electricity Generation 1.4.3.1 First Electricity Generation Electric Power
107 First Full Power Generation 1.4.3.2 First Full Power Generation Electric Power
108 FP Gas 1 2.1.8.1 Maximum Fuel Burnup FP Gas Volume
109 FP Gas 1 8.1.2.1 Averaged Fuel Burnup FP Gas Pressure
110 FP Gas 2 2.1.8.2 Fuel Element Length FP Gas Plenum Length
111 FP Gas 2 8.1.2.2 Maximum Burnup Cladding Thickness
112 FP Gas 3 8.1.3.1 Thermal Power Circumferential Stress
113 FP Gas 4 8.1.3.2 Maximum Cladding
Temperature
Circumferential Stress
114 FP Gas 5 8.1.3.3 Mean Residence Time in Core Circumferential Stress
115 Fuel Kind 1 1.2.1.1 Construction Start Fuel Kind
116 Fuel Kind 2 1.2.1.2 Thermal Power Fuel Kind
117 Fuel Spacer 2.2.2 Thermal Power Type of Fuel Pin Separation
118 Fuel and Blanket Length 2.1.7 Plant Thermal Power Fuel and Blanket Length in
Element
119 Fuel Density 2.1.3.1 Thermal Power Density of fuel
120 Fuel Density Smeared 2.1.3.2 Thermal Power Smeared Density of fuel
121 Fuel Element Component Length 2.1.9 Average Fuel Burnup Ratio of Composition Length
122 Fuel Element Length 2.1.6.1 Thermal Power Length of Fuel Element
123 Fuel Element Length/Core height 2.1.6.3 Thermal Power Ratio of Fuel Element
Length/Core Height
124 Fuel Element Number 2.2.4 Thermal Power Number of Fuel Element
125 Fuel/Blanket 1 2.3.8.1 Diameter of Fuel Diameter of Blanket
126 Fuel/Blanket 2 2.3.8.2 Thermal Power Blanket/Fuel Diameter
127 Fuel/Blanket 3 2.3.8.3 Thermal Power Number of Pins
128 Fuel/Blanket 4 2.3.8.4 Thermal Power Fuel/Blanket Pins Number
129 Fuel/Control Rod 2.3.12.3 Thermal Power Fuel/Control rod
130 IHX Area 3.3.2.2 IHX Heat Transfer Capacity IHX Heat Transfer Area
131 IHX capacity 3.3.2.1 Thermal Power IHX Heat Transfer Capacity
132 IHX Flow Side 3.3.1.1. Thermal Power IHX 1ry Flowing Side
133 IHX Material 8.7.1 Temperature of Coolant at
Pump
Material of IHX Shell and Tube
134 IHX Number 3.3.1.2. Thermal Power Number of IHX
135 IHX Shell 1 3.3.4.1 IHX Heat Transfer Capacity IHX Shell Diameter
136 IHX Shell 2 3.3.4.2 IHX Shell Diameter IHX Shell Thickness
137 IHX Tube 1 3.3.3.1 Thermal Power IHX Heat Transfer Tube
86
Diameter
138 IHX Tube 2 3.3.3.2 IHX Heat Transfer Capacity IHX Heat Transfer Tube
Thickness/Diameter
139 IHX Tube 3 3.3.3.3 IHX Heat Transfer Capacity IHX Heat Transfer Tube Length
140 IHX Tube 4 3.3.3.4 IHX Heat Transfer Area IHX Heat Transfer Tube
Number
141 In-Service Inspection 1 10.6.1.1 Thermal Power ISI Reactor Vessel and Internal
142 In-Service Inspection 2 10.6.1.2 Thermal Power ISI Reactor Vessel Outer
Surface
143 In-Service Inspection 3 10.6.2.1 Thermal Power ISI 1ry Piping
144 In-Service Inspection 4 10.6.2.2 Thermal Power ISI 2ry Piping
145 In-Service Inspection 5 10.6.3 Thermal Power ISI IHX Pipe
146 In-Service Inspection 6 10.6.4 Thermal Power ISI SG Pipie
147 Linear Power 1 5.3.1.1 Thermal Power Maximum Linear Power
148 Linear Power 2 5.3.1.2 Thermal Power Maximum/Mean Linear Power
149 Linear Power, Neutron Flux, Power
Density
5.4.1.1 Thermal Power Neutron Flux, Linear Power,
Power Density
150 Linear Power/Neutron Flux 5.3.2. Max Neutron Flux Maximum Linear Power
151 Neutron Flux 1 5.2.1.1 Thermal Power Maximum Neutron flux
152 Neutron Flux 2 5.2.1.2 Thermal Power Maximum/Mean Neutron flux
153 Pellet Diameter 2.1.5.1 Thermal Power Diameter of Pellet
154 Pellet/Cladding Diameter 2.1.5.2 Thermal Power Ratio of Pellet/Cladding
Diameter
155 Piping Material 8.5.1 Temperature of Reactor
Outlet Coolant
1ry Piping Material
156 Plant Classification 0.1.1 Plant Name
157 Plant Classification 1 1.1.2 Thermal Power Plant Classification
158 Plant Classification 2 1.1.3 Electric Power Plant Classification
159 Plant Control 1 10.1.1.1 Thermal Power Power Control
160 Plant Control 2 10.1.1.2 Thermal Power Control Specification
161 Plant Control 3 10.1.1.3 Thermal Power Electric Power Supply
162 Plant Control 4 10.1.2 Thermal Power Minimum 1ry Pump Speed
163 Plant Development 1 1.5.1.1 Thermal Power Nation
164 Plant Development 2 1.5.1.2 Electric Power Nation
165 Plant Development 3 1.5.2.1 First Criticality Thermal Power by Nation
166 Plant Development 4 1.5.2.2 First Electricity Generation Electric Power by Nation
167 Plant Development 5 1.5.3 Nation Ratio of Plant Power
168 Plant Major Event 1.4.4 Date of Major Event Major Event
169 Plant Operation Method 10.1.1 Plant Operatin Method
170 Plant Operation 1 10.2.1.1 Thermal Power Reactor Run Time(day)
171 Plant Operation 2 10.2.1.2 Thermal Power Reactor Run Time(year)
172 Plant Operation 3 10.2.1.3 Start of Construction Actual Reactor Run Time
173 Plant Operation 4 10.2.2 Thermal Power Availability Plant Operation
174 Plant Operation Time 1 1.4.5 Plant Name Term Years
87
175 Plant Operation Time 2 1.4.6 Classification Plant
Operation Term
Reactor years
176 Plant Power 0.2.1 Plant Power
177 Plant Size 1 0.2.2 Plant Name Thermal Power
178 Plant Size 2 0.2.3 Plant Name Electric Power
179 Power Density 5.4.1.2 Thermal Power Max/Mean Power Density
180 Preheating 1 10.4.1.1 Thermal Power Preheating Method of Loop 1ry
181 Preheating 2 10.4.1.2 Thermal Power Preheating Method of Pool 1ry
182 Preheating 3 10.4.1.3 Thermal Power Preheating Method of 2ry
183 Preheating 4 10.4.2.1 Thermal Power 1ry Preheating Temperature
184 Preheating 5 10.4.2.2 Thermal Power 2ry/1ry Preheating
Temperature
185 Primary circuit Configuration 1 1.3.1.1 Thermal Power 1ry Circuit Configuration
186 Primary circuit Configuration 2 1.3.1.2 Electric Power 1ry Circuit Configuration
187 Primary circuit Configuration 3 1.3.1.3 Construction Start 1ry Circuit Configuration
188 Pump Head 1 3.2.2.2 1ry Pump Capacity Pump Head
189 Pump Head 2 3.2.4 2ry Pump Capacity 2ry Pump Head
190 Pump Location 3.2.1 Thermal Power Location of Pump
191 Pump Material 1 8.6.1.1 Material of Pump Hard
Facing
Material of Pump Shaft
192 Pump Material 2 8.6.1.2 Temperature of Coolant at
Pump
Material of Pump Impeller
193 Pump Material 3 8.6.1.3 Matrial of pump Shsft Material of Pump Diffuser
194 Pump Power 1 3.2.3.1 1ry Pump Capacity Electric Power Input
195 Pump Power 2 3.2.3.2 Thermal Power Ratio of Pump Electric Power
Input
196 Pump Speed 3.2.2.3 1ry Pump Capacity Pump Speed
197 Pump Speed Control 3.2.2.4 1ry Pump Capacity Principle of 1ry Pump Speed
Control
198 Pump Type 3.2.2.1 1ry Pump Capacity Type of Pump
199 Reactivity Coefficient 1 5.8.1.1 Thermal Power Reactivity Coefficient
200 Reactivity Coefficient 2 5.8.1.2 Thermal Power Maximum Coolant Void
Coefficient
201 Reactivity Coefficient 3 5.8.2 Thermal Power Doppler Coefficient
202 Reactor Vessel 1 2.3.14.1 Thermal Power Inner Diameter of Reactor
Vessel
203 Reactor Vessel 2 2.3.14.2 Inner Diameter of Reactor
Vessel
Height of Reactor Vessel
204 Reactor Vessel 3 2.3.14.3 Inner Diameter of Reactor
Vessel
Thickness of Reactor Vessel
205 Reactor Vessel Material 8.4.1 Temperature of Reactor
Outlet Coolant
Reactor Vessel Material
206 Reactor Vessel-Piping Material 8.5.2 Reactor Vessel Material 1ry Piping Material
207 Reactor Vessel-Piping Thickness 8.5.3 Reactor Vessel Thickness 1ry Piping Thickness
208 Refueling 1 4.1.1 Electric Power Refueling Method
88
209 Refueling 2 10.3.1.1 Residence Time of Inner Core
Fuel
Residence Time of Outer Core
Fuel
210 Refueling 3 10.3.1.2 Residence Time Refueling Ratio
211 Refueling 4 10.3.1.3 Residence Time Refueling Ratio
212 Refueling 5 10.3.1.4 Refueling Ratio of Inner Core
Fuel
Refueling Ratio of Outer Core
Fuel
213 Refueling 6 10.3.2.1 Refueling Ratio of Inner Core
Fuel
Exchange Ratio of Blanket
214 Refueling 7 10.3.2.2 Refueling Ratio of Inner Core
Fuel
Exchange Ratio of Control Rod
215 Safety Control Rod 1 9.2.2.1 Stroke of Control Rod Stroke of Fine-Coarce Rod
216 Safety Control Rod 2 9.2.2.2 Thermal Power Rod Drop Time
217 Safety Control Rod 3 9.2.2.3 Stroke of Control Rod Rod Drop Time
218 Safety Control Rod 4 9.2.3 Thermal Power Number of Additional
Shutdown Rod
219 Safety Coolant Leakage 1 9.3.1 Thermal Power Safety Feature for Coolant
Leakage
220 Safety Coolant Leakage 2 9.5.1 Thermal Power Leak Detection Method
221 Safety Decay Heat Removal 1 9.3.2.1 Thermal Power Decay Heat Removal System
222 Safety Decay Heat Removal 2 9.3.2.2 Thermal Power Capacity for Decay Heat
Removal
223 Safety Decay Heat Removal 3 9.3.2.3 Thermal Power Decay Heat/Full Power
224 Safety Decay Heat Removal 4 9.3.2.4 Thermal Power Delay before Operation
225 Safety Decay Heat Removal 5 9.3.2.5 Capacity of Decay Heat
Removal
Delay before Operation
226 Safety Decay Heat Removal 6 9.3.3 Thermal Power Safety Method
227 Safety Detection Failed pin 1 9.4.1 Thermal Power Method of Detecting Failed Pin
228 Safety Detection Failed pin 2 9.4.2.1 Thermal Power Gas Tagging
229 Safety Detection Failed pin 3 9.4.2.2 Number of Fuel Pin Gas Tagging
230 Safety Detection Failed pin 4 9.4.2.3 Start of Construction Gas Tagging
231 Safety Reactor Scram 9.2.1 Total Reactivity Worth Reactivity Worth
232 Safety Shutdown 1 9.1.1.1 Thermal Power Main Criteria for Shutdown
233 Safety Shutdown 2 9.1.1.2 Main Criteria for Shutdown Number of Adopted Plant
234 Safety Shutdown 3 9.1.2 Electric Power Shutdown System
235 Safety Water Leakage 9.6.1 Oxygen Impurity Hydrogen, Carbon impurity
236 Safety Water Leakage 9.6.2 Delay Time Hydrogen Impurity
237 SG Configuration 3.4.1 Thermal Power Configuration of SG
238 SG Heat Capacity 3.4.6 SG Heat Capacity EV SH RH Heat Capacity
239 SG Material 1 8.8.1.1 Thermal Power Material of SG Tube
240 SG Material 2 8.8.1.2 Ev Maximum Temperature EV Material
241 SG Material 3 8.8.1.3 SH/RH Maximum
Temperature
SH/RH Material
242 SG Material 4 8.8.2 stress of SH Tube Material of SH tube
243 SG Number 1 3.4.2.1 Thermal Power SG Number
244 SG Number 2 3.4.2.2 Thermal Power SG Number
89
245 SG Tube 1 3.4.4.1 Thermal Power EV Tube Diameter and
Thickness
246 SG Tube 2 3.4.4.2 Thermal Power SH Tube Diameter and
Thickness
247 SG Tube 3 3.4.4.3 Thermal Power RH Tube Diameter and
Thickness
248 SG Tube 4 3.4.4.4 EV Tube Diameter EV Tube Thickness/Diameter
249 SG Tube 5 3.4.4.5 SH Tube Diameter SH Tube Thickness/Diameter
250 SG Tube 6 3.4.4.6 RH Tube Diameter RH Tube Thickness/Diameter
251 SG Tube 7 3.4.4.7 EV Tube Diameter SH/EV Tube Diameter
252 SG Tube 8 3.4.4.8 EV Tube Diameter RH/EV Tube Diameter
253 SG Tube NUmber 3.4.5 SG Heat Capacity Number of Tube
254 SG Tube Type 3.4.3 Thermal Power Type of SG Tube
255 Sodium Inventory 1 3.1.7.1 Thermal Power Coolant Inventory
256 Sodium Inventory 2 3.1.7.2 1ry Coolant Inventory 2ry Coolant Inventory
257 Spacer Material 8.2.1 Maximum Temperature of
Cladding
Spacer Material
258 Spent Fuel Store 4.1.2 Thermal Power Spent Fuel Storing Method
259 Spent Fuel Transport 4.1.3 Thermal Power Fuel Transfer Method
260 Subassembly Clearance 1 2.3.2.1 Width across Subassembly Clearance between
Subassembly
261 Subassembly Clearance 2 2.3.2.2 Width across Subassembly Clearance/Width
262 Subassembly Length 2.2.3.1 Thermal Power Length of Subassembly
263 Subassembly Number 1 2.3.6.1 Thermal Power Number of Subassembly
264 Subassembly Number 1 2.3.9.1 Thermal Power Number of Subassembly
265 Subassembly Number 2 2.3.6.2 Thermal Power Ratio of Inner/Outer
266 Subassembly Number 2 2.3.9.2 Thermal Power Number of Subassembly(a part)
267 Subassembly Width 1 2.2.5.1 Thermal Power Width across Flat
268 Subassembly Width 2 2.2.5.2 Number of Fuel Element Width across Flat
269 Subassembly/Fuel Element Length 1 2.2.3.2 Thermal Power Ratio of Subassembly/Fuel
Element Length
270 Subassembly/Fuel Element Length 2 2.2.3.3 Height of Core Subassembly and Fuel Element
Length
271 Thermal Efficiency 1 7.4.1 Electric Power Thermal Efficiency
272 Thermal Efficiency 2 7.4.2.1 Turbine Inlet Steam
temperature
Thermal Efficiency
273 Thermal Efficiency 3 7.4.2.2 Reactor Outlet Temperature Thermal Efficiency
274 Thermal Efficiency 4 7.4.3.1 Reactor Outlet-Turbine Inlet
Temperature
Thermal Efficiency
275 Thermal Efficiency 5 7.4.3.2 Turbine Inlet -Feed Water
Temperature
Thermal Efficiency
276 Turbine Number 3.5.1 Electric Power Number of Turbine
277 Turbine Speed 3.5.2 Electric Power Speed of Turbine
278 Turbine Steam Condition 3.5.3 Steam Temperature Steam Pressure
279 Weight Pu/U 1 2.3.10.1 Thermal Power Total Pu Weight
90
280 Weight Pu/U 2 2.3.10.2 Pu239 Weight U235 Weight
281 Wrapper Material 8.2.3 Temperature of 1ry Hot Leg
Coolant
Material of Wrapper
