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March 2013
Fast Breeder Reactor Research and Development
Center (Monju)
Japan Atomic Energy Agency
Takehide Deshimaru
RECENT PROGRESS AND STATUS
OF MONJU
�Introduction
�Historical perspectives
� Restart of system start-up tests
� Recovery from Hardware Troubles
�Post-Fukushima Safety Measures
� Safety characteristics and measures
� Comprehensive safety assessment
�National Debates on NuclearProgram and Monju Research Plan
�Concluding Remarks
1CONTENTS
Sodium leak detection
and monitoring system
Modification of 2ry
sodium piping
History of Monju
Dec.1995 Sodium leak accident
2005-2007 Plant modification
to improve sodium safety
Aug.1995 First grid
Apr. 1994 Criticality
May 2010 Restart of SST-1
Jul. 2010 Completion of SST-1
A future research plan and schedule of Monju is
under discussion in the MEXT WG, with its
report being expected in summer 2013.
2
Beyond 45 % reactor output, steam is introduced to the superheaters, and the turbine is operated with superheated steam.
1) Start up systems on a step-by-step basis from reactor, turbine, generator, etc. to confirm the overall plant performance.
2) The SST consists of the three steps of tests.
Core ConfirmationTest (CCT)
Inspection
40%-power Confirmation Test (40%CT)
Inspection, evaluation and confirmation
Power RisingTest (PRT)
Refueling
Refueling
Core characteristics check at 0%*1 power output
Entire plant functions and performances are checked including water/steam and turbine-generator systemsat 0 – 40% power output.
Overall plant performance is checked at 0 – 100% power output
Reactor output (RO)
Electricity output (EO)Reactor, Primary Na loop, Secondary Na loop
Water/Steam System – includes the by-pass system for start-up
Turbine - Generator
EO:40%
RO:45% 45%
40%
79%
75%
100%
RO: 0%*1
*1 : The precise power ranges in 0.001 – 1.3% .
CCT core An expected core configuration for 40%CT
An expected core configuration for PRT
Initial core
: loaded fuel
: fresh fuelRefueling
(84 core fuels)
100%RO
EO
Overview of System Startup Tests 3
Achievement of Core Confirmation Test
� Safe startup and operation of the reactor and cooling system
� Reactor core with 14-year-old fuel and some new fuelStartup and operation
� Reactivity worth of all the 19 control rods
� Safe control and shutdown of the reactorSafe control of reactor
Inherent self-stability � Negative reactivity feedback characteristics
� Inherent self-stability upon power increase
� Complex reactor core composition with three different
types of fuel subassemblies including Am-rich 14-year-old
fuel
Accurate prediction of criticality
New technologies � Basic physics studies in collaboration with universities
� Test with an advanced ultrasonic thermometer
Reactor physics data
Major achievement
� Valuable reactor physics data with the fuel containing about
1.5% americium
� Successful operation, after a long blank for more than 14 years, with no major troubles
� Extremely valuable data with a complicated fuel composition
SST-1 (Core confirmation test) successfully conducted
4
Recovery from hardware troubles 5
� Hardware troubles in recent years, “a drop of in-vessel transfer machine (IVTM) (Aug. 2010)”,
“cracking in cylinder liner of DG (Dec. 2011) and other minor troubles, have all been restored.
� The trouble with IVTM took nearly two years to completely bring the plant back to normal state.
Reactor vessel
Core
Auxiliary Handling Machine (AHM)
Cross-section view of reactor vessel when the IVTM is hung up
Dropped IVTM
Schematic view of the IVTM
Opening andclosing rod
AHMgripper
90mm
Auxiliary Handling Machine (AHM) gripper failed to fully open, due to rotation of the rod.
Reactor auxiliary building
Ex-vessel transfer machine
Reactor building
Fuel Handling Machine(FHM)
In-VesselTransferMachine(IVTM)
Ex-Vessel Storage Tank(EVST)
Reactor vessel
Control roddrive
mechanism
5
The IVTM dropped about 2m when hung up from predetermined position on August 26, 2010 succeeding to the refueling.
� IVTM removed from reactor vessel (June 2011)
� Confirmed vessel structure integrity and no missing components of IVTM.
� Conducted test refueling operation with a new IVTM and a modified gripper (June 2012)
� The IVTM trouble recovered completely (Aug. 2012)
Ultimate heat sink = sea water Ultimate heat sink = atmosphere
� Decay heat is transferred through
sodium systems to air coolers (auxiliary
cooling system)
� Built-in passive safety feature of natural
convection heat transport with sodium
� Inherently strong against tsunamis
� Decay heat removal strongly relies on
heat transport to sea water
� Seawater pumps (and even DGs) are
sometimes placed at lower elevation
� Disadvantageous features against
tsunamis
Decay heat removal and heat sinks
(Comparison between LWR and SFR)
BWR Monju
6
around
17mmmm
IHX
Reactor
vessel
Air cooler
Heat source(Core)
Heat sink(Air cooler)
Power-supply vehicle
Power supply
Features of Monju and Measures for SBO
Ingenious layout of equipments and pipes
Heat to be emitted into atmosphere
Air
Center of reactor
around
7mmmm
T.P.+0m ▽▽▽▽
T.P.-6.5m
T.P.+5.2m▽▽▽▽
Intake
T.P.+21m
Reactorbuilding
Reactor auxiliary building6.4mT.P.+31m
T.P.+42.8m
Breakwater Curtain wall
Screen pump room
Spent fuelpool
Diesel building
Ex-vessel fuel storage tank
Important facilities, including sodium systems and spent fuel storage facility, locate at 21m above sea level. (JAEA envisages the tsunami height around 5.2m.)
During SBO, the spent fuels in the Ex-vessel Fuel Storage Tank (EVST) are cooled by natural circulation.
After reactor shutdown, decay heat is removed by natural circulation during SBO.
7
Sea level +0m ▽▽▽▽
-6.5m
++++5.2m▽▽▽▽
取水口取水口取水口取水口
6.4m
� Assurance of decay heat removal and ultimate heat sink� Inherent safety features of natural-convection decay heat removal have been
re-evaluated for both the core and EVST
� Forced-convection cooling has been made available as well, with electricity from the power supply vehicle.
+21m
� Sealing of sea-
water piping
The penetrations to
the buildings were
water-tightened.
� Water supply to spent fuel pool
The fuel is cooled in EVST and then in
fuel pool. The power in the pool is low
enough to avoid boiling, but water
supply is prepared using fire engines.
<Earthquake>
Fukushima-Daiichi accident and consequence
� Reactor shutdown successfully
� Emergency DGs actuated normally
� Reactor cooling systems operated as intended
� Loss of off-site power supply due to failure of power transmission line
� Essential power equipment such as DGs, switch boards, and
butteries were all flooded
� Seawater pumps failure, leading to loss of ultimate heat sink
� Station black-out (loss of off-site and on-site DG power)
Long-lasting station blackout and loss of ultimate heat sink conditions led
to severe fuel damage, loss of confinement capability, and serious off-
site release of radioactive materials.
Safety measures implemented in LWRs
� Measures under the SBO condition
� Diverse power supply for plant monitoring
� Measures for loss of cooling in fuel pool
� Preparation of water supply to spent fuel pool
� Measures to avoid seawater intrusion
� Water-tightening of seawater piping
Seawater pumpsCurtain wallBreakwater
afterbefore
� Measures for cooling� Insulators wereInsulators wereInsulators wereInsulators were packaged packaged packaged packaged
for easy manual access for easy manual access for easy manual access for easy manual access to to to to
the valves inthe valves inthe valves inthe valves in the auxiliary the auxiliary the auxiliary the auxiliary
cooling system (air coolers).cooling system (air coolers).cooling system (air coolers).cooling system (air coolers).
� Inspection and drills�Repetition of drills
�Manuals
Reactor bldg.
+42.8m
DG bldg.
Reactor auxiliary bldg.
Emergency
DGs (3)
Control room,
power
systems, etc.
Fuel pool
� Disposition of power vehicles
Buttery charging
Butteries
EVST cooling
Plant monitoringPlant
protection
systems
Control room
air conditioning
Air coolers
Vehicles (300kVA x 2)
A larger-capacity power vehicle
with gas turbine (4000kVA) is
to be disposed in 2013.
Post-Fukushima Safety Improvement in Monju 8
<Tsunami>
<Consequence>
給水給水給水給水ポンプポンプポンプポンプ
循環水循環水循環水循環水ポンプポンプポンプポンプ
冷却水冷却水冷却水冷却水
(海水)(海水)(海水)(海水)
復水器復水器復水器復水器
発電機発電機発電機発電機タービンタービンタービンタービン
蒸気蒸気蒸気蒸気
蒸気発生器蒸気発生器蒸気発生器蒸気発生器
(過熱器)(過熱器)(過熱器)(過熱器)
蒸気発生器蒸気発生器蒸気発生器蒸気発生器
(蒸発器)(蒸発器)(蒸発器)(蒸発器)
伝熱管伝熱管伝熱管伝熱管
Under station blackout (SBO) condition
No power
to pumps
No power to
blower
(natural convection)
Decay heat removal from the core
Air cooler
Secondary
sodium
Primary
sodium
IHX
Shield wall
CV
N2
Core
Guard vessel Reactor vessel
SG inlet valve (close)
AC outlet valve (open)
Primary pump
Secondary
pump
9
10
Secondary systemPrimary system
� Coolant temperature is stably reduced below 250 C in 3 days.
� DHR by natural convection is possible only in 1 loop (out of 3)
� When a DG is recovered, coolant temperature is further stabilized
� Fuel and cladding temperatures stay below the safety criteria
� Conclusions unchanged even with more pessimistic assumptions
0
20
40
60
80
100
120
140
160
150
200
250
300
350
400
450
500
550
0 1 2 3 4 5 6 7
流
量
(
%
)
温
度
(
℃
)
時間(日)津波来襲津波来襲津波来襲津波来襲地震発生地震発生地震発生地震発生 ディーゼル発電機1台復旧ディーゼル発電機1台復旧ディーゼル発電機1台復旧ディーゼル発電機1台復旧
0
20
40
60
80
100
120
140
160
150
200
250
300
350
400
450
500
550
0 1 2 3 4 5 6 7
流
量
(
%
)
温
度
(
℃
)
時間(日)津波来襲津波来襲津波来襲津波来襲地震発生地震発生地震発生地震発生 ディーゼル発電機1台復旧ディーゼル発電機1台復旧ディーゼル発電機1台復旧ディーゼル発電機1台復旧
N.C. F.C.
1ry flowrate
RV outlet
RV inlet
Pony motor restart
DHR from core under SBO (long term)
Tem
pera
ture
, C
Tem
pera
ture
, C
Flo
wra
te,
%
Flo
wra
te,
%
1ry flowrate
RV outlet
RV inletPony motor restart
Time (day)Time (day)
N.C. F.C.
10
Comprehensive safety assessment 11
Station Blackout
PlantShut-down
Emergency Power Supply
CoolingDown
Natural Circulation
CoolingDown
Failure
CoreDamage
1.25*
>2.2* 1.79* 1.53*
DieselGeneratorStart-up
1.25*
1.86*
: cliff edge
Air CoolerForced
Circulation
Success
FailureFailure Failure
Failure
Success Success Success
Success
Tolerance in design-
basis earthquake
acceleration(760 gal)
*:
Core damage sequence in the case of extreme earthquake
� In the case of extreme earthquake, the weakest safety-related component
was evaluated to be a valve at the outlet sodium piping of the air cooler,
which needs to be operated to establish a coolant path to the heat sink.
The valve can withstand the acceleration level 1.86 times larger than the
design basis earthquake acceleration (760 gal (0.78g)).
� For tsunami, our design-basis tsunami height is 5.2m above the standard
sea level. Since the plant is built on a ground level of 21m above the sea
level, our tsunami design has a safety margin of a factor of 4.0.
� Monju has an advantageous safety feature for decay heat removal with the
air being the ultimate heat sink. There is no cliff-edge effect under the
conditions of SBO or loss of ultimate heat sink.
Changes in national nuclear program 12
� A series of public debates and hearings were held during the
summer 2012 with three nuclear options: 0%, 15%, 20-25%
nuclear in 2030.
� A majority of the public seemed to support a nuclear phase-
out scenario, and this led to a government report on energy
and environment strategy *.
� For Monju, the report states that:
� a limited-term research plan shall be developed; and
� a research plan for reducing the volume and toxicity of
radioactive wastes.
� A working group was formed in MEXT in Oct. 2012 to
develop a research plan of Monju. The WG will continue
until the summer 2013 to formulate a detailed plan.
� Prioritization of R&D subjects and
� With emphasis on safety technologies and international
cooperation.
*:The present government revises the former strategy by a zero base to construct a responsible energy
policy . (Third meeting of the Headquarters for Japan's Economic Revitalization on 25th Jan. 2013)
Prioritization on Monju R&Ds (example) 13
1. Only by Monju 2. Reasonable by Monju 3. Possible without Monju
A. E
ssentialS
FR
technolo
gy
A-1
� Core design and management with
fuel containing higher-order Pu/Am .
� Design and evaluation of loop-type
cooling system, hot reactor vessel,
fuel handling system, etc.
� In-service-inspection devices for R/V
and primary cooling piping.
� Maintenance of primary cooling
system.
� Sever accident evaluation.
� Design and evaluation of heat
removal by Natural convection in
loop-type cooling system.
� Design and evaluation of
Sodium/Water reaction prevention
and mitigation.
A-2
� Design of large scale fuel
assembly
� Design and evaluation of large
scale main sodium components.
� Instrumentation for sodium leakage,
Sodium/Water reaction & failed fuel
detection.
� In-service-inspection devices for
Heat exchange pipe of Steam
generator
� Sodium handling technique.
� Maintenance of Fuel handling
system.
A-3
� Design and evaluation of
components under high
temperature.
� Prevention and mitigation
design of sever accident in
large scale SFR
B. U
sefu
l
SF
R
technolo
gy B-1
� Design and evaluation of transient
and control characteristics of the
steam/water system.
B-2
� Design and evaluation of steam
generators.
B-3
� Design and evaluation of
secondary pumps.
A. N
on S
FR
technolo
gy C-1
� Design of support systems (utilities,
air, etc.)
� Turbine and generators
C-2
(none)
C-3
(none)
Imp
orta
nc
e o
f R
&D
Su
bje
cts
Necessity of using Monju
Prospect for future 14
� In a wake of Fukushima accident, national law and regulations for ascertain
the safety of nuclear installations have been amended significantly in Japan.
� Although the safety standard under development is to be applied to LWRs, but
is to be considered also in Monju, taking detailed consideration of the
differences in safety features and design characteristics between LWRs and
sodium-cooled fast reactors.
� The future plan of SSTs and operation of Monju will be judged based on a
research plan developed by the MEXT Working Group through 2013.
� In parallel to this, we will keep our effort to improve the safety of Monju taking
the lessons learned from Fukushima. We believe the risk level of Monju can be
kept low and with added accident management measures to further improve
safety the level will be made much lower.
� The roles of Monju as a prototype continue to be important. In addition, it must
be emphasized that some of the international joint research programs using
Monju are still actively continuing, because the reactor is one of the very few
fast reactor plants that are operable today.
� Thus Monju is expected to play a role as an international asset to provide
research facility and knowledge/technology transfer to future generations.
Concluding remarks 15
� Monju, still in a stage of system start-up tests, has
encountered a series of challenges: accident/troubles in itself,
Fukushima-Daiichi severe accident, political and public
debates, and so on. Each of them so far has been overcome.
� We have confirmed SFRs have advantageous safety features
against loss-of-heat-sink-type accidents, and our safety
improvement effort will make a risk level of Monju much lower.
� Even in a wake of Fukushima, the roles of Monju has not
changed, especially in countries like Japan having no energy
resources. An on-going discussion in MEXT WG positively
supports this.
� Monju is expected to play a role as an international asset to
provide a research field and to transfer knowledge/technology
base to future SFRs.