progress on fast reactor development in japan
TRANSCRIPT
Fast Reactor Cycle Technology Development Project
0
Progress on Fast Reactor Development
in Japan
May 19-23, 2014
Hiroaki OHIRA
Nariaki UTO
Japan Atomic Energy Agency (JAEA)
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Contents
1. Experimental Fast Reactor JOYO
2. Prototype Fast Reactor MONJU
3. SFR Development in Japan
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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1. Experimental Fast Reactor JOYO
Current Status and Future Plan of Joyo Mark-III
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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JFY 2007 2008 2009 2010 2011 2012 2013 2014
15th
Periodical Inspection
Design of MARICO-2 Retrieval Device
Planning of UCS
Replacement Work
Manufacturing of new UCS and
Device for Replacement
In-vessel Visual
Inspection
UCS
Replacement
MARICO-2
Retrieval
▼ MK-III 6’ cycle
Manufacturing of
MARICO-2 Retrieval Device
▼Failure of the test subassembly disconnection
Lifting-up test
of MARICO-2
Re-installation of
auxiliary equipment
Progress of Joyo resumption
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Cask
(2) Retrieval of
ed-UCS
Door valve
(3) Retrieval of
MARICO-2 S/A
(4) Installation of
n-UCS
L-R/P
S-R/P Temporary pit cover
Screw jack-up equipment
Guide tube
(1) Jack-up of
ed-UCS
UCS
Cask
MARICO-2
S/A
MARICO-2 S/A retrieval equipment
Wire jack-up equipment
Ed-UCS is extracted through
small-diameter part.
UCS
ed-UCS : existing damaged UCS
n-UCS : new UCS
Outline of restoration work plan
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Mock-up test for UCS replacement Mock-up test for MARICO-2 retrieval
Ex-vessel mock-up test for resumption work
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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• The ex-vessel mock-up test for replacement of the UCS and MARICO-2 retrieval have completed to confirm the performance of related equipment and the work procedure.
• The new UCS installation aims to complete within this year (2014).
• Various irradiation experiments are expected in Joyo after the resumption to develop FR fuel technology. Irradiation experiments using Joyo to support the R&D program for Pu and MA burning in Monju is being discussed now, according to the R&D plan by the working group in MEXT*.
*Ministry of Education, Culture, Sports, Science & Technology
• The new regulatory requirements for research reactors were launched on December 18, 2013. In order to apply for permission to restart, safety improvement measures are under consideration.
Current Status of Joyo
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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2. Prototype Fast Reactor MONJU
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Sodium leak detection
and monitoring system
Modification of 2ry
sodium piping
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 were decided under discussion in the MEXT WG, with its report were published in September 2013.
History of Monju
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Beyond 45 % reactor output, whole steam is able to be 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… to confirm the performances
2) Following the three steps below with inspections, evaluations and confirmations:
Core ConfirmationTest (CCT)
Inspection
40%-power Confirmation Test (40%CT)
Inspection, evaluation and confirmation
Power RisingTest (PRT)
RefuelingRefueling
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.
Entire plant performances are 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- Current configuration
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
Beyond 45 % reactor output, whole steam is able to be 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… to confirm the performances
2) Following the three steps below with inspections, evaluations and confirmations:
Core ConfirmationTest (CCT)
Inspection
40%-power Confirmation Test (40%CT)
Inspection, evaluation and confirmation
Power RisingTest (PRT)
RefuelingRefueling
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.
Entire plant performances are 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- Current configuration
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 Start-up Test (SST)
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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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 fuel Startup and operation
Reactivity worth of all the 19 control rods
Safe control and shutdown of the reactor Safe 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
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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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
Cross-section view of reactor
vessel when the IVTM is hung up
Dropped
IVTM
Schematic view of
the IVTM
AHM gripper
90mm
Auxiliary Handling Machine (AHM) gripper failed to fully open, due to rotation of the rod.
Ex-vessel transfer machine
Reactor building
In-Vessel
Transfer
Machine
(IVTM)
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)
Auxiliary Handling Machine
(AHM)
Reactor auxiliary building
Fuel Handling Machine
(FHM)
Control rod drive mechanism
Ex-Vessel Storage Tank
(EVST) Reactor vessel
Opening and closing rod
Recovery from hardware troubles
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Features of Monju and Measures for SBO
around 17m
IHX
Reactor vessel
Air cooler
Heat source (Core)
Heat sink (Air cooler)
Power-supply vehicle
Power supply
Ingenious layout of equipments and pipes
Heat to be released to atmosphere
Air
Center of reactor
around 7m
T.P.+0m ▽
T.P.-6.5m
T.P.+5.2m ▽
Intake
T.P.+21m
Reactor building
Reactor auxiliary building
6.4m T.P.+31m
T.P. +42.8m
Breakwater Curtain wall
Screen pump room
Diesel building
EVST
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.
SFP
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Post-Fukushima Safety Improvement in Monju
Sea level +0m ▽
-6.5m
+5.2m ▽
取水口
6.4m
(6) 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
(3) Sealing of sea-water piping
The penetrations to the buildings were water-tightened.
(2) 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 pumps Curtain wall Breakwater
after before
(4) Measures for cooling Insulators were packaged
for easy manual access to the valves in the auxiliary cooling system (air coolers).
(5) 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
(1) Disposition of power vehicles
Buttery charging
Butteries
EVST cooling
Plant monitoring Plant
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. <Tsunami>
<Consequence>
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Decay heat removal from the core
seawater
DG
Reactor BuildingPump
Shutdown
Pump
Shutdown
Air Cooler Blower Shutdown
Air Cooler Blower Shutdown
ContainmentVessel
Primary MainCirculation Pump
Primary Sodium
N2 Gas
IntermediateHeat Exchanger
Fuel Control Rod
Guard VesselReactor Vessel
Secondary MainCirculation Pump
Air Cooler
SecondarySodium
Steam GeneratorInlet Stop Valve(Closed)
Air Cooler
Outlet Stop Valve
(Opened)seawater
DG
Reactor BuildingPump
Shutdown
Pump
Shutdown
Air Cooler Blower Shutdown
Air Cooler Blower Shutdown
ContainmentVessel
Primary MainCirculation Pump
Primary Sodium
N2 Gas
IntermediateHeat Exchanger
Fuel Control Rod
Guard VesselReactor Vessel
Secondary MainCirculation Pump
Air Cooler
SecondarySodium
Steam GeneratorInlet Stop Valve(Closed)
Air Cooler
Outlet Stop Valve
(Opened)
» The main pumps of the primary and secondary main cooling systems are inoperable.
» The blowers of AC are also inoperable and AC power supply is stopped in a big earthquake.
» The SG inlet stop valves are
closed, and the AC outlet stop
valves are opened just after
the reactor scrum.
» The pony motors are still
operating by the emergency
DGs.
» However, SBO occurred by a
huge Tsunami.
seawater
DG
Reactor BuildingPump
Shutdown
Pump
Shutdown
Air Cooler Blower Shutdown
Air Cooler Blower Shutdown
ContainmentVessel
Primary MainCirculation Pump
Primary Sodium
N2 Gas
IntermediateHeat Exchanger
Fuel Control Rod
Guard VesselReactor Vessel
Secondary MainCirculation Pump
Air Cooler
SecondarySodium
Steam GeneratorInlet Stop Valve(Closed)
Air Cooler
Outlet Stop Valve
(Opened)seawater
DG
Reactor BuildingPump
Shutdown
Pump
Shutdown
Air Cooler Blower Shutdown
Air Cooler Blower Shutdown
ContainmentVessel
Primary MainCirculation Pump
Primary Sodium
N2 Gas
IntermediateHeat Exchanger
Fuel Control Rod
Guard VesselReactor Vessel
Secondary MainCirculation Pump
Air Cooler
SecondarySodium
Steam GeneratorInlet Stop Valve(Closed)
Air Cooler
Outlet Stop Valve
(Opened)
» The main pumps of the primary and secondary main cooling systems are inoperable.
» The blowers of AC are also inoperable and AC power supply is stopped in a big earthquake.
» The SG inlet stop valves are
closed, and the AC outlet stop
valves are opened just after
the reactor scrum.
» The pony motors are still
operating by the emergency
DGs.
» However, SBO occurred by a
huge Tsunami.
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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DHR from core under SBO (long term)
Coolant temperature is stably reduced below 250ºC 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
(a) Primary Heat Transport System
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)津波来襲地震発生 ディーゼル発電機1台復旧
RV outlet
RV inlet
Flow rate
Natural circulation
Forced circulation
Te
mp
era
ture
(ºC
)
Flo
w r
ate
(%
)
160
140
120
100
80
60
40
20
00 1 2 3 4 5 6 7
Time (day)
550
500
450
400
350
300
250
200
150
Pony motor restarted
Earthquake~SBO
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)津波来襲地震発生 ディーゼル発電機1台復旧
(b) Secondary Heat Transport System
Time (day)0 1 2 3 4 5 6 7
550
500
450
400
350
300
250
200
150
Natural circulation
Forced circulation
Te
mp
era
ture
(ºC
)
160
140
120
100
80
60
40
20
0
Flo
w r
ate
(%
)
AC outlet
AC inlet
Flow rate
Pony motor restarted
Earthquake~SBO
(a) Primary Heat Transport System
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)津波来襲地震発生 ディーゼル発電機1台復旧
RV outlet
RV inlet
Flow rate
Natural circulation
Forced circulation
Te
mp
era
ture
(ºC
)
Flo
w r
ate
(%
)
160
140
120
100
80
60
40
20
00 1 2 3 4 5 6 7
Time (day)
550
500
450
400
350
300
250
200
150
Pony motor restarted
Earthquake~SBO
(a) Primary Heat Transport System
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)津波来襲地震発生 ディーゼル発電機1台復旧
RV outlet
RV inlet
Flow rate
Natural circulation
Forced circulation
Te
mp
era
ture
(ºC
)
Flo
w r
ate
(%
)
160
140
120
100
80
60
40
20
00 1 2 3 4 5 6 7
Time (day)
550
500
450
400
350
300
250
200
150
Pony motor restarted
Earthquake~SBO
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)津波来襲地震発生 ディーゼル発電機1台復旧
(b) Secondary Heat Transport System
Time (day)0 1 2 3 4 5 6 7
550
500
450
400
350
300
250
200
150
Natural circulation
Forced circulation
Te
mp
era
ture
(ºC
)
160
140
120
100
80
60
40
20
0
Flo
w r
ate
(%
)
AC outlet
AC inlet
Flow rate
Pony motor restarted
Earthquake~SBO
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)津波来襲地震発生 ディーゼル発電機1台復旧
(b) Secondary Heat Transport System
Time (day)0 1 2 3 4 5 6 7
550
500
450
400
350
300
250
200
150
Natural circulation
Forced circulation
Te
mp
era
ture
(ºC
)
160
140
120
100
80
60
40
20
0
Flo
w r
ate
(%
)
AC outlet
AC inlet
Flow rate
Pony motor restarted
Earthquake~SBO
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Comprehensive safety assessment
Station Blackout
Plant Shut-down
Emergency Power Supply
Cooling Down
Natural Circulation
Cooling Down
Failure
Core Damage
1.25* >2.2* 1.79* 1.53*
Diesel Generator Start-up
1.25*
1.86*
: cliff edge
Air Cooler Forced
Circulation
Success
Failure Failure 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.
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Current Status of Monju
JFY2012 JFY013
Technical Investigations
Safety Confirmations
1. Planning of Future R&D activities
2. Confirmation of facility integrity (function tests of water-vapor system, etc.)
3. Compliance with the revised nuclear reactor regulation law
4. Accommodation with anti-seismic evaluation
Interim report
Annual inspection
Regulation standards formulation by the Nuclear Regulation Authority (NRA)
On-site investigations on crush zones
Review by NRA (Aug.) Additional
inspections
Promulgation and
Effectuation (July)
Final report (Sep.)
Current status
JFY2014
Order on Measures (May)
The Basic Energy Plan, which includes the Monju Research Plan, has been finalized as a cabinet decision. Monju is positioned as a research center for the radwaste reduction (MA burning).
Inspections for the restart have been suspended due to an order on measures for plant safety (maintenance system and
safety culture) issued by NRA. Confirmation is ongoing by NRA.
A final report is being drafted by an Advisory Committee on the philosophy for securing safety of Monju. The safety measures will be examined after finalizing the details by NRA.
The results of the additional geological inspections (no indications for active crush
zones) were reported to NRA. On-site inspections by NRA is being scheduled in the nearest future.
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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(1) Planning of Future R&D Activities:
● The Monju Research Plan has been finalized as a cabinet decision
on April 11, 2014.
- Sep. 25: The Monju Research Plan was drafted by the Working Group (WG) in
MEXT after 12-times discussions by the WG members.
- Dec. 13: The drafted Research Plan was finalized by the Advisory Committee
on Energy and Natural Resources as a part of the Basic Energy Plan
in Japan.
- Feb. 25: The discussions on the drafted Basic Energy Plan, proposed by the
cabinet, was initiated by the members of the Diet.
- Apr. 11: The Basic Energy Plan, including the Monju Research Plan, was
finalized as a cabinet decision based on the consensus within the
ruling parties.
Recent Progress in Monju (1/4)
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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【1】 Compilation of outcomes of FR development
Research items to challenge in Monju : 【1】~【3】
【2】 Reduction of the amount and toxic level of radioactive waste
【3】 Safety enhancement of FR
To confirm the technical feasibility as an FR plant Core/fuel, Component/system design, Sodium handling, Operation/maintenance
To confirm the effectiveness of reduction of environmental burden by FR Fuel fabrication, Irradiation test, Fuel/material development, Core/reactor system design, Reprocessing
To establish safety technology system for FR SA evaluation technology by using PSA for SA sequence and measures for safety enhancement, Training/operation for SAM
C&R of research plan
After ca. 2yrs at completion of SST incl. 1st cycle : Interim C&R
After ca. 6yrs at completion of 5th cycle : Compilation of research results
Rigorous C&R of technical achievement level
Judgment of whether the research should be continued, in consideration of the
C&R result, national policy of energy and international situation at that time
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
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(2) Confirmation of facility integrity:
● Preparatory activities for the upcoming pre-operational tests have
been suspended in May 2013.
- Feb. 14-15: Some excesses of inspection intervals were identified by
NRA to be against the safety regulations due to lack of
adequate paperwork.
- Nov. 19: A report was submitted to NRA, which described on the
reconstruction the plant maintenance management system
and quality assurance system, and revision of the plant
maintenance plan.
- Dec. 26: An application was submitted to NAR for revision on the
safety regulations to reform the Monju organization for
enhancement of the safety culture.
- Apr. 16: Some inappropriate paperwork for plant maintenance was
suspected. Further investigation and root cause analysis were
requested by NRA.
Recent Progress in Monju (2/4)
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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(3) Compliance with the revised nuclear reactor regulation law:
● Detailed discussions on the new regulation standards are ongoing.
- Dec. 24: An Advisory Committee on Basic Approach to Securing Safety
of Monju was established by JAEA. Discussions were initiated
in order to reconstruct the philosophy for securing the safety of
Monju.
- Apr. 23: A final report was discussed on the Monju safety philosophy
based on the 7-times discussions by the committee members.
The summary report will be finalized at the next meeting to be
held shortly.
Recent Progress in Monju (3/4)
● The details of the renewed standards for Monju will be discussed by
NRA based on the above-mentioned final draft. The adequacy of the
safety measures will be examined by NRA after finalizing the details
of the regulation standards.
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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Recent Progress in Monju (4/4)
(4) Accommodation with anti-seismic evaluation:
● Discussions on the on-site crush zones in Monju are ongoing.
- Nov. 29: The results of additional geological investigations on the on-site
crush zones were reported to NRA as a first interim report.
- Jan. 31: Additional results were reported to NRA as a second interim
report.
- Mar. 28: The final report was submitted to NRA.
● No evidence was found, which indicates any possibility of the
crush zones to be active and existence of additional active faults
around the Monju site.
● On-site geological investigations by NRA are being scheduled in
the nearest future.
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
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- Objective
Research for the safety improvement of SFR, support the safe
and stable operation of “MONJU”, and plays a role of the center
of international and regional research collaboration.
- Research Subjects
a) Investigation of severe accident of SFR
b) Development of ISI&R technology of SFR
c) Development of measurement and evaluation technique for
sodium cooling system of SFR
- Outline of the Facility
Building : steel construction, three-story, approx. 700 m2
Sodium storage capacity : approx. 6 tons
- Schedule
JFY2013 : Construction of the building
JFY2014 : Installation of experimental equipments
Sodium Engineering Research Facility (SERF)
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- Experimental equipment in SERF
a) Sodium test laboratory
Basic tests for under sodium viewer, and search and retrieval of lost parts.
b) Multi-purpose sodium tank and loop
Development of ISI&R technique for sodium adhering equipment and investigation of the propagation behavior of ultrasonic through liquid sodium.
c) Steel shell and chamber
Development of safety evaluation method for severe accidents such as sodium-concrete reaction.
Glove boxes and analyzers
Sodium Test Laboratory
Multipurpose sodium tank & loop
Storage Tank
EMP
Tank 1 C/T Tank 2
EMP
Steel sell and chamber
Controlled Atmosphere chamber
Steel cell for sodium burning test
Sodium Engineering Research Facility (SERF)
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3. SFR Development in Japan
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FS phase-I(’99-’00) Conceptual design study on wide range of various coolant/fuel and
selection of four systems (sodium, helium gas, lead-bismuth and water)
FS phase-II(’01-’05) Detail comparison on selected concepts and selection of JSFR concept
(sodium cooled + MOX fuel)
FaCT phase-I(’06-’10) Evaluation on key technologies for commercial JSFR (Still suspended due to the Great East Japan Earthquake)
After 1F accident Design study on safety enhancement was initiated.
Safety Design Criteria (SDC)/Safety Design Guides (SDG): SDC were initiated in October 2010 in GIF to establish global safety
requirements for SFR, and were approved in the GIF policy group meeting in May 2013.
SDG establishment work is ongoing.
SFR development progress in Japan
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
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Secondary pump
SG
Integrated pump-IHX
Reactor Vessel Reactor Core
Advanced Loop-type SFR Design concept
Items Specifications Electricity/Thermal output 1,500 / 3,530 MW Configuration Loop Primary sodium temp. 550 degree C Reactor vessel material 316 FR stainless steel Piping material Mod. 9Cr-1Mo steel Plant efficiency Approx. 42% Fuel type TRU-MOX
Safety
CDA prevention : SASS CDA mitigation: FAIDUS + - void reactivity <$6, - core height < 100cm, - specific power > 40kW/kg
Burn-up (ave.) for core fuel Approx. 150GWd/t Breeding ratio Break even (1.03) ~ 1.2 Cycle length 26 months or less, 4 batches
Japan Sodium-cooled Fast Reactor (JSFR)
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Activities on SFR enhanced safety (1)
Further safety enhancement to meet SDC (Safety Design Criteria) for
Gen IV. SFR (==> Later slides in detail)
Feasibility study on design measures for JSFR against DEC at the following
points:
• LOHS (Loss of Heat Sink) and Loss of sodium level to:
DHRS (Decay Heat Removal System)
EVST (External Vessel Storage Tank)
SFP (Spent Fuel Pool)
• Sodium leak from double boundary of 2nd loop, followed by the sodium fire on
the steel-concrete plate of reactor building
Establishment of SDG (Safety Design Guides)
SDG (Safety Design Guides) : a set of safety design measures to apply SDC to SFR
design
Study in a framework of 2nd phase of GIF SDC-TF, aiming at organizing:
• Design philosophy/concept of advanced reactors
• Safety design measures and conditions for each system (core, heat
transport system, containment vessel)
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
Fast Reactor Cycle Technology Development Project
Activities on SFR enhanced safety (2)
AtheNa*-SA experiment program
A candidate of new GIF-SFR project to conduct sodium experiments for various
cooling functions against SA (severe accident)
Continuous efforts to coordinate the following items have been made among
countries joining the program:
• Specifications of the experiments required from each country
• Scope of contribution of each country to the program
* Advanced Technology Experiment Sodium (Na) Facility
Maintenance and upgrading of safety-related analytical codes
Safety-related numerical analysis aiming at V&V of analytical codes
• Sodium-water chemical reaction
• High cycle thermal fatigue
• Gas entrainment through free surface of upper sodium plenum
• Flow-induced vibration of large-diameter piping, etc.
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Multiple DHRS (DRACS×1+PRACS×2)
Full natural circulation DHRS
Low operation loads (Activated by Air cooler damper operation
connected to Direct Current(DC)-power)
Redundant air cooler dampers (50%×2 at inlet/outlet damper)
Complete double boundary design (Measures against core uncovering with
sodium and sodium combustion in CV)
Low frequency of LOHS (under 10-8/ry)
DRACS PRACS×2
SG
IHX/Pump
Double Boundary (GV, Outer pipe, Enclosure)
Air cooler
DHRS design approach against LOHRS (1)
Heat Removal Design Approach for JSFR
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DHRS design approach against LOHRS (2)
Identification of plant failure state
L
O
H
S
Loss of flow at air cooler Stack collapse induced by aircraft crash etc.
Air cooler damper closure induced by damper failure or DC power loss
Loss of flow at DHRS
secondary loop Air cooler damper opening induced by
damper failure or DC power loss (Na freeze)
《Typical failure condition》 《DHRS failure》
L
O
R
L
Na level decrease (Core exposure)
Stack collapse induced by aircraft crash etc.
Air cooler damper closure induced by damper failure or DC power loss Na level decrease under
EsL
(Main loop Siphon break) Air cooler damper opening induced by
damper failure or DC power loss (Na freeze)
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DHRS design approach against LOHRS (3)
Measures against LOHRS
AM of air cooler damper
Assuring heat sink
diversity
Prevention of double boundary failure
of RV&GV
Functionality extension of DRACS
(submerged)
Improvement of air cooler
stack integrity or
dispersed layout of stacks
Preparation of
emergency
power supply
Against LOHS
Against LORL
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Measures against LOHRS to EVST (1)
Loop type forced
convection (4 lines)
Ultimate heat sink:Air
Air cooler damper:
・battery operated
・doubled at inlet / outlet
(manually control)
EVST vessel (double
boundary = inner/outer
vessel)
Time margin (max temp.
550℃) ・Normal fuel exchange:9 days
・Whole core evacuation:0.5 days
In addition, against SBO,
Natural circulation heat
removal at whole core
evacuation
EMP
EM F
Air
IHX
EM F
Dump tank s econdary
cooling line
Dump tank prim ary
cooling line EVST
Drain tank
Drain tank
Drain tank
Drain tank
Drain tank
Drain tank
Air Air
Air
Air
Pot (degassing)
Cooling line B
Cooling line C
Cooling line B
Cooling line C
Cooling line D
Cooling line A Cooling line B Cooling line C Cooling line D
d amper
Air Cooler
Air
Cooling line D
Air
Air
DBE Cooling system for EVST
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Measures against LOHRS to EVST (2)
Alternative cooling systems as coutermeasure against LOHRS
are under consideration.
Dipped heat exchanger type
Vessel auxiliary cooling type (Multiple/Diversification for DHX type)
Against LOHS Against Loss of sodium level
Prevention of double boundary failure
Inner vessel and outer vessel are to be designed,
manufactured, installed and inspected in the
highest design standards.
EVST Cooling pipe(8 piping)
N2 gas cooler
N2 gas
Buffer
tank
BlowerUltimate
Heat sink
(Sea water)
Ultimate
heat sink
EVST
EMP
Cooling
panel
Expansion
tank
Sodium cooler
Buffer
tank
N2 gas
Blower
N2 gas cooler
Ultimate
Heat sink
(Sea water)
Ultimate
heat sink
Ultimate
heat sink
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Measures against LOHRS to SFP (1)
DBE Cooling system for SFP
Failure assumption:loss of offsite power and failure of one active component
or failure of one static component
No. Failure assumption Power
supply
Cooling system No. of
Available
Systems
Operationa
l Number System
#1
System
#2
1 - ○ ○ ○ 2 2
2 Failure of one active
component
×
○ × 1 1
3 Failure of one static
component ○ × 1 1
4
Startup failure of one
emergency power
generator
○ × 1 1
2 independent cooling systems
(equipped with 2 independent emergency power supply)
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Candidate countermeasure
Heat exchanger
Purification
system
Auxiliary water tank Cooling
tower
Fuel water supplying system
Cooling system
(fuel transfer
system)
Coupling nozzle(for direct water supplying)
Coupling nozzle (direct water supplying to SFP)
Coupling nozzle (direct water supplying to cooling tower)
Pump car, etc. Makeup tank
SFP
Auxiliary water tank
Heat exchanger
Cooling tower
Purification
system
Cooling line A Cooling line B
Cooling system
(fuel transfer
system)
Enhanced cooling systems as candidate coutermeasure against
LOHRS are under consideration.
Measures against LOHRS to SFP (2)
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JSFR Measures against Sodium Leak No energetic release due to Core Disruptive Accident
Simple piping arrangement with Mod. 9Cr-1Mo steel with small
thermal expansion
Double wall piping with inert gas in annulus region
SC structure (concrete covered with steel) for CV and reactor building
格納バウンダリ(SCCV)
1次アルゴンガス系
1次冷却系
給水
蒸気
2次冷却系
2次ポンプ
1次ポンプ
Measures against secondary sodium fire (1)
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Measures against secondary sodium fire (2)
Sketch of SG room
The SG room has been selected as
analysis item, since it has large volume
and accommodates the tallest sodium
component.
Cases Evaluation points
Large spray Room pressure
Small spray &
subsequent local pool
Steel temperature of
SC structure
Large pool Concrete temperature
Sodium fire analysis was performed to evaluate measures
against sodium fire assuming hypothetical double
boundary failures of secondary loop.
Analysis Cases
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria
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Measures against secondary sodium fire (3)
Evaluated value Critical event Without measures With measure
Room pressure Large spray 0.13MPa (gage) 0.03MPa (gage)
(Limitation of spray height)
SC steel plate
temperature Small spray 805℃
539℃
(catch pan)
Concrete
temperature Large pool 155℃ at 500mm
148℃ at 500mm
(catch pan)
The pressure raise in the SG room could be
mitigated by limiting spray height.
For small spray, catch pan will effectively
reduce the maximum temperature of the SC
steel plate.
Combination of catch pans and leak
sodium drain system effectively protect
building structures from high temperature due
to large sodium pool fire.
Support floor
at 6th floor
Catch pan
47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria