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Fast Reactor Cycle Technology Development Project

Progress on Fast Reactor Development

in Japan

May 21-24, 2013

Hiroaki OHIRA

Nariaki UTO

Japan Atomic Energy Agency (JAEA)

46th TWG-FR Annual Meeting, May 21-24, IAEA HQ, Vienna, Austria


1

Contents

1. Changes in national nuclear program

2. Experimental Fast Reactor JOYO

3. Prototype Fast Reactor MONJU

4. SFR Development in Japan



2

1. Changes in national nuclear program

� 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)



3

2. Experimental Fast Reactor JOYO

Current Status and Future Plan of

Joyo Mark-III



4

Experimental Fast Reactor Joyo

Apply to Monju

JSFR

Overview of Main Cooling Bldg.

Air Blower

Main Control Room

Rotating Plug

・ To demonstrate the basic FBR technology

・ To conduct irradiation of fuel and materials

・ To validate innovative technologies for the

development of future FR

Attain Initial Criticality：1977

First Operation (MK-I; 50/75MWt; Breeder core) ：1978/1980

ibid (MK-II; 100MWt; Irradiation core) ：1983

ibid (MK-III; 140MWt; Upgraded irradi. core) ：2004

Role of Joyo


5

Safety of Joyo in case of a tsunami and blackout

38m

Approx. 300m

• Joyo is located at 38 meters above

sea level.

• The heat is released into air in the

secondary loop finally.

• Decay heat can be removed by a

natural circulation of sodium.

Natural circulation

Location of Joyo

Ocean

~~

Air



6

UCS retrieval device (cask)

UCS

MARICO-2 retrieval device

Replace UCS Retrieve MARICO-2

In-vessel observation device

Small rotating plug

Transfer pot

MARICO-2Finger

MARICO: Material testing rigs with temperature control

Replace UCS and Retrieve MARICO-2



7

Screw Jack-up

Equipment

Layout of equipment in

mock-up test (an example) Objective :To confirm the function of devices and methods for- Monitoring and control of jack-up speed, hanging load and inclination

- Visual monitoring of states of the UCS in the retrieval operation

UCS

模擬体

Appearance of mock-up test equipment

in the UCS manufacturer’s machinery works

Approx.

9m

Approx.

7mApprox.

5mDummy

UCS

DummyReactorvessel

Dummy

Rotating

Plug

Temporary

Boundary

Load-

Measuring

Device

Level-

Measuring

DeviceHanging Plate

Visual Monitoring Device

Dummy UCS

Guide Sleeve

of Dummy

Rotating Plug

Ex-vessel mock-up test for UCS replacement



8

JFY 2007 2008 2009 2010 2011 2012 2013 2014

15th Periodical Inspection

Design of MARICO-2 Retrieval Device

Planning of UCS Replacement Work

Manufacturing ofNew UCS

In-vessel Visual Inspection

UCS Replacement

MARICO-2 Retrieval

Lifting-up testof MARICO-2

▼ MK-III 6’ cycle

Manufacturing ofMARICO-2 Retrieval Device

▼Failure of the test subassembly disconnection

Re-installation of auxiliary equipment

Completion of Resumption

MARICO: Material testing rigs

with temperature control

Master Schedule of Joyo



9

3. Prototype Fast Reactor MONJU


Current Status and Future Plan


10

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 is under discussion in the MEXT WG, with its report being expected in summer 2013.


History of Monju


11


Overview of System Start-up Test (SST)


12Achievement 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



13

� 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

AHMgripper

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 drivemechanism

Ex-Vessel Storage Tank

(EVST)Reactor vessel

Opening andclosing rod


Recovery from hardware troubles


14Features of Monju and Measures for SBO

around

17ｍｍｍｍ

IHX

Reactorvessel

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

7ｍｍｍｍ

T.P.+0m ▽▽▽▽

T.P.-6.5m

T.P.+5.2m▽▽▽▽

Intake

T.P.+21m

Reactorbuilding

Reactor auxiliary building

6.4mT.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



15Post-Fukushima Safety Improvement in Monju

Sea level +0m ▽▽▽▽

-6.5m

++++5.2m▽▽▽▽

取水口取水口取水口取水口

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 pumpsCurtain wallBreakwater

afterbefore

(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 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.

<Tsunami>

<Consequence>



16Decay heat removal from the core

seawater

DG

Reactor BuildingPump

Shutdown

Pump

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 CoolerOutlet Stop Valve

(Opened)seawater

DG


Shutdown

Pump

Shutdown



ContainmentVessel


Primary Sodium

N2 Gas


Fuel Control Rod



Air Cooler

SecondarySodium



(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


Shutdown

Pump

Shutdown



ContainmentVessel


Primary Sodium

N2 Gas


Fuel Control Rod



Air Cooler

SecondarySodium



(Opened)seawater

DG


Shutdown

Pump

Shutdown



ContainmentVessel


Primary Sodium

N2 Gas


Fuel Control Rod



Air Cooler

SecondarySodium



(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.



17DHR 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

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

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250

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350

400

450

500

550

0 1 2 3 4 5 6 7

流

量

（

％

）

温

度

（

℃

）


(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


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

（

℃

）


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


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

（

℃

）


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


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

（

℃

）



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


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

（

℃

）



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


Earthquake~SBO



18Comprehensive safety assessment

Station Blackout

Plant

Shut-down

Emergency

Power Supply

Cooling

Down

Natural Circulation

CoolingDown

Failure

Core

Damage

1.25＊

＞2.2＊ 1.79＊ 1.53＊

Diesel

GeneratorStart-up

1.25＊

1.86＊

： cliff edge

Air Cooler

Forced 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.



19

� 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.


Prospect for future


4. SFR Development in Japan

20



21

SFR Development in Japan99 00 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15

FS P-I

FS P-II Earthquake and 1F accident

FaCT P-I

(FaCT P-II)

Safety Enhance

SDC SDG

(JFY: April to March)

� FS phase-I: 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: Detail comparison on selected concepts and selection of JSFR concept (sodium cooled + MOX fuel)

� FaCT phase-I: Evaluation on key technologies for commercial JSFR

� FaCT phase-II (suspended due to the TEPCO Fukushima Dai-ichi (1F) Accident): Demonstration of key technologies and conceptual design of demonstration JSFR

� After 1F accident: Design study on safety enhancement was initiated.

� Safety Design Criteria (SDC): SDC was initiated in October 2010 in GIF to establish global safety requirements for SFR. Necessity of SDC was reconfirmed by 1F accident and effort on SDC was enhanced. SDG should be established secondary to SDC completion, too.



Secondary pump

SG

Integratedpump-IHX

Reactor Vessel

Reactor Core

Japan Sodium-cooled Fast Reactor (JSFR)

22

Advanced Loop-type SFR Design concept

Items SpecificationsElectricity/Thermal output 1,500 / 3,530 MW

Configuration LoopPrimary 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



23

SDC key concept� Safety Goals in Gen-IV reactors

� Excel in Operational Safety and Reliability

� Very low likelihood & degree of reactor core damage

� Elimination of the need for offsite emergency response

� Key Safety Approach� Applying reinforced Defence-in-Depth (DiD)

� Built-in, and not by add-on

� Measures for prevention & mitigation of Severe Accident (SA) in DiD Level 4

� Passive safety utilization for DEC

Design Basis

Level 1: Prevention of abnormal operation and failures

Level 2: Control of abnormal operation and detection of failures

Level 3: Control of accidents within the design basis

Design Extension

Condition (DEC)

Level 4: Control of severe conditions including prevention of accident

progression and mitigation of the consequences of a severe accident

Off-site response Level 5: Mitigation of radiological consequences of significant external

releases of radioactive materials

* Defence-In-Depth, ** Severe Accident46th TWG-FR Annual Meeting, May 21-24, IAEA HQ, Vienna, Austria


24

Hypothetical 1F conditions on JSFR version 2010- Power supply and CCWS/SCWS -

<Conventional system>

DHRS safety grade components

� Pony motor on the primary pump

� DHRS 2nd circuit EMP

� Blowers for air cooler

Emergency DG

Component cooling

water system (CCWS)

Sea water cooling system

(SWCS)

depend depend

<JSFR with Natural circulation DHRS>

DHRS safety grade components

� Air cooler damper Emergency battery (DC power)

� JSFR does not require quick emergency power activation because of full NC

DHRS and is able to introduces emergency gas turbine generator(GTG) with air

cooling which requires a few dozen seconds for activation.

� JSFR DHRS is independent electric supply thanks for NC DHRS except for

dumper control of air cooler.

� JSFR CCWS and SWCS are non-safety grade as in 2010 design.

depend

Affected by tsunami



25

Hypothetical 1F conditions on JSFR version 2010- Hypothetical Tsunami -

Tsunami Loss of SWCS

<realistic for JSFR>

<hypothetical assumption for SBO transient analysis>

GTG activation* Stable cooling

*: independent form SWCS

TsunamiLoss

of GTG

DHRS

activation*

*:NC mode

*: AC damper open by battery

Out of battery

in 2h*Loss of air cooler

damper control

Sodium freeze

at air cooler

PLOHS

*:As in 2010 design

� Transient analysis on hypothetical SBO condition has been

conducted.

� Time margin to protected loss of heat sink (PLOHS) is important.

� If time margin is sufficient, PLOHS can be managed by AM such

as manual dumper control.



26

Hypothetical 1F conditions on JSFR version 2010- Hypothetical SBO transient analysis -

� SBO analysis shows sufficient time margin to initiate AM


27

Hypothetical 1F conditions on JSFR version 2010- Seismic isolation system and conditions -

Oil damper Laminated rubber bearings

� Avoid resonances between seismic frequency and natural frequencies of components.

� Mitigate horizontal seismic forces on components.

� The 1F-envelop condition is basically less than the JSFR safety shutdown earthquake.

Horizontal seismic condition

Vertical seismic condition


28

Hypothetical 1F conditions on JSFR version 2010- Seismic margins -

� The 1F-envelop condition has been used in this seismic evaluation.

� Seismic design margins of JSFR ver. 2010 are sufficient for 1F seismic conditions.

Component Item Margin

RVBuckling 2.4

Control rod insertion capability 1.7

Integrated-IHX BucklingY piece 22.8

Skirt 50.0

SG BucklingSkirt 21.1

Main Body 20.9

Primary pipingStress (Hot-leg) Support piping 17.6

Stress (Cold-leg) Elbow center 5.3

Secondary pipingStress (Hot-leg) Elbow center 13.7


DRACS secondary pipingStress (Hot-leg) Elbow center 4.7


PRACS secondary pipingStress (Hot-leg) T piping 3.9

Stress (Cold-leg) T piping 4.7

EVST Stress

Rotating rack 2.1

Inner tank 2.5

Outer tank 23.8

EVTM Stress

Main body 23.4

Rail hook 6.4

Locator pin 1.9



29

Lessons learned from 1F Accident- Consideration for safety enhancement -

� Enhancement of systems that may needed in order to decrease the likelihood of a severe

accident due to extreme external hazards. Namely, robustness on power supplies and

cooling functions including final heat sink.

� Enhancement of response measures against severe accidents. The means should be

provided to prevent severe loads on CV and the instrumentation should be prepared to

identify status of reactor core and CV.

� Reinforcement of safety infrastructure by ensuring independency and diversity of safety

systems.

Design should consider

� Long term loss of all AC power

� Passive safety functions reducing dependency on power supply

� Severe external events which could be SA initiators such as earthquakes, tsunami and flood

� JSFR already introduced full NC DHRS and dependency on power supply is low. However,

sufficiency of DHRS from the DiD point of view has to be reconfirmed.

� External events should be analyzed and design measure should be introduced. For the 1F

conditions, the previous analysis showed JSFR toughness.



30

Lessons learned from 1F Accident- Situations to be practically eliminated -

<ATWS>

� In-balance of power and cooling might causes core damage within a shorter time period.

� Relocation of damaged core could cause potential mechanical energy release.

� Severe mechanical energy release due to coherent sodium boiling or molten fuel compaction,

failure of decay heat removal from degraded core in ATWS type events

� JSFR already meet this requirement as JSFR 2010 version

<LOHRS>

� Decay heat is at a few percent of the nominal power, the temperature of the reactor coolant

system would be slow with sufficient time margin to make recovery action for failed DHRSs

and/or implementation of back up cooling measures.

� If no heat sink is available, core and coolant boundary failure might occur due to high

temperature causing significant thermal loads not to cope with on the containment.

� Significant core damage in LOHRS type events

� Even though JSFR has full NC DHRS, additional DHRS for DEC which is independent of DBA

DHRS is under discussion.



31

Lessons learned from 1F Accident- DHRS and additional improvements -

Original DHRS

Redundant and Diversified DHRS against DBA� DRACS + PRACSx2 (Natural circulation)

ACS

DRACS PRACS

SG Nitrogen

Gas Cooling

Additional improvements against DEC

Further defense lines against DEC� Improved DRACS against loop siphon break� Accident management on DRACS and PRACS� Auxiliary Cooling System (ACS)� SG Nitrogen Gas Cooling



Thank you for your attention!


Appendix



34

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


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


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


Prioritization on Monju R&Ds (example)

Fast Reactor Cycle Technology Development ProjectJAEA is proposing AtheNa-SA experiments for various DHRS in GIF framework.

International cooperation on SA cooling measures

・Dimension:130m x 62m x 55m

・Total floor area: 11,000m2

・Sodium inventory: 240ton

� Criteria to practical elimination of LOHRS and sufficient design measures should be discussed

internationally.

� Design measures and design tools should be developed and validated internationally regarding

objectivity and cross-check.

AtheNa facility


Documents

Progress on Fast Reactor Development in Japan · Overview of Main Cooling Bldg. ... 250 300 350 400 450 500 ... Progress on Fast Reactor Development in Japan ()