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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

RECENT PROGRESS AND STATUS OF MONJU · PDF fileMarch 2013 Fast Breeder Reactor Research and Development Center(Monju) Japan Atomic Energy Agency Takehide Deshimaru RECENT PROGRESS

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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

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

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.

Thank you for you attention.

ごごごご静聴静聴静聴静聴ありあとありあとありあとありあとうごうごうごうござしましたざしましたざしましたざしました。。。。