progress on fast reactor development in japan

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


1

Contents

1. Experimental Fast Reactor JOYO

2. Prototype Fast Reactor MONJU

3. SFR Development in Japan



2

1. Experimental Fast Reactor JOYO

Current Status and Future Plan of Joyo Mark-III



3

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



4

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



5

Mock-up test for UCS replacement Mock-up test for MARICO-2 retrieval

Ex-vessel mock-up test for resumption work



6

• 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



7

2. Prototype Fast Reactor MONJU



8

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



9

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)



10

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



11

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



12

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



13

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>



14

Decay 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 Cooler

Outlet Stop Valve

(Opened)seawater

DG


Shutdown

Pump

Shutdown



ContainmentVessel


Primary Sodium

N2 Gas


Fuel Control Rod



Air Cooler

SecondarySodium


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


Shutdown

Pump

Shutdown



ContainmentVessel


Primary Sodium

N2 Gas


Fuel Control Rod



Air Cooler

SecondarySodium


Air Cooler

Outlet Stop Valve

(Opened)seawater

DG


Shutdown

Pump

Shutdown



ContainmentVessel


Primary Sodium

N2 Gas


Fuel Control Rod



Air Cooler

SecondarySodium


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.



15

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（℃）


(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



16

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.



17

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.



18

(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)



19

【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



20

(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.




21

(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.


● 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.



22


(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.



23

- 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)



24

- 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)



25

3. SFR Development in Japan



26

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



27

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)



28

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)



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.

29 47th TWG-FR Annual Meeting, May 19-23, 2014, IAEA HQ, Vienna, Austria


30

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



31


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)



32


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



33

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



34

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



35

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)



36

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)



37

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)



38


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



39


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


progress on fast reactor development in japan

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