Status of Fast Reactor Technology Development in Korea€¦ · · 2013-05-30Status of Fast...

Status of Fast Reactor Technology Development in Korea

21-24 May 2013

Hyung-Kook Joo

The 46th IAEA TWG-FR Meeting, Wien, IAEA

IAEA TWG-FR, Wien, 21-24 May 2013 2

Outline

Korean Nuclear Power Program I

SFR Technology Development Program II

Conceptual Design of a Prototype SFR III

R&D Activities IV


Korean Nuclear Power Program I


Korea’s Energy & Oil Import

Year 2010

Status of Energy Supply in Korea

Energy Import : 252 Mtoe

(118 Billion USD, 27 % in total import)

Oil Import : 147 Mtoe

(69 Billion USD)

* Ref: BP (2011), Statistical Review of World Energy

* IEA, Energy balance of OECD countries 2011

96.6 % of energy was imported in 2010

Energy self-sufficiency (without nuclear power)

Energy self-sufficiency (including nuclear power)

Se

lf-s

uff

icie

ncy [

%]

20

40

60

80

100

120

140

160

3

Japan Germany

France U.S.

U.K. Canada

4

28

65

8

74

142

19 20

40

51

78

81

153

Korea

Korea’s Energy Consumption

* Ref: Korea Energy Economics Institute (2011)

Energy Consumption : 261 Mtoe


Installed capacity in 2012 was 82GWe (Nuclear share 25%)

Electricity generation in 2012 was 510 TWh (Nuclear share 30%)

Nuclear share is estimated to be 22.7% in 2027

“KEPCO in Brief," www.kepco.co.kr, Dec 2012 As of December 2012

Status of Electricity Generation

24.5

20.7 20.1

7.76.4

2.3

0

5

10

15

20

25

30

Coal Nuclear Gas Oil Hydro Alternative

Ca

pa

city

(G

We

)

Energy Resource

Installed Capacity

Coal

Nuclear

Gas

Oil

Hydro

Alternative

(30.0%)

(25.3%)

(24.6%)

(9.4%)

(7.9%)

(2.9%)

180.8

150.3

114.0

48.2

7.7 8.6

0

20

40

60

80

100

120

140

160

180

200

Coal Nuclear Gas Oil Hydro Alternative

Ge

ne

rati

on

(T

Wh

)

Energy Resource

Electricity Generation

Coal

Nuclear

Gas

Oil

Hydro

Alternative

(35.5%)

(29.5%)

(22.4%)

(9.5%)

(1.5%)

(1.7%)

http://www.kepco.co.kr/


Seoul

Wolsong

Hanbit

(yonggwang)

Hanwool

(Uljin)

As of February 2013

In Operation – 23 Units

• 19 PWRs (9 OPR1000)

• 4 PHWRs at Wolsong

Under Construction – 1 OPR1000

– 4 APR1400

Planned by 2024 – 4 APR1400

– 2 APR1500

NPP

Sites

Kori

Hanbit

Hanwool

Wolsong

Total

Generation

’12 (GWh)

37,174

46,479

39,382

27,292

150,327

Capacity

(MWe)

5,137

5,900

5,900

3,779

20,716

Current Status of NPPs

Under construction

In operation

Preparation for Construction

OPR1000

APR1400

APR1500

Kori


Fast Reactors

Option for LWR Spent Fuel Management

Why Fast Reactor?

’09년말 현재 10,761톤 누적

’16년부터 소내저장용량 포화

Domestic SF Cumulative

Onsite SF storage will be

saturated from 2016.

• Direct disposal

-A formidable construction cost and huge site for final disposal repository

- Radiotoxicity lasting over 300 thousand years

Reuse of Spent Fuel

• Safe management of SF through reduction of radwaste volume,heat load and radiotoxicity environment friendly

Cu

mu

lati

ve

Sp

en

t F

ue

l (M

TU

)

Year

Cumulative LWR Spent Fuel Inventory


SFR Technology Development Program II


SFR-Pyroprocess development plan authorized by KAEC on Dec. 22, 2008

SFR-Pyroprocess Development Plan

Gen IV

SFR

Electrical

Heater

7 MWt

Air cooler

FW pump

SG

PHTS

pump IHTS

pumpIHX

545.0 oC

390.0 oC

30kg/s

320.7 oC

526.0 oC

320.0 oC

503.1 oC

23

0.0oC

230.0 oC

Pump

Drain tank

Plugging

indicator

Cold trap

hot air out

Air stack

hot sodium in

AHX

cold sodium out

cold air in

PDRC

AHX

Expansion tankArgon

LSDT

DHX

IRACS

Air Blower

Active AHX

System Performance

Test

Standard Design

Detailed Design

Demonstration Plant

Mock-up Facility

(Nat. U, 10t/Yr)

Eng.-scale Facility (10t/Yr)

Prototype Facility

(100t/Yr)

Prototype Facility

Operation

Pyro-

process

Advanced Design

Concept

‘07 ’11 ’16 ’20 ’28 ’26

Licensing Technology Development

Metal Fuel Irradiation Test

Viability & Economics Licensibility Construction


Revised SFR Development Plan

BOP and

Component

Design

BOP /Component

Concept. Design

BOP /Comp

Design


SFRA

Organized on 16th of May, 2012

Affiliated organization of KAERI

Goal of SFRA : acquisition of design approval for prototype SFR

Background of Organizing SFRA

Phase change in SFR development program

From key technology development in the past

To overall system engineering including SFR system design and

optimization, design V&V tests, major component development etc.

In order to perform prototype SFR development efficiently and

consistently

Project Period : 2012 ~ 2020 (9 years)

SFR development Agency (SFRA)


Organization of SFRA

Ministry of Science, ICT and Future

Planning (MSIP)

Committee for Promotion of

SFR Development

National Research

Foundation of Korea (NRF)

Director Steering Committee

Advisory Board Executive Office

NSSS Design※

(KAERI)

Technology

Verification

(KAERI)

Fuel

Development

(KAERI)

BOP Design,

Component Design,

… (industry)

SFRA

Fund and Manage the SFR Development Project including NSSS, BOP,

Component Design, Development of Related Technology

※ : Performed by the Collaboration Program with Argonne National Laboratory


’92

’11 Basic Research

’97

’02

’20 ’07

KALIMER-150 Conceptual

Design

KALIMER-600 Conceptual

Design

Sodium integral effect Test Loop for simuLation

and Assessment

Specific Design

Approval

Prototype SFR

Advanced SFR

Concept ’28

’16

Goal

Construction of prototype SFR by 2028

Work Scope

Advanced design concept development

Design validation

Metal fuel development

BOP and component design

Proliferation resistant core without blankets

Metallic fuel

Enhanced safety with passive systems

Prototype SFR Development


Conceptual Design of a Prototype SFR III


Pool-type Reactor

150 MWe

Fuel : U-Zr -> U-TRU-Zr

Core I/O Temp. : 390/545 ℃

DHR System : PDHRS/ADHRS

2-loop IHTS/SGS

Single-wall tube SG

Superheated Steam Rankine Cycle

(SCO2 cycle option)

Key Design Features (Draft)

AHX #1

AHX #2FHX #1

FHX #2

Steam

Generator#1

Steam Generator

#2

DRAC Piping

IHTS Piping

Guard Vessel

IHTS Pump#1

IHTS Pump#2

Reactor Vessel

Core

Reactor Support


Conceptual Design-Core Design

Core Configuration of 150MWe Prototype SFR

Trade-off Study – 50/75/100/125/150/200/600 MWe

– Two Concerns to Power Level

Economy: Smaller Fuel inventory

TRU Fuel Irradiation Capability: Higher

Fast Neutron Flux

Prototype SFR Core – Power Level: 150 MWe

– Initial Core : Uranium Metal Fueled Core

Test and Demonstrate TRU Fuel

– will evolve into TRU Core

Determination of Power Level

Inner Core

Secondary control rod

Primary control rod

Reflector

B4C shield

IVS

Radial shield

33

7

3

102

60

138

78

Outer Core 90


Core U LTRU MTRU

EFPD / # of Batches [day / #] 290 / 5 290 / 5 290 / 5

# of Fuel Assembly (IC/OC) 33/90 33/90 33/66

Fuel Pin Diameter [cm] 0.74 0.74 0.70

P/D Ratio 1.125 1.125 1.189

Active Core Height [cm] 100.0 100.0 100.0

Lower Shield Height 90.0 90.0 90.0

Fission Gas Plenum Height[cm] 150.0 150.0 150.0

Enrichment (IC/OC) [w/o] 14.0 / 19.5 14.9 / 21.8 20.2/29.6

Fuel Loading Amount [Ton/GWe] 107.9 107.8 76.4

Charged Amount [kg] Heavy Metal 2205 2204 1576

TRU 0 438 415

MA 0 40 51

Fissile 397 237 195

Reactivity Swing [pcm] 1184 695 1493

Burnup [MWD/kg] Average 50.2 50.4 70.5

Peak 78.7 81.8 110.1

Fast N. Flux [x1015 n/cm2·sec] Average 0.98 1.18 1.41

Peak 1.54 1.87 2.22

Peek Fast N. Fluence [x1023 n/cm2] 1.95 / 1.93 2.37 / 2.36 2.83 / 2.71

Linear Power Density [W/cm] Average 104.7 105.1 129.7

Peak 180.0 178.6 219.7

Pressure Drop [MPa] 0.255 0.255 0.204

Cladding Midwall Temp. [oC] 645 645 645

Core Function

Core Design

Core Configuration

U / LTRU core MTRU core

<Core Performance>


Fluid System Design

Key Design Features

System Concept

System Heat Balance

System design concept from trade-off study

to enhance plant safety and to improve economic &

performance

PHTS – Pool-type

– Two PHTS pumps

– Four IHXs

IHTS – Two loops

– Two straight tube steam generators

– Two mechanical pumps on cold legs

– Passive protection of IHX from SG tube failure by cold

leg piping layout

DHRS: Passive (PDHRS) / Active (ADHRS)

Energy conversion system

– Superheated steam Rankine cycle


Loop Cover gas High

Low

Relief valve

SodiumAHX

DHX

Reactor head

Hot-leg

Cold-leg

Expansion

vessel

Plane A

Plane B

Plane C

x

y

z

Rx.

Support

wall

4.05 m *

9.65 m

4.15 m

2.5 m

6.65 m 7.65 m

2.5 m2.5 m

7.71 m

2.5 m

3.5 m

2.5 m

4.2 m

1.0 m

1.34 m

3.58 m

16.4

1 m

15.4

1 m

36.0 m

4.45 m

~ 2

5.1

2 m

22.7

m

1.7

55

m

3.0

8 m

2.0

3 m

1.54 m

2.18 m

2.18 m 4.2 m

Hot-leg (48.07 m)

- Vertical length : 19.99 m

- Horizontal length : 28.08 m

- Coaxial part : 5.79 m

Cold-leg (40.99 m)




2.5 m

1.85 m

2.85 m

1.0

5 m

( 10 inch SCH40 )

( 10 inch SCH40 ) 1.0

5 m

*: Distance from DHX center

0.5 m

Loop Cover gas High

Low

Relief valve

Sodium

DHX

Reactor head

Hot-leg

Cold-leg

Expansion

vessel

Plane A

Plane B

Plane C

x

y

z

Rx.

Support

wall

4.05 m *

9.65 m

4.15 m

2.5 m

6.65 m 7.65 m

2.5 m2.5 m

7.71 m

2.5 m

3.5 m

1.0 m

1.34 m

3.58 m

16.4

1 m

15.4

1 m

39.0 m

4.45 m

24.6

8 m

22.7

m

1.7

55

m

2.0

3 m

1.10 m

Hot-leg (48.07 m)




Cold-leg (40.99 m)




2.5 m

1.85 m

2.85 m

1.0

5 m

( 10 inch SCH40 )

( 10 inch SCH40 ) 1.0

5 m

*: Distance from DHX center

0.26 m

FHX

2.5 m

6.38 m

3.3

2 m

6.38 m

1.50 m

EM

P

Decay Heat Removal System

Safety-grade DHR system

– Ultimate heat sink for DBA

PDHRS (Passive Decay Heat Removal System)

– Two independent heat removal loops

– DHX, AHX, Expansion tank, Air dampers

– Use of natural circulation of Sodium & Air

– Operation

Emergency heat rejection w/o operators’ action

ADHRS (Active Decay Heat Removal System)

– Two independent heat removal loops

– DHX, FHX, Expansion tank, EMP, Air blower, Air dampers

– Use of forced (and/or natural) circulation of Sodium & Air

– Operation

Scheduled & Emergency operation

50% heat removal capacity for SBO condition

PDHRS loop configuration ADHRS loop configuration

DHX AHX FHX


Mechanical Structure Design

Core

Primary Pump

IHX

Steam Generator

AHX

FHX

DHXUpper Internal Structure

Double Vessel

- Reactor Vessel

- Containment Vessel

Main Design Features

- Simple reactor enclosure system

Uniform vessel thickness

No penetrations and no attachment on

vessel

- Double vessels (Reactor and Guard vessels)

- Skirt type core support structure excluding weld

joint with reactor vessel

- Significantly reduced IHTS piping length using

9Cr-1Mo-V steel

- Seismic Isolation Design for Reactor Island (Rx

Bldg, Aux Bldg and Wastage/Maintenance Bldg)

. – ASME BVP III division 5 is applied to high

temperature design

Key Design Features


Evaluation of its safety functions of the PGSFR on DBEs

MARS-LMR Nodalization

Safety Evaluation of the PGSFR

1 10 100 1000 10000

300

350

400

450

500

550

600

650

700

750

DHRS heat removal ~ power

Reactor trip

Reactor Vessel Leak

initiation

Reactor Vessel Leak

Core inlet temperature

Core outlet temperature

Co

ola

nt T

em

pe

ratu

re,

oC

Time, s

16.3

17.2168

1.1

CORE

180175

178170

130135

160165

16.8

17.658

InletPlenum

18.486

0.0

1.26

2.0

Upper Hot Pool

0.3

Hot PoolRiser

7.896

10.946

137

150

145MP

140

Buffer

265,275285,295

260,270280,290

0.6096

Buffer

120

115MP

110

100

105

200

235

107

215

205 210

220

225

230

237

IHX

240

300,350

pipe

310,360

pipe

390,395

sngljun320,370

pipe

345

pipe

380

pump

385

pipe

0.3

5.2

245242

250247

DHX

5.3

5.554

7.055

7.1557.4455

11.733

10310.946

12.257

185190

195

11.257

11.84

6.326.9

7.0

315,365

branch

11.307 11.307

10.0

1.755

237

240

6.801

345

pipe

385

pipe

SG

325,375

sngljun333

sngljun

335

pipe

338

sngljun

340

pipe

28m

610

625

615 825

613

617

620

835

820

AHX

Reactor Vessel Leak Transient of Over Power

Safety evaluation of the PGSFR

- Representative transients of TOP, LOF, LOHS,

Primary pipe break, reactor vessel break have been

evaluated with the MARS-LMR code.

- PGSFR design satisfied the safety criteria with an

appropriate margin.

ATWS transients of UTOP, ULOF, ULOHS have

been evaluated with the MARS-LMR code.

The 6 sub-channel flow blockage was also

evaluated and ensured to satisfy the safety limits.

The performance of the DHRS during typical DBE

and BDBE has been checked to show that the

DHRS design has ability to prevent the fuel rod

heat-up due to inherent natural circulation function.

Further work

- Safety analysis for final conceptual design will be

followed. Typical Temperatures


R&D Activities IV


STELLA Program

STELLA (Sodium Test Loop for Safety Simulation and Assessment)

– Phase 1: STELLA-1

• Performance evaluation of key sodium components

• Heat exchanger design codes V&V

– Phase 2: STELLA-2

• Verification of dynamic plant response after reactor shutdown

• Construction of test DB to support specific design approval for the prototype SFR

Schedule


Overall Characteristics of STELLA-1

Main test loop

– Test components

• Sodium-to-sodium heat exchanger (DHX)

• Sodium-to-air heat exchanger (AHX)

• Mechanical sodium pump (PHTS pump)

– Electrical loop heaters, EM pumps, Flow meters,

Expansion tanks, Sodium storage tank

Sodium purification system

– Cold trap, Plugging meter, etc.

Auxiliary Systems

– Gas supply & Vacuum system

– Power supply system

– Fire protection system

Working fluid Liquid sodium Total electric power 2.5 MW

Total sodium inventory ~ 18 ton Heat capacity of HXs 1.0MW

Design temperature 600oC Design pressure 10 bar

Max. flowrate for HX test 10 kg/s Max. flowrate for Pump test 125 kg/s

STELLA-1 Layout

Overall Size (W×L×H): 15m×8m×22m

Main Characteristics


Tube-side (Na)

Shell-side (Air)

Forced Circulation Flow (kg/s)

3.5 3.9 10.0

Forced-draft

Flow (kg/s)

3.5 ○ ○ ○

5.0 ○ ○ ○

Natural-draft N/A ○ ○ ○

Test Scope & Conditions

Tube-side (Na)

Shell-side (Na)

Forced Circulation Flow (kg/s)

3.5 3.9 10.0

Forced

Circulation

Flow (kg/s)

3.6 ○ ○ ○

5.3 ○ ○ ○

10.0 ○ ○ ○

< Matrix for DHX performance test >

AHX Shell-side

DHX Shell-side

Forced-draft Flow (kg/s) Natural-

draft 3.5 5.0

Forced

Circulation

Flow (kg/s)

3.6 ○ ○ ○

5.3 ○ ○ ○

10.0 ○ ○ ○

(Air)

(Na)

< Matrix for AHX performance test >

< Matrix for DHX-AHX natural circulation test >

DHX performance test

- Nominal & Transient conditions by flowrate change

Function of sodium Peclet number

- Heat transfer & Pressure drop characteristics

AHX performance test

- Nominal & Transient conditions by flowrate change

Function of air Reynolds number

- Heat transfer & Pressure drop characteristics

DHX-AHX natural circulation test

- DHX shell-side sodium: Forced circulation

- AHX shell-side air: Forced- & Natural-draft

- Natural circulation flowrate inside loop piping

Mechanical sodium pump test

- Sodium temperature: ~350oC

- Constitution of test matrix

N/NR : 25 ~ 100% (every 10%)

(Q/QR) : 5 ~ 125% (every 5%)


S-CO2 Brayton Cycle

Development of S-CO2 Brayton cycle Integral Experiment Loop (SCIEL)

Objectives

- Develop the S-CO2 Brayton Cycle Integral Experiment

Loop(SCIEL) and verify the characteristics of the S-CO2

recompression cycle and operation technologies

• SCIEL: 550oC-20MPa turbine inlet temperature, heat

input 730kW, net power generation 200kWe

SCIEL Schedule

- 2013: Installation starting of SCIEL

• Step I: Compressor Performance Test Loop (‘13.June)

• Step II: Simple Cycle Test Loop

• Step III: Simple Recuperated Cycle Test Loop

- 2014: Completion of SCIEL operation test

• Step IV: Recompression Cycle Test Loop

Schematics and major specification of the SCIEL


Development of under-sodium ultrasonic waveguide sensor module and

Performance tests in sodium

C-Scan test of waveguide sensor in sodium R&D Contents

- Development of under-sodium ultrasonic waveguide sensor

• Be and Ni coating for the well-developed beam profile

generation in sodium

- Performance tests of under-sodium waveguide sensor

modules in sodium

• Ultrasonic wave propagation and sensitivity test in sodium

• C-scan imaging resolution test

- Design and construction of sodium wetting test facility

• Sodium wetting test chamber

Operation in high temperature condition (up to 350 ℃)

Sodium test tank, sodium storage tank, enclosure box

Ar purging system

• Steel enclosure structure for the protection of sodium fire

accident for the long duration operation

t

Under-Sodium Test

Test Target 2 mm C-Scan Image

Sodium wetting test facility with enclosure structure

Ultrasonic Transient Signal (10 m)

Ultrasonic Waveguide Sensor for Under-sodium Viewing


Tube fabrication process - Intermediate heat treatment effects

Increase of hardness in modified fabrication process

- Final heat treatment effects

Increase of UTS and YS in modified fabrication process

Heat treatment process - Evaluation of intermediate heat treatment conditions

IHT temperature : 700oC & 720oC

IHT time : 10, 30, 60min

Selection of IHT conditions for cold working process

- Evaluation of final heat treatment conditions

Normalizing temperature : 950 ~ 1100oC

Tempering temperature : 700 ~ 800oC

Selection of candidate FHT conditions

Development cladding tube fabrication process

Fuel Cladding Development

Hardness change with IHT Tensile properties with FHT

Tensile strength with FHT Mechanical property diagram

M.T

1st drawing

2nd 3rd

4th 5th

M.T

2nd 3rd

4th

1st drawing

Normalized at 1050oC for 30min 650oC tension test

UTS of AR HT9

YS of AR HT9

+12.3%

+16.3%

Hardness change with drawing

UTS change with FHT


Fuel Fabrication and Performance Evaluation

Fuel casting technology and Fuel irradiation test in HANARO

Fuel irradiation test in HANARO - Irradiation of U-Zr-(Ce) fuel up to 3 at%

- Completion of Irradiation

•2010.11.15~2012.1.5

-Nondestructive Test

•γ- scanning : 12 rodlets completion

-Destructive Test

•Fission gas release measurement : 3 rodlets

•Microstructure by OM, SEM etc

Lower fuel rodlet

Upper fuel rodlet

Coolant

Fuel slug

Hf tube

A-AB-B Neutral plane of core

Bottom view

Fuel slug casting - Modification of casting process and equipment:

melting and casting temperature, preheating

temperature of melt distributer, and melting batch

size

- Modification results

• Fuel dimensions: Φ5-L300 mm

• Deviation of Zr content and fuel density:

0.1~0.2wt%, 0.1~0.2g/cm3

• Feeding rate of melt to mold: 95%

• Fuel loss: 0.1%

U-10Zr fuel slugs

Gamma radiography


Development of Codes and Verification

Reactor Physics Experiments

- Collaboration with IPPE (Institute for Physics and Power Engineering) is going on.

- Three critical assemblies were constructed :

BFS-73-1, BFS-75-1

BFS-76-1A for TRU burner concept (without blanket, deep insertion of CR at BOEC)

- Representative Experiment for Prototype SFR

BFS-109-2A : high enriched uranium fuel (U-10Zr)

Thermal-Hydraulic Validation Test

- Three test activities are planned

Validation tests of reactor core thermal-hydraulic characteristics: Test DB is established and

Conceptual facility design has been developed

performance tests of the finned-tube sodium-to-air heat exchanger (FHX): Concept of finned

tube heat exchanger has been developed for verification of FHX thermal sizing.

V&V of steam generator design code: to be conducted untill 2016


Numerical codes were developed to estimate energy release and core

expansion behavior in the case of CDA

Scoping analysis of Severe Accident

Analysis of energy release from CDA

- Energy release from CDA was estimated for

KALIMER-150 core using CDA-ER code

based on the Bethe-Tait methods

- The influences of Doppler effect on the power

excursion were estimated, and the obtained

results were used as initial conditions for core

expansion analysis

Analysis of core expansion behavior

- Numerical code was developed to investigate

core expansion behavior during the power

excursion of KALIMER-150 core

- The transient behaviors of a vaporized core

expansion and its effects on coolant were

analyzed through numerical computations

He

core

sodium

18m

7.4m

0

200

400

600

800

0.000 0.005 0.010 0.015

rad

ial ve

locity

(m/s

)

time (s)

(b) Core expansion velocity (c) Computation domain

0

2

4

6

8

10

12

14

3.0 3.5 4.0 4.5

time (msec)

E (GJ)

Power (TW)

Pressure (GPa)

(a) Released energy from CDA


Summary

Long-term Advanced SFR Development Plan was revised by KAEC

in November 2011

– Specific design by 2017

– Specific design approval by 2020

– Construction of a prototype SFR by 2028

Activities for development of an Advanced SFR include

– Conceptual core design from U core to MTRU core

– Conceptual design of fluid system & mechanical structure

– Development of metal fuel

– Under sodium viewing for in-service inspection

– STELLA for major components test and integral effect test including decay heat

removal system

– Reactor physics experiment for U-Zr core

– Evaluation of representative DBEs and BDBEs using MARS-LMR code

Overall System Design for Prototype SFR is started this year

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