HTTR Test Reactor - Atoms for Peace and Development · HTTR Test Reactor Japan Atomic Energy ......

HTTR Test Reactor

Japan Atomic Energy Agency

Training Course on

High Temperature Gas-cooled Reactor Technology

October 19-23, Serpong, Indonesia

p.2

HTTR Outtline

Major specification

Thermal power 30 MW

Fuel Coated fuel particle / Prismatic block type

Core material Graphite

Coolant Helium

Inlet temperature 395C

Outlet temperature 950C

Pressure 4 MPa

Containment vessel

Reactor pressure vessel

Intermediate heat

exchanger (IHX)

Hot- gas duct

HTTR Graphite-moderated and helium-cooled VHTR

Fuel Rods Graphite

Block

First criticality : 1998 Full power operation : 2001 50 days continuous 950oC operation : 2010 Loss of forced cooling test at 9MW : 2010

p.3

HTTR History

Reactor physics

Very High Temperature Reactor Critical Assembly (VHTRC)

Thermo-hydraulics Start Construction

Work

Installation of

Reactor Pressure

Vessel

Conceptual design

System integrity design

Basic design

Detail design

Application and

permission of construction

Construct ion

F i rs t cr i t i ca l i t y

Reactor thermal power（30MW）

Reactor outlet coolant

temperature 850℃

Reactor outlet coolant

temperature 950℃

850℃/30 days operation

950℃/50 days operation

Safety demonstration test

1973

1969 ～

1980

1974

～

1984

1981 ～

1985 ～

1988

1989

1990

1991 ～

1997

1998

2001

2002

2004

2007

2010

Research development and design

～

Construction of Reactor

Establishment of fundamental technology

Start of loss of forced

cooling test

Research and development

Fuel / material

In-pile Gas Loop(OGL-1)

Outside and inside of the HTTR

HTTR is the first HTGR in Japan. Reactor thermal power : 30MW Reactor inlet/outlet coolant temperature:

395/850,950C Primary coolant pressure: 4MPa Reactor core height/diameter: 2.9m/2.3m Average power density: 2.5W/cc Uranium enrichment: 3~10%(average6%)

H T T R

Long term high temperature operation

Construction

2013 New regulation standard issued

Confirmation of conformity of

the standard toward restart

p.4

HTGR Technology Developed in HTTR Project

Japanese HTGR technologies are front runner in the world.

High temperature resistant metal, Hastelloy XR (Mitsubishi material)

Experiences of HTTR design, construction, operation

(MHI, Toshiba/IHI, Hitachi, Fuji Electric, KHI, etc.)

Fuel (Nuclear Fuel Industry)

Graphite, IG-110 (Toyo tanso)

Graphite core component in the HTTR

Intermediate

heat exchanger

(IHX)

World highest quality graphite (isotropic, high density)

Japan can construct HTGR by domestic technology.

Hastelloy XR is applicable at 950C as the nuclear structural material . IHX can deliver hot helium gas at 950C to outside the reactor pressure vessel.

Coated fuel particle

Fuel compact

A lot of technical data of HTTR was accumulated. Optimum design of commercial HTGR can be conducted by only Japanese technology.

Ceramics coating is stable for long-term. (3 times higher burnup than LWR)

Ceramics coating layer retains fission products inside the coated fuel particle at extreme low leak level.

High strength, high thermal conductivity, irradiation resistance p.4

p.5

Bird’s-eye View of HTTR Reactor Building

Control room

Spent fuel storage pool


Reactor containment vessel

Intermediate heat

exchanger

Air cooler

Primary pressurized

water cooler

Secondary pressurized

water cooler

Fuel handling machine

Reactor core

p.6

Major Specifications

Thermal power

Average power density

Outlet coolant temperature

Inlet coolant temperature

Primary coolant pressure

Direction of coolant flow (core)

Moderator / Reflector

Core height

Core diameter

Fuel

Uranium enrichment

Fuel element type

Pressure vessel

Containment vessel

30MW

2.5MW/m3

850C/950C

395C

4MPa

Downward

Graphite

2.9m

2.3m

Low enriched UO2

3 10% (Ave. 6%)

Prismatic block

2 1/4Cr-1Mo steel

13m(H)×6m(ID)

Steel containment

30m(H)×18.5m(ID)

System Configuration

p.7

RPV

p.8


Design pressure 4.8 MPa

Design temperature 440 C

Inner diameter 5.5 m

Height 13.2 m

Thickness

Cylinder, Bottom dome 122 mm

Top dome 160 mm

Stand pipe

Control rod 16

Irradiation 5

Others 13

Material 2.25Cr-1Mo steel

Y. Tachibana et al., JAEA-Technology 2009-063 (2009).

Primary Pressurized Water Cooler (PPWC)

p.9

Primary pressurized water cooler

type Vertical U-tube

Design pressure 4.8MPa

Design temperature 430 C

Flow rate

Primary coolant Max. 45t/h

Cooling water Max. 640t/h

Temperature

Primary coolant Max. 950 C

Cooling water Inlet 150 C /outlet 190 C

Tube

Number 136

Outer diameter 25.4 mm

Thickness 2.6 mm

Outer diameter of shell 2.1 m

Overall height 7.1 m

Material 2.25Cr-1Mo steel (shell)

Stainless steel (tube)

T. Furusawa et al., Nucl. Eng. Des., 233113-124 (2004).

p.10

Intermediate Heat Exchanger (IHX)

Intermediate heat exchanger

Type Vertical helically-coiled

counter-flow heat

exchanger

Heat transfer rate 10 MW

Design pressure 4.8 MPa (shell)

0.3 MPa (tube)

Design temperature 430 C (shell)

955 C (tube)

Shell

Outer diameter 2 m

Overall height 11.0 m

Tube

Number 96

Outer diameter 31.8 mm

Thickness 3.5 mm

Material 2.25Cr-1Mo steel (shell)

Hastelloy XR (tube)

T. Furusawa et al., Nucl. Eng. Des., 233113-124 (2004).

p.11

Gas Circulator

IHX HGC PPWC HGC

Type Vertical centrifugal/gas bearing type

Number 1 3

Flow mass rate (max.) 15t/h 15t/h

Head (max.) 794kPa 1,080kPa

Allowable working press. 4.8MPaG 4.8MPaG

Allowable working temp. 430C 430C

Filter type Sintering material

Motor type Cage type induction motor

Motor power 190kW 260kW

Number of revolutions 3,000-12,000rpm 3,000-12,000rpm

Frequency converter

type

Thyristor-converter

T. Furusawa et al., Nucl. Eng. Des., 233, 113-124 (2004).

p.12

Containment Vessel

Containment type Steel containment

Maximum service pressure 0.4MPa

Maximum service temperature 150C

Major size

Inner diameter 18.5m

Overall height 30.3m

Body thickness 30mm

Top lid thickness 38mm

Refueling hatch diameter 8.5m

Maintenance hatch diameter 2.4m

Air lock diameter 2.5m

Free volume 2800m3

Material Carbon steel

Leak rate 0.1%/day at R.T.&

0.36MPa

C/V solely stands on the base mat without any support

S. Saito et al., JAERI-1332 (1994).

p.13

Purification System

Type Number Helium flow rate

Pre-charcoal trap Vertical cylinder 1 200kg/h

CuO fixed bed Vertical cylinder 2 200kg/h

Molecular sieve trap Vertical cylinder 2 200kg/h

Cold charcoal trap Vertical cylinder 2 50kg/h

N. Sakaba et al., NED, 233, 147-154 (2004).

p.14

Helium Storage and Supply System

Storage tank Number 6

Tank volume About 18m3/tank

Capacity About 1320kg (total)

Max. pressure 8.6MPa

Supply tank Number 1

Tank volume About 10m3/tank

Capacity About 110kg (total)

Max. pressure 8.6MPa


p.15

Instrumentation and Control

K. Saito et al., Nucl. Eng. Des., 233, 125-133 (2004).

p.16

Nuclear Instrumentation


p.17

Reactor Power Control System

Operation mode selector

Rated power operation

High temperature operation

Safety demonstration test operation

Reactor power control system

Power control

Reactor outlet temperature control

Detector

Neutron flux

Reactor outlet coolant temperature


p.18

Temperature Measurement in Reactor


p.19

Reactor Protection and Control Systems

Reactor scram signals in the reactor

protection system

WRMS High

PRMS High

IHX primary coolant flow rate Low

PPWC He flow rate Low

Primary coolant radioactivity High

IHX outlet primary coolant temperature High

Reactor outlet temperature High

Core differential pressure Low

PPWC pressurized water flow rate Low

Primary / pressurized water differential

pressure

High

Low

Primary / secondary He differential

pressure

High

Secondary He flow rate Low

Seismic acceleration High

Plant control system

Reactor inlet coolant temperature control

IHX primary coolant flow rate control

PPWC primary coolant flow rate control

Primary He pressure control

Primary / secondary He differential pressure

control

Primary / pressurized water differential pressure

control

Pressurized water temperature control


p.20

Control Rod

Control

rod

Type Double circular cylinders

with lid and vent

Number 16 pairs (32 rods)

Total length 3.1m

Outer dia. 113mm

Inner dia. 65mm

Sleeve material Alloy 800H

Neutron

absorber

Outer dia. 105mm

Inner dia. 75mm

Material Sintered compact of

B4C/C

Excess

reactivity

Temp. loss and FP buildup 0.088k/k

Burnup 0.043 k/k

Reactivity margin 0.018 k/k

Uncertainties 0.016 k/k

Total 0.165 k/k

Shutdown

margin

CR controllable reactivity > 0.18 k/k

CR shutdown margin > 0.01 k/k


p.21

Fuel Failure Detection System

Type Precipitator

Number 2

Sampling nuclide 88Kr, 138Xe

Flowrate 0.5kg/h/channel

Temperature up to 50 C

• Fuel failure detection system detects the

failure of a CFP by detecting short life FPs,

such as Kr-88 asn Xe-138, which are

gathered with precipitating wiring.

• Helium gas of the outlet core from seven

regions are transferred to the line selector.

p.22

Vessel Cooling System

Line 2

Removal

heat

Rated operation <0.6MW / 2 Lines

In accident >0.3MW / 1 Line

Cooling pipe inlet water

temperature

44C

Available working

temperature

90C

Water flow rate 86t/h

• VCS is used as a residual heat removal

system when the forced helium circulation can

not maintained due to failure of concentric hot

gas duct.

• VCS is an engineering safety system so that

two independent complete sets are provided.


p.23

Auxiliary Cooling System

Line 1

Removal heat 3.5MW

Available working

temperature

Cold leg 430C

Hot leg 980C

Primary coolant mass flow rate 3t/h (one GC)

4.3t/h (two GCs)

• Auxiliary cooling system

automatically starts up at reactor

scram.

• ACS is an engineering safety

system so that a pair of GCs and

water pumps are provided.


p.24

Support Structures for RPV, IHX and PPWC

• RPV is supported by skirt, stabilizer at cylinder and vibration control beam at standpipe.

• IHX and PWC is supported by constant hanger and snubber.

Snubbers

Hangers Skirt

Stabilizer

Vibration Control beam


p.25

HTTR Operation Experiences (1/2)

• Unexpected temperature rise was observed in upper shielding.

• The temperature rise is due to bypass flow which flow upwards along standpipe wall


• Unexpected temperature rise was observed in upper shielding.

• The temperature rise is due to bypass flow which flow upwards along standpipe wall

• To reduce the bypass flow, flow resistance in main flow path is reduced by adding holes on the side of guide pipe and top plate of the CR guide tube.

• The gaps for the CR drive mechanism is sealed with synthetic rubber gaskets.

p.26

HTTR Operation Experiences (2/2)

• Unexpected temperature rise was observed in center part of core support plate.

• The temperature rise is due to gap flow from core hot plenum to inside of concentric hot gas duct.


• Detail calculation was conducted considering the gap flow to confirm that the temperature would not exceed the design temperature limit.

p.27

High Temperature Continuous Operation

To establish fundamental technologies of HTGR To demonstrate stable high temperature heat supply to a heat

application system

Purpose

Evaluation of core physics

Evaluation of fuel performance (FP retention)

Evaluation of impurity control technology in helium coolant

Evaluation of IHX performance

Evaluation of structural integrity of components

Accumulation of operation and maintenance technologies

p.28

Operation Record

0

10

20

30

40

1/4 1/11 1/18 1/25 2/1 2/8 2/15 2/22 3/1 3/8 3/15 3/22

Date　（month/day）

The

rmal

pow

er （

MW

)

50 days high-temperature operationHigh-temp. operation for 50 days

Control room

Date (month/day/2010)

50-days continuous operation

Rea

cto

r p

ow

er [

MW

t]

Date in 2010

K. Takamatsu et al., JAEA-Research 2010-038 (2010).

p.29

Main Results (1/2)

Long stability of condition

Flow rate and temperature of coolant were kept very stable.

Parameters Fluctuation

Nuclear power ±0.4 %

Coolant temp. About 3 oC

Coolant flow ±0.5 %

High controllability of excess activity

Insert depth of control rod was kept shallow and mostly constant.

Good confinement of FPs

Fractional release of Kr88 from coated fuel was kept very low.

p.30

Main Results (2/2)

Stable shielding

Temperature of shielding concrete around RPV was kept stable.

Low impurity of coolant

Impurity such as water, CO and H2 in the coolant was kept low .

High integrity of IHX

High temperature heat could be stably supplied through IHX.

p.31

Fuel Performance (1/3)

to

auxiliary

cooling

system

Reactor

PPWC

IHX

to SPWC

By-pass line

HGC

HGC

To pressurized water cooling system

(1) ionization chamber of PCR instrumentation

(3) the primary coolant sampling

Precipitators

(2) FFD system

Measurement of the radioactivity in the primary cooling system is performed by

(1) The primary coolant radioactivity (PCR) instrumentation of the safety protection system

(2) The fuel failure detection (FFD) system

(3) The primary coolant sampling system


p.32


Radioactivity in the primary cooling system was below the lower detection limit of about 14Bq/cm3

Detector: gamma-ray detection

0

25

50

75

100

1e-6

1e-5

1e-4

1e-3

1e-2

1/5 1/15 1/25 2/4 2/14 2/24 3/6 3/1610-6

10-4

10-3

10-2

10-5

104

103

102

10

1

100

75

50

25

0

Reactor power

Lower detection limit

Radioactivity in the primary cooling system

Reacto

r po

wer (%

) R

adio

acti

vity

(B

q/c

m3)

Date in 2010

1/5 1/15 1/25 2/4 2/14 2/24 3/6 3/16

(1) The primary coolant radioactivity (PCR) instrumentation of the safety protection system

p.33


Precipitator(A)

FFD system

Precipitator(B)

１

7 2

3

4 5

6

(2) The fuel failure detection (FFD) system

No significant differences of counting rate were observed among 7 regions.

Detector: NaI (Tl) scintillation counter 1.05

1.03

1.05

0.97

0.95

1.07

1.04

1.08

0.98

0.96

Rat

io o

f co

un

tin

g ra

te

0.88

0.92

0.96

1.00

1.04

1.08

FFD-1/2 FFD-3/2 FFD-4/2 FFD-6/5 FFD-7/5

K. Takamatsu et al., JAEA-Research 2010-038 (2010).

p.34


1/4 1/11 1/18 1/25 2/1 2/8 2/15 2/22 3/1 3/8 3/15 3/22

(3) The primary coolant sampling system

Kr concentrations were lower than about 0.1 Bq/cm3

10-4

10-5

10-6

10-7

10-8

10-9

HTTR

50-days continuous operation

1/4 1/18 2/1 3/1 3/15 2/15 3/22

Date in 2010

(R/B

) o

f 8

8K

r

Ref 1: T. Tochio et al., JAEA-Technology 2010-038 (2010). Ref 2: K. Takamatsu et al., JAEA-Research 2010-038 (2010).

p.35

Safety Demonstration Test

Reduction of primary coolant flow rate

Increase of reactivity by withdrawal of the central pair of control rods

Reactor

Primary Pressurized

Water Cooler

Core

Gas Circulators

Control Rods

Fuel Region Reflector Region

Core

Reactivity Insertion Test Coolant Flow Reduction Test

from partial reduction to total loss

p.36

Reactivity insertion test – CR withdrawal test -

Test conditions

Reactor power 30% - 80%

(30%, 50% and 60% tests has been conducted)

Control rods to be withdrawn Central pair

(16 pairs (total))

Withdrawal speed 1 and 5 mm/s

Withdrawal distance 10mm - 40mm

Reactor power control system Disabled


Initial Below 850℃

During Test Below 950℃

p.37

CR withdrawal test

Arrangement of control rods

Permanent reflector block Replaceable reflector

block

Fuel block

Control rod

Reserved shutdown pellets insertion hole

Central pair of control rods

Irradiation test column

p.38

Coolant flow reduction test

Test conditions

Reactor power 30% - 100%

(30% and 60% tests has been conducted)

Gas circulators to be stopped 1 or 2 (out of 3)





Scram set point of flow rate

93% → 27%

Reactor

Primary Pressurized Water

Cooler

Core

Gas Circulators

p.39

Loss of Forced Cooling Test

Auxiliary cooling system

Secondary pressurized water cooler (SPWC)

Reactor

Vessel cooling system (2 units)

Loss of forced cooling (LOFC) & Loss of vessel cooling (LOVC) simulation of station blackout

Upper panel

Side panel

Heat removal adjustment

panel

Water cooling tube

Upper radiation shielding

Lower panel

RPV

Radiation shielding

Thermal reflector plate

Vessel cooling system (VCS)

Cooling water

： Stop of circulator and pump

Primary pressurized water cooler (PPWC)

Intermediate heat exchanger (IHX)

Heat

Core is cooled from the outside by radiation and natural convection.

p.40


Test conditions

Reactor power 30%, 80%, 100%

Gas Circulators to be Stopped 3 (All the GCs)


Reactor Outlet Coolant Temperature



Reactor scram due to decrease of primary coolant flow rate Bypassed during test

p.41


Loss of forced cooling test (LOCF test) : Stop of all circulators in primary circuit

Increase of reactor core temperature

Reactor power and fuel temperature

remain in stable condition

Stop of all primary circulators flowrate of primary coolant : 100% → 0%）

Decrease of reactor power due to negative reactivity feedback

（ resonance absorption of neutron in U238）

41

Test Condition

• Initial reactor power 30% (9MW) • Without scram (no movement of CR)

Test Results

Elapsed time (hr) 0 1 2 3 4 5 6

Stop of circulator

Core flow rate Test result

Reactor power

Test result

Peak fuel temperature

Analytical result

100

50

0

30

15

0

1600

800

0

Flo

w r

ate

(%)

Pow

er

(%

)

Tem

p.

(oC

)

Dec.21, 2010 Test Date

p.42

Loss of Vessel Cooling Test

Test conditions

Reactor power 30%

Gas circulators to be stopped 3 (All the GCs)

VCS pumps to be Stopped 1 or 2 (VCS as an engineered safety feature has two pumps)



Initial about 320℃

Reactor scram due to decrease of primary coolant flow rate

Bypassed during test

HTTR Test Reactor - Atoms for Peace and Development · HTTR Test Reactor Japan Atomic Energy ......

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