Re evaluation of Maximum Fuel Temperature of the …€evaluation of Maximum Fuel Temperature of the HTTRat Normal Operation Hirofumi OHASHI Nuclear Hydrogen and Heat Application Research

Re‐evaluation of Maximum Fuel Temperature of the HTTR at Normal Operation

Hirofumi OHASHI

Nuclear Hydrogen and Heat Application Research CenterJapan Atomic Energy Agency (JAEA)

IAEA Technical Meeting on on Re‐evaluation of Maximum Operating Temperatures and Accident Conditions for High Temperature Reactor Fuel and Structural Materials, 10‐12 July 2012, Vienna, Austria

Outline

1. HTTR overview

2. Evaluation of HTTR fuel temperature at design stage Evaluation method Evaluation results

3. Re‐evaluation of fuel temperature 1st modification using the operation data of the rise‐

to‐power test up to 850oC 2nd modification using new analysis model 3rd modification using the 950˚C operation data

4. Related future tests using HTTR

5. Summary1

High Temperature Engineering Test Reactor (HTTR)

Major specificationFuel Low enriched UO2

Uranium enrichment 3~10wt% (avg. 6wt%)Core PrismaticFuel assembly Pin‐in‐blockModerator GraphitePrimary coolant HeliumThermal power 30 MWInlet temperature 395COutlet temperature 850oC / 950C (Max.)Primary coolant pressure 4 MPaPrimary coolant flow rate 12.4 / 10.2 kg/s

Containment vessel

Reactor pressure vessel

Intermediate heat

exchanger(IHX)

Hot‐ gas duct

HTTR

Fuel Rods Graphite Block

First criticality : 1998 Full power operation (850oC/30 MWt): 2001 950oC operation at full power: 2004 50‐days continuous operation at high outlet

temperature (950oC/30MWt) : 2010

2

Structure of Fuel Assembly

3

Fuel compact Fuel rod Fuel assembly

26mm

39mm

8mm

Coated fuel particle

920μmLow density PyC

SiCHigh density PyC

Fuel kernel,600μm

34mm

Graphite sleeve

Fuel compact

PlugDowel pin

Dowel socket360mm

Fuel handing hole

580mm

Reactor pressurevessel

Core supportplate

Primary heliumgas tube

Core restraintmechanism

Fuel assembly

Replaceablereflector

Permanentreflector

Control rodstandpipe

Control rodguide block

Permanentreflector

Reactor pressurevessel

Core restraintmechanism

Fuel assembly


Reactor Core Structure

4

Evaluation Method of HTTR Fuel Temperature

FLOWNETIn‐vessel thermal and hydraulic

analysis code

Fuel, control rod, core component, core internal structure design data

Power density and neutron fluence distributions

Nuclear design code

TEMDIMFuel temperature analysis code

Coolant flow rate and coolant temperature distribution

Fuel temperature

(1)

(2)

(3)

5

Top shieldingReplaceable reflector

Fuel assembly

Replaceable reflectorHot plenum

: Gap between each block: Control rod column flow path : Fuel channel

In‐vessel Thermal and Hydraulic Analysis Code “FLOWNET”

6

One‐dimensional model using nodes and branches The flow channels are represented by node, and the nodes are connected by branch. The heat transfer between the branches are taking into account. Flow paths: the main coolant flow, the bypass flow in the inter‐column gaps, the leakage flow through the

permanent reflectors and the cross flow in the horizontal interface gaps of the hexagonal graphite blocks

Ref: S. Saito et al., “Design of High Temperature Engineering Test Reactor (HTTR)”, JAERI 1332 (1994).

: Fuel temperature (℃)

: Coolant inlet temperature (℃)

: Temperature rising (℃)

: Hot spot factor (‐)

: Random factor (e.g., manufacturing tolerances, uncertainties on physical properties) (‐)

: Systematic factor (e.g. total reactor power, coolant flow rate, inlet coolant temperature )(‐)

Gap

Graphite sleeve

Fuel compact

Annular flow path

Fuel rod

Graphite block

Coolant

Coolant

A

A‐Acrosssection

Estimatedpoint

NT

N1T

ginT

A

N2T N

5TN4TN

3T

5

1iii

Gasin

FUEL ∆TFTT

Fundamental equation

n(i)

1j

m(i)

1k

2ki,ji,i fr1fsF

FUELTGasinT

iFi∆T

ji,fs

ki,fr

Fuel Temperature Analysis Code “TEMDIM”

7

Two dimensional cylindrical model based on the power distribution including local power peaking, coolant flow distribution including

redistribution in the fuel column and hot spot factors

i= 1 : Coolant temperature rising2 : Film temperature rising3 : Temperature rising in graphite sleeve4 : Fuel compact‐graphite sleeve gap temperature rising5 : Temperature rising in fuel compact

Hot plenum

(Unit : %) Outlet Inlet

Top plenum 98.9

Control rod

coo

ling channe

l

Fuel coo

ling channe

l

Inne

r colum

n gap

100 100

5.5

5.5

5.5

5.7

7.2

8.8

8.2

7.1

99.30.7

0.4

92.2

92.1

91.6

88.4

88.8

90.1

91.4

1.3

1.4

1.8

3.7

1.7

1.2

1.0

0.4Plenumblock


Fuelblock


Uppershield

Evaluation Result of Fuel Temperature at Design Stage

8

Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).

The effective flow rate for the fuel cooling is about 88% of total flow.

Remains are the leakage flow through the permanent reflectors, flow in the control rod cooling channel, and the bypass flow in the inner column gap.

Evaluation Result of Fuel Temperature at Design Stage

Nominal temp.Maximum temp.

Core Center

12931475

13051474

13021476

12931475

13211492

12951476

13021476

13051474

13211492

12951476

13021476

Temperature limits of HTTR fuel Anticipated operation occurrence: 1600 ˚C Normal Operation : 1495˚C

9

400 600 800 1000 1200 1400 1600Temperature (oC)

(maximum)Compact inner surface (nominal)

1

2

3

4

5

Vertical position

at a fu

el colum

n

(Top)

(Bottom)

Sleeve outer surface

Coolant

Graphite block


Vertical temp. distribution for the fuel where the maximum fuel temp. appeared.

Horizontal temp. distribution

Re‐evaluation of HTTR Fuel Temperature

(1) 1st modification using the operation data of the rise‐to‐power test up to 850˚C Validation of power and helium flow distributions Revision of operating conditions (e.g., core inlet coolant temperature) Revision of hot spot factors

(2) 2nd modification using new analysis model Detailed mesh model

(3) 3rd modification using the 950˚C operation data in 2004 Revision of core inlet coolant temperature and core coolant flow rate

10

Power den

sity (W

/cc)

0.50

0.75

1.00

F1 F2 F3,F4 F5Column number

1.25

CalculatedMeasured

CenterF4

F5 F2

F4

F1

F3

F5

F1

F3

F2

Power distribution

T/C for core inlet coolant temp.

Center region

Outside region(6 points)

T/C for hot plenum coolant temp.

Coolant tem

perature risin

g (oC)

440

450

460

470

480

490

Center region Outside region

MeasuredCalculated

Flow distribution

Validation of Evaluation Results using HTTR data Estimated power and helium flow distributions were validated using the operation data of the rise‐to‐power test up to 850oC.

Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006). 11

The gross gamma ray from the fuel assemblies was measured by GM counter

Design Revised Reasons

Operation day 440 EFPD 160 EFPD Operating condition at 1st 950oC operation

Core inlet coolant temperature 415oC 409oC Operation data

Core coolant flow rate 10.2 kg/s 10.1 kg/s Re‐evaluation of heat and mass balance using operation data

Control rod position 2610 mm 2900 mm Operation data

Systematic factors

Factor for thermalpower

Coolant temp. rise 2.5% 0% No effect of the thermal power error on the coolant temp. rise

Others 2.5% 2.0% Calibration result of thermal power

Factor for axial power distribution 4.0% 0% Measurement results of power

distribution

Factor for flow

distribution

Coolant temp. rise 4.0% 2.0% Re‐evaluation of flow distribution using operation data Film temp. rise 3.2% 1.6%

1st Modification: Modification of Calculation ConditionsBoundary conditions concerning to the operating conditions, and hot spot factors (i.e., systematic factors) were revised using the operation data up to 850oC operation.

12Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).

Hot plenum


Top plenum 99.6

Control rod

coo

ling channe

l

Fuel coo

ling channe

l

Inne

r colum

n gap

100 100

6.2

6.2

6.2

6.2

6.3

6.0

5.2

5.0

99.20.2

0.2

90.5

90.4

90.2

89.6

90.3

91.4

92.9

1.7

1.7

1.9

2.4

2.1

2.1

0.9

0.5Plenumblock


Fuelblock


Uppershield

1st Modification: Re‐evaluation Result of Fuel Temperature

13

Hot plenum


Topplenum 98.9

Control rod

coo

ling channe

l

Fuel coo

ling channe

l

Inne

r colum

n gap

100 100

5.5

5.5

5.5

5.7

7.2

8.8

8.2

7.1

99.30.7

0.4

92.2

92.1

91.6

88.4

88.8

90.1

91.4

1.3

1.4

1.8

3.7

1.7

1.2

1.0

0.4Plenumblock


Fuelblock


Uppershield

Design stage 1st modificationFlow distribution


14


13101452

13071463

12911434

13101452

12861431

13111453

13021476

13071463

12861431

13111452

12911434


Core Center

Estimated maximum fuel temperature was decreased from 1492oC at design stage to 1463oC by the re‐evaluation using the operation data.

Fuel temperature


Core Center

12931475

13051474

13021476

12931475

13211492

12951476

13021476

13051474

13211492

12951476

13021476

Design stage 1st modification




400 600 800 1000 1200 1400 1600Temperature (oC)

1

2

3

4

5Vertical position

at a

fuel colum

n

(Top)

(Bottom)


Coolant

Graphite block

Design stage 1st modification

Temperature distribution

15

400 600 800 1000 1200 1400 1600Temperature (oC)

1

2

3

4

5

Vertical position

at a

fuel colum

n(maximum)

Compact inner surface (nominal)


(Top)

(Bottom)

Coolant

Graphite block


Each fuel block is divided into 6 triangular‐meshes for the models of nuclear design andthe fuel temperature analysis.

One fuel rod is represented by the triangular‐mesh of the fuel temperature analysis model .

The rector power calculated by the nucleardesign is allocated for each mesh of the fueltemperature analysis model using the peakingfactor.

The reactor power is multiplied by the hotspot factor to take into account theheterogeneous effect of the nuclear designmodel (i.e., power: +7%).

Allocated rector power×hot spot factor

2nd Modification: Modification of Analysis Model

Old model

16

The hot spot factor related to the heterogeneous effect of the nuclear design model is eliminated by using detailed mesh.

New model

1/6‐divided‐block model Each fuel rod model

Horizontal

Vertical

1

2

3

4

Fuel rod

4 meshes 14 meshes

1234567891011121314

Old model New model

2nd Modification: Modification of Analysis Model

17Ref: D. Tochio et al., “Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas‐Cooled Reactor ‘HTTR’”, Trans. At. Energy Soc. Japan, 5 [1], p.57‐67(2006).

FLOWNETIn‐vessel thermal and hydraulic

analysis code

Fuel, control rod, core component, core internal structure design data

Power density and neutron fluence distributions

Nuclear design code

TEMDIMFuel temperature analysis code

Coolant flow rate distribution

(1)

(2)

(4)

MVPContinuous energy Monte Carlo code

Power distribution for each fuel rod

(3)

Fuel temperature

Added

2nd Modification: Modified Evaluation Method

18

12611382

12561368

13141428

12511372

12211330

12651387

13021476

12581371

12141322

12551377

13031417

13101452

13071463

12911434

13101452

12861431

13111453

13021476

13071463

12861431

13111452

12911434

Old model (1st modification) New model (2nd modification)

Old model New model Temp. difference

Maximum fuel temp. 1463oC 1428oC ‐35oC

Core average fuel temp. 1178oC 1018oC ‐160oC


2nd Modification: Re‐evaluation Result of Fuel Temperature

19

Fuel temperature


2nd Modification: Re‐evaluation Result of Fuel Temperature

20

400 600 800 1000 1200 1400 1600Temperature (oC)

1

2

3

4

5

Vertical position

at a

fuel colum

n(maximum)

Compact inner surface (nominal)


(Top)

(Bottom)

Coolant

Graphite block

400 600 800 1000 1200 1400 1600Temperature (oC)

1

2

3

4

5

Vertical position

at a

fuel colum

n



(Top)

(Bottom)

Coolant

Graphite block

Old model (1st modification) New model (2nd modification)

Temperature distribution


3rd Modification: Re‐evaluation Result of Fuel Temperature

21

Analysis result(2nd modification)

Operation data at 950oC operation in 2004

Coolant temp. (oC)

Core inlet

Center region 399 396

Outside region 405 402

Core outlet



Temperature rising



Maximum fuel temp. (oC) 1,428Analysis result using operation

data (3rd modification)1,435


Fuel Temperature Measurement (Future Test)

22

Objectives To develop the fuel temperature measuring technique for the prismatic HTGR To upgrade the core design technology for the prismatic HTGR

Methods Temperature monitors using 22 kinds of melting wire in temperature range 600 ‐1390oC Irradiation performance of the melt‐wire temperature monitor has been confirmed by

JMTR capsule irradiation test. Temperature monitors will be inserted into the fuel assembly to measure temperature

distribution of the HTTR core.

Melt‐wire temperature monitor

Monitors will be stacked into hole under the dowel pin of the fuel assembly

Fuel assembly

HTTR

Ref: Y. Tachibana et al., “Test Plan using the HTTR for Commercialization of GTHTR300C”, JAEA‐Technology 2009‐063 (2009).

Melt‐wire temp. monitor

High Temperature Irradiation of the HTGR Fuel/ Metallic FP Plate‐out Test (Future Test)Objectives

To optimize the limitation of fuel failure under accident condition

To investigate metallic FP (Sr, Cs, etc.) plate‐out behavior by using the real HTGR facility

Methods A test fuel element loaded at the center

column of HTTR and heated step‐by‐step up to 2000oC

Using temperature monitor to measure fuel temperature

On‐line measurement of primary coolant radioactivity to estimate additional fuel failure fraction

Plate‐out probe( ) will be settled in the primary circuit to measure metallic FP plate‐out concentration by PIE

PIEs to investigate fuel failure and FP plate‐out to

auxiliarycoolingsystem

Reactor

PPWC

IHX

to SPWC

By-pass line

HGC

HGC

Forsingle-loadedoperation

Forparallel-loadedoperation

To pressurized water cooling system23

Test fuel element

Ref: Y. Tachibana et al., “Test Plan using the HTTR for Commercialization of GTHTR300C”, JAEA‐Technology 2009‐063 (2009).

Summary

24

HTTR fuel temperature was re‐evaluated using the HTTR operating data and new analysis model.

The summary of the re‐evaluation results of the maximum fuel temperature is the following: Design stage : 1492oC 1st modification using 850˚C operation data : 1463oC 2nd modification new analysis model : 1428oC 3rd modification using 950˚C operation data : 1435oC

We are planning to measure the HTTR fuel temperature using the melt‐wire temperature monitor.

Documents

Re evaluation of Maximum Fuel Temperature of the …€evaluation of Maximum Fuel Temperature of the HTTRat Normal Operation Hirofumi OHASHI Nuclear Hydrogen and Heat Application Research