HTGR Technology Course for the Nuclear Ct Rl i … Modules HTGR Fundamentals/Module... · HTGR...

HTGR Technology Course for the Nuclear R l t C i i Regulatory Commission

May 24 – 27, 2010ay , 0 0

Module 6bPebble Bed HTGR Nuclear Design

Pieter VenterPebble Bed Modular Reactor (Pty) Ltd.

1Slide 1

Outline

• Pebble bed nuclear design characteristicsPebble bed nuclear design characteristics• Nuclear design considerations• Analytical toolsy

2Slide 2

Pebble Bed Nuclear DesignCore Neutronics / Thermal Hydraulics

Analyze Neutronic Design for

Shielding and Activation

Analyze Shielding and Activation of core structures and surrounding

feedback to both engineering and safety

Steady-State (VSOP, MCNP) and Transient (TINTE) Analysis

Monte Carlo Analysis (MCNP) or simplified transport analysis (MicroShield)

Input to engineering on coreInput to engineering on core component temperatures, power profiles etc.

Input to safety on maximum fuel

Input to engineering on core component activities for maintenance / decommissioning

Input to safety for worker doseInput to safety on maximum fuel temperatures, control rod worth etc.

Fission Product Releases

D t i Fi i P d t

Dust Generation and Activation

Graphite and metallic dust generation in the core and fuel handling system and activation of the dustDetermine Fission Product

Releases for both normal operation and accident scenarios

Input to the rest of the source term analysis chain

Fuel Source Term

Neutron and photon sources from spent and used fuel

3Slide 3

analysis chain p

Key Pebble Bed Nuclear Characteristics

• Continuous fueling provides core design Continuous fueling provides core design flexibility to introduce different fuel cycles on-line

• Burnup measurement of each sphere instead of calculations reduces core design uncertaintyuncertainty

• No reload analyses are required and approach to criticality is only required for approach to criticality is only required for initial core loading or after reflector replacement is performed

4Slide 4

Fuel Cycle Flexibility

• Pebble bed can also be used for U-Th, Pu Pebble bed can also be used for U Th, Pu disposition and MOX

• AVR experience demonstrated core operation with between 4 and 14 different fuel elements

Heavy metal loadings ranging from 5g to 20g– Heavy metal loadings ranging from 5g to 20g– Enrichments ranging from 5% to 93% U-235

• Different fuel cycle can be introduced on-line • Different fuel cycle can be introduced on-line and the reactivity effect can be monitored continuously during the transition period

5Slide 5

Outline


6Slide 6

Key Considerations

• Equilibrium core parametersq p• Excess reactivity• Neutron control and shutdown• Temperature coefficients• Xenon stability

P di t ib ti• Power distribution• Neutron flux• Fuel temperatures• Fuel temperatures• Burnup• Decay heat

7Slide 7

y

Typical Equilibrium Core Parameters

Description Units PBMR-400MW

PBMR-250MW

Thermal power MW 400 250

Core diameter (inner/outer) m 2.0/3.7 1.5

Core height (average) M 11.0 10.5

Helium coolant temperatures (Inlet/Outlet) °C 500/900 250/750Helium coolant temperatures (Inlet/Outlet) C 500/900 250/750

U-235 enrichment wt% 9.6 7.8

Number of average fuel sphere cycles 6 15

Average residence time in core Days ~ 930 ~ 1080

Average discharge burn-up MWd/t 91 450 81 565

Number of Fuel Spheres (FS) ~ 452 000 ~ 400 300p ( )

Average number of fresh fuel spheres to be loaded per day ~ 486 ~ 438

Average number of fuel spheres recirculated per day ~ 2 913 ~ 6 566

8Slide 8

Load Following RequirementXenon Reactivity Effect Due To Power Changes

1.50

180

200

Cylindrical 250MW 700°C

0.50

1.00

ct o

f Xen

on

k]

120

140

160

of T

otal

]

Cylindrical, 250MW, 700 C

-0.50

0.000 20 40 60 80 100 120

eact

ivity

Effe

[%Δ

k

60

80

100

Pow

er [%

o

-1.50

-1.00

Time [h]

Re

0

20

40%ΔkPower

• Changes in reactor power causes Xe changes in core that affects the reactivity

• This lowering of reactivity following a power change needs to be compensated for by adding excess reactivity

Time [h]

9Slide 9

compensated for by adding excess reactivity

Excess Reactivity• The excess reactivity is the additional reactivity

available in the core during operating conditions by the loading of a fuel mixture that is more reactive (less the loading of a fuel mixture that is more reactive (less burned) than what is required to keep the reactor critical at the full power operational conditions (temperatures and equilibrium fission products)(temperatures and equilibrium fission products)

• The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical

• The excess reactivity can be changed by changing • The excess reactivity can be changed by changing the position of the control rods and the adjusting loading of fresh fuel into the core

10Slide 10

Neutron Control and Instrumentation Requirements

• Provide two independent and diverse Provide two independent and diverse systems of reactivity control for reactor shutdown

• Each system shall be capable of maintaining hot subcriticalityO t h ll b bl f i t i i • One system shall be capable of maintaining cold shutdown

• Provide neutron flux (low intermediate and • Provide neutron flux (low, intermediate and high range) and axial profile measurements

11Slide 11

Typical Pebble Bed Neutron Control Systems

• Control rods are used for operational (ROT) control and hot shutdown

• Small absorber spheres Small absorber spheres (SAS) are used to achieve cold shutdown

• Control rods and SAS

Control Rod

• Control rods and SAS are located in reflectors (SR/SR or SR/CR)

SAS Channel

• Control rods can be operated in banks SAS

Extraction Point

12Slide 12

Reactor Shutdown with Control Rods

Annular 400MW 900°CAnnular, 400MW, 900 C

• Rapid shutdown of reactor for two different control rod speeds• 1 cm/s is the normal controlled insertion

13Slide 13

1 cm/s is the normal controlled insertion• 50 cm/s is the scram speed when the power is cut to the CRDMs

Shutdown by Interruption of Coolant Flow

Annular, 400MW, 900°C

• When the coolant flow is stopped through the core, the sphere temperatures increase and the reactor is shutdown, even with no movement of the control rods

i ffi i ff f i

14Slide 14

• This temperature coefficient effect has been successfully demonstrated in the AVR and HTR-10

Temperature Coefficients

Temperature Coefficients Unit At Operating Conditions


Fuel (Doppler coefficient of mainly 238U) Δρ/°C - 4.4 x10-5

Moderator Δρ/°C - 1.0 x10-5

Reflector regions (all together) Δρ/°C + 1.8 x10-5

• The Doppler temperature coefficient acts promptly and stabilizes the nuclear chain ti

g ( g ) ρ

TOTAL Δρ/°C - 3.6 x10-5

reaction.

• The moderator coefficient acts promptly when temperature change of the moderator is the primary event, and causes a response in the neutron flux and fission rate.

• The reflector temperature changes are not so strongly coupled to power changes in the fuel. The side reflector temperature is dominated by the coolant temperature in the riser channel and combined with the large heat capacity will cause a considerable delay of effect of the reflector temperature coefficient.

15Slide 15

Xenon Stability

• Xenon stability refer to the degree to which the spatial flux distribution varies for a specific reactor design due to distribution varies for a specific reactor design due to spatial xenon dynamics

• The main question is whether the change in xenon q gconcentration with time exhibits a damped or un-damped oscillatory behavior

• Previous studies on HTGR specific xenon stability reported • Previous studies on HTGR-specific xenon stability reported the following conclusions:– Un-damped axial xenon oscillations only occurred for HTR

cores when the core height was increased to larger than 8 m, with a simultaneous power density increase to more than 20 with a simultaneous power density increase to more than 20 MW/m3

– No un-damped radial xenon oscillations were observed for cylindrical cores of up to 6.4 m in radius

16Slide 16

Xenon Oscillation Results from100%-40%-100% Power Variation


• Legend refers to different heights in the core with 100 being near the top of the bed and 1000 being at the bottom of the bed

17Slide 17

the bed

Power Shaping

• Pebble beds do not have to use fuel loading Pebble beds do not have to use fuel loading or burnable poison to affect power shaping

• Fuel can be circulated faster through the core to flatten the axial power profile

• Once equilibrium conditions are established t l d d t d t b d t control rods do not need to be moved to

compensate for burnup

18Slide 18

Axial Power Distribution• The power profile is biased

towards the top of the core due to the lower fuel temperatures and fresh fuel

Axial Power Distribution

0Cylindrical, 250MW, 700°C

temperatures and fresh fuel• The top of the power profile is

depressed due to the partially inserted control rods

• Since no burnable poisons

200

400[cm

]

Since no burnable poisons are used with on-line refueling, this power profile is representative of the greater part of plant operation

400

600

ght (

From

Top

) [

Note: channels are not physical but represent sphere flow regions for modeling with

800

1000

Axi

al H

eig

Channel 1Channel 2Channel 3 g g

channel 1 towards the centre of the core and channel 5 towards the outside of the core

1000

12000 0.5 1 1.5 2

Channel 4Channel 5

19Slide 19

Relative Power Density

Power Distribution per Pass

• The power produced

Power Per Pass for Channel 1 (Center of Core)

0Cylindrical, 250MW, 700°C p p

per fuel sphere reduces with each pass through the core

0

200

cm]

400

600

t (fr

om T

op) [

c

800

1000

Axia

l Hei

ght

Pass 1Pass 5

1000

12000 0.5 1 1.5 2

Pass 10Pass 15

20Slide 20

Power Per Pebble [kW]

Neutron Flux Distribution (Radial)Radial Flux Profile

1.6E+14CORE SIDE REFLECTOR

Cylindrical, 250MW, 700°C

1.2E+14

1.4E+14

] > 0 1 MeV

CORE SIDE REFLECTOR

8E+13

1E+14

x [n

cm

-2 s

-1 > 0.1 MeV0.1 MeV - 0.009 MeV0.009 MeV - 130 eV130 eV - 1.86 eV< 1 86 eV

2E 13

4E+13

6E+13

Flux

< 1.86 eV

0

2E+13

0 50 100 150 200 250 300

R di l Di t (f t ) [ ]

21Slide 21

Radial Distance (from center) [cm]

Neutron Flux Distribution (Axial)Axial Flux Profile (Center of Core)

1.6E+14 CORETOP REFLECTOR

BOTTOM REFLECTOR


1.2E+14

1.4E+14

> 0.1 MeV

REFLECTOR REFLECTOR

8E+13

1E+14

x [n

cm

-2 s

-1]

0.1 MeV0.1 MeV - 0.009 MeV0.009 MeV - 130 eV130 eV - 1.86 eV< 1.86 eV

4E+13

6E+13Flux

0

2E+13

0 200 400 600 800 1000 1200 1400 1600 1800

22Slide 22

Axial Distance (from top) [cm]

Temperature Distribution for Equilibrium Core

Axial Fuel Temperature Distribution

0Cylindrical, 250MW, 700°C

0

200

[cm

]

400

600

ght (

from

top)

800

1000

Axi

al H

eig Core Center

Between Core Center and Side Reflector

Near Side Reflector

1200200 300 400 500 600 700 800 900 1000

Average Fuel Temperature [°C]

23Slide 23

Fuel Temperature DistributionTemperature Distribution

18Average Pebble Surface Temperature


14

16 Average Fuel (UO2)Temperature(Doppler)

8

10

12

age

of S

pher

es

4

6

8

Perc

enta

0

2

24Slide 24

295 341 387 433 479 525 571 617 663 709 755 801 847 893 939

Temperature [°C]

Burnup• Fuel spheres are circulated through the core continuously and

each sphere is measured for burnup as it comes out of the core

• Fuel spheres that have not yet reached target burn-up, or more specifically the burnup limit setpoint, can be reloaded and recycled continuously during normal reactor operation.

• Fuel spheres with higher burn-up as the setpoint are discharged from the refueling line, and replaced with a fresh fuel sphere. In practice some fuel may pass through the reactor less than the practice some fuel may pass through the reactor less than the average number of passes while others may pass more times before reaching the setpoint burnup value.

The Burnup Measurement System discriminates between spent • The Burnup Measurement System discriminates between spent and used fuel spheres by analyzing the gamma energy spectrum to determine the inventory of specific nuclides (specifically Cs-137).

25Slide 25

Decay Heat• German standard DIN is used to evaluate decay heat• The standard provide the methodology to calculate the heat power

generated by the decay of the fission products (valid for all kinds of thermal reactors) and rules concerning the additional sources of ) gdecay heat (activation) in the fuel of pebble-bed high temperature reactors

7


4

5

6

of P

ower

]

2

3

4

Dec

ay H

eat [

% o

0

1

0 00001 0 0001 0 001 0 01 0 1 1 10 100 1000 10000 100000 1000000

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0.00001 0.0001 0.001 0.01 0.1 1 10 100 1000 10000 100000 1000000Time [h]

Outline


27Slide 27

PBMR Nuclear Codes• Analysis Codes

– VSOP99/5: HTR core neutronics code including cell calculations, 2D and 3D reactor physics simulation, depletion and 2D quasi-static thermal-hydraulichydraulic

– T-REX: 1-D transport solution used to model control rods for VSOP99/5– TINTE: HTR transient analysis tool based on 2-D spatial kinetics– GETTER: Metallic (long lived) fission product release calculation software

for normal and accident conditions.– NOBLEG: Steady state gaseous (short lived) fission product release

calculation software– SCALE5: Used for the calculation of the fuel source term and fuel

depletionDAMD: Calculation of radioactive source terms on the surfaces and in the – DAMD: Calculation of radioactive source terms on the surfaces and in the coolant due to the plate-out of condensable atomic fission products released from the fuel in the core. The code also calculates the deposition of dust and the amount of dust circulating in the coolant.

– MCNP5: Monte Carlo code for neutron, photon and electron transport d th l l ti f iti litand the calculation of criticality

– FISPACT: Code for the calculation of neutron induced activation source terms

28Slide 28

Summary• The nuclear design of a pebble bed reactor is

simple and straightforwardsimple and straightforward• The on-line refueling provides significant

flexibility to the core designer to choose a fuel y gcycle whilst reducing uncertainty in the core calculations

i• The core designer has well proven and validated nuclear design codes at his disposal to calculate the core behavior and to calculate the core behavior and characteristics

29Slide 29

Suggested Reading• PBMR Nuclear Design and Safety Analysis: An Overview, Stoker, C.C.,

PHYSOR 2006, Vancouver, BC, Canada, September 10-14. • Plutonium Disposition in the PBMR-400 HTGR, E. Mulder, PHYSOR 2004,

A il 2004April 2004• The Pebble Bed Modular Reactor Layout and Neutronics Design of the

Equilibrium Cycle, Reitsma, F., Proceedings of PHYSOR2004 Meeting, Chicago, USA, April 25-29, 2004.

• An Overview of the FZJ Tools for HTR Core Design and Reactor • An Overview of the FZJ-Tools for HTR Core Design and Reactor Dynamics, the Past, Present and Future, Reitsma, F., Rütten, H. J. and Scherer, W., Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications, Palais des Papes, Avignon, France, September 12-15, 2006

• Comparison of VSOP and MCNP Results of PBMR Equilibrium Core Model, Sen, S., Albornoz, F., and Reitsma, F., Proceedings of the 3rd International Topical Meeting on High Temperature Reactor Technology, Johannesburg, Gauteng, South Africa, 2006

• The Re evaluation of the AVR Melt Wire Experiment Using Modern • The Re-evaluation of the AVR Melt-Wire Experiment Using Modern Methods with Specific Focus on Bounding the Bypass Flow Effects, Viljoen, C. F., et al Proceedings of the 4th International Topical Meeting on High Temperature Reactor Technology, HTR-2008, Washington DC, USA, 2008

30Slide 30

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