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Thermal Fluid Characteristics for Prismatic HTGRs

IAEA Course on High Temperature

Gas Cooled Reactor Technology

Tsinghua University, Beijing

October 22-26, 2012

Javier Ortensi

Idaho National Laboratory (INL)

Overview

• General Thermal Fluids Characteristics

• Design Requirements

• Bypass Flow and Cross Flow

• Thermal Fluids Modeling

• Some of the Remaining Challenges – Lower plenum

– RCCS & more

• Conclusions

General: Helium Flow Path

General: Passive Safety Design

• Annular core design – Lower inner reflector and fuel temperatures

– More graphite (heat storage)

• Large negative temperature coefficient – Temperature rise always leads to power decrease

• Passive decay heat removal – RCCS with natural convective flow

– RCCS without natural convective flow to ground

• No powered reactor safety systems

General: Materials

• Graphite core – Large heat capacity (Cp)

– Large thermal inertia (I)

– Very isotropic, low impurities

– Stability – sublimates above 3600oC

• Multi-hole block design optimized to: • reduce fuel temperature

• reduce pressure drop

• maintain graphite structural integrity

• Thermo-physical properties depend on temperature and fluence (burnup)

General: Local Heating

• Primarily from large fission fragments in fuel

• Conduction in solid material (fuel & graphite)

• Conduction + radiation in gaps

• Convection in fluid regions

fuel

graphite

coolant

gap

General: Non-local heating

• Approximately 6% of the power is generated via gamma + neutron heating – Heat neighboring blocks

– Use specialized cross sections

– Require a gamma transport calculation or correction

• ~2% of this is deposited in the inner and outer reflector which may have non-negligible effect on the temperature distribution

• Affects time of criticality during LOFC without SCRAM

General: Flow

• Down through core – Forces transition to natural convection during loss of flow

• Well defined flow channels

• Mainly low mach number with velocities ~60 m/s max.

• Turbulent flow during normal operation

• Flow redistribution driven by high flow resistance in hot regions – Large core with large DTs

– Flow rate variability between columns and outlet temperatures

• Small temperature rise from parasitic radial flow

• Reduced pressure drop to avoid flow induced vibrations

Overview

• General Thermal Fluids Characteristics

• Design Requirements

• Bypass Flow and Cross Flow

• Thermal Fluids Modeling

• Some of the Remaining Challenges – Lower plenum

– RCCS & more

• Conclusions

PMR Thermal Fluids Design Requirements (GA)

Temperature profile of the MHTGR (GA)

At steady state core mid height – high power

location

DT~100 from fuel to moderator

DT~125 moderator to coolant

DT~300 across a block near inner reflector

Overview

• General PMR Thermal Fluids

• Design Requirements

• Bypass Flow and Cross Flow

• Thermal Fluids Modeling

• Some of the Remaining Challenges – Lower plenum

– RCCS & more

• Conclusions

PMR Bypass Flow (1/2)

• Prediction of the bypass flow is the most important thermal fluids phenomena

• Flow that bypasses coolant holes

• Gaps between fuel blocks – Large temperature gradients and fast neutron irradiation result in uneven

block expansion (uncertainty in gap geometry) – Needed for refueling – Provide cooling to peripheral compacts

• Needed to cool control rods

13

PMR Bypass Flow (2/2)

0 mm gap

5 mm gap

Crossflow

• Horizontal gaps between stacked graphite elements form leakage paths to/from primary coolant flow path

• Driven by lateral pressure gradient

• Leakage crossflow characteristics in an HTGR core have been studied experimentally and numerically by H. Kaburaki and T. Takizuka – Empirical cross flow equations developed

15

Overview

• General PMR Thermal Fluids

• Design Requirements

• Bypass Flow and Cross Flow

• Thermal Fluids Modeling

• Some of the Remaining Challenges – Lower plenum

– RCCS & more

• Conclusions

17

PMR Thermal Fluids Modeling Requirements

• Incompressible flow models OK for regions with no heat transfer

– Can assume that pressure is constant

• Compressible flow models for low Mach number

– Heat transfer regions

– Max He velocity (~60 m/s) (acoustic velocity 1500 m/s)

– Might not be the case for some transients

• Momentum flux terms should be included in conservation equations

– Large DT across core causes significant localized fluid acceleration

– Significant for depressurization event (blowdown)

• Frictionless flow

– Fluid press change from plenum to duct can be addressed with good duct design - produces flow acceleration with minor losses

18

PMR Thermal Fluids Modeling Requirements

• Capturing the bypass flow is essential

– 10%~25% of total coolant flow moves between the blocks (ANL-GenIV-071)

• Ability to capture flow reversal

– Forced circulation flow is from top of core to bottom of core

– Natural circulation flow (LOFC) is from bottom to top

– Need those gravitational terms for transients

• Complex low velocity, buoyancy-driven flow in upper plenum during LOFC: limiting safety case for core barrel top plate

Approach to Core Thermal Fluids Modeling

• Traditional TF – 1-D pipe flow networks

– Subchannel codes

– Fast for core transient analysis but details are lost

• Homogeneous multi-physics methods are also in use – Solve tightly coupled neutronics and thermal fluids equations

– Better representation of the core

– Require properly homogenized parameters from CFD or other correlations

• Computational Fluid Dynamics – Used in small domains where details needed or other methods

fail

– Not used for full core transient analysis

Approach to Core Thermal Fluids Modeling (1D System Codes e.g. RELAP5)

• Reactor is represented by a series of 1-D pipes

• Pros

– Flexible

– History of successful use in LWRs

• Cons

– Ability to model 3-D bypass flow phenomena is questionable

– Existing codes would require substantial modification for use with HTRs

• Modeling concepts have been borrowed from these codes (i.e. single and time dependent junctions/volumes) for use in AGREE code

20

RELAP5 model

Example: AGREE PMR Fluids Modeling

• A 3-D core is represented by a series of cross-connected 1-D subchannels

• Subchannel method is based on proven LWR core thermal-hydraulic analysis methodologies (i.e. COBRA/VIPRE)

1-D subchannels flow into and out of slide

Approach to Core Thermal Fluids Modeling: Multi-Physics (3-D homogenized)

• Partial 3-D representation of homogenized core with many (103-104) cells

• Pros – Improved fidelity solution

– Tightly coupled allows improved

time stepping with less error

– Easy to add multi-scale

– Computationally reasonable

• Cons – Need multi-mesh capability

– Complex systems to maintain

– Parallelization is required

Approach to Core Thermal Fluids Modeling: (Computational Fluid Dynamics)

• Full 3-D representation of core using large number of cells (>106)

• Pros – High fidelity solution

• Cons – Computationally expensive

– Parametric studies may be cumbersome/slow

– So far loosely coupled to neutronics in SHARP with DeCART to STAR-CD

• Use CFD as a verification tool

23

(Image: ANL-GenIV-121)

Effective Thermal Conductivity

• NRC contracted AMEC to determine thermal conductivity

• Computed for – TRISO

– Compact, pebble

– Block

• Results agree well with reference CFD computations

Temperature

[K]

Ref. k

[W/m.K]

AMEC k

[W/m.K]

500 29.401 29.349

1000 19.557 19.518

1500 15.245 15.214

0

5

10

15

20

25

30

35

-0.4

-0.2

0

0.2

0.4

0.6

0.8

1

400 900 1400

Th

erm

al C

on

du

ctiv

ity

[W

/m.K

]

Geo

met

ric

Co

rrec

tio

n F

act

or

Temperature [K]

UO2

buffer

PyC1

PyC2

SiC

matrix

k

Courtesy of Ivor Clifford, PSU, 2012

Overview

• Helium Flow Path

• General PMR Thermal Fluids

• Design Requirements

• Bypass Flow and Cross Flow

• Thermal Fluids Modeling

• Some of the Remaining Challenges – Lower plenum

– RCCS & more

• Conclusions

Analysis Challenges: Lower Plenum (1/2)

• Graphite core support structure – Forms exit plenum

– Provides inspection/surveillance capability

Analysis Challenges: Lower Plenum (2/2)

• Multiple jets enter lower plenum (hot streaks ± 200 oC of average)

• Cross flow moves towards outlet

• Most full core TF models cannot resolve this

• CFD can, but slow given the domain size

• New CFD-like 3-D flow models needed for these regions

Courtesy of Fluent

Analysis Challenges: RCCS

• Used to remove heat from RPV to containment structure

• Safety grade passive or redundant forced convection

• Air versus water cooled designs (doubt about reliability of air cooled)

• Needs to accommodate all scenarios

• Two-phase flow behavior RCCS water

cooling channels – natural circulation with flow stagnation

– forward and backward flow

Other - Thermal Fluid Challenges

• Effect of the gamma + neutron heating on the thermal fluids calculation

• Modeling changes in the by-pass and cross flow channels (gaps) due to graphite irradiation and thermal expansion/contraction

• Changes in the thermo-physical properties due to temperatures and irradiation

• Low flow phenomena (natural circulation plumes on top plate)

• Graphite oxidation highly dependent on TF solutions of ingress

Conclusions

• Design requirements are well established

• Single phase fluid flow is straight forward to model in most regions of the core with various techniques

• All heat transfer mechanisms: conduction, convection, and radiation

• Modeling of bypass is primary challenge

• Some difficulties with modeling of plenums and transition to natural convection

• Interplay of power peaking, bypass & crossflow to determine maximum fuel temperatures

• Other challenges in hot streaking, RCCS design

Bibliography

– “HTGR Technology Course for the Nuclear Regulatory Commission” May 24-27, 2010, Idaho National Laboratory and General Atomics.

– “Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR,” H. Sato, R. Johnson, R. Schultz, Annals. Nuc. Ener., 37 (2010) pp. 1172-1185.

– “Crossflow Characteristics of HTGR Fuel Blocks,” H. Kaburaki, T. Takizuka, Nucl. Eng. Des., 120 (1990) pp 425-434.

– “Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor,” N. Tak, M. Kim, W.J. Lee, Annals Nuc. Ener., 35 (2008) pp. 1892-1899.

– “RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor”, C. Oh, C. Davis, G.C. Park, NURETH-12, Pittburg, Pennsylvanya, USA, 2007.

javier.ortensi@inl.gov (208) 526-4256

Presented by:

Javier Ortensi (INL)

Content:

Javier Ortensi (INL)

Gerhard Strydom (INL)

Volkan Seker (U-Michigan)

RELAP5-3D Results MHTGR