The Mutual Influence of Materials and Thermal … of Nuclear Technology and Energy Systems The...
Transcript of The Mutual Influence of Materials and Thermal … of Nuclear Technology and Energy Systems The...
Institute of Nuclear Technology
and Energy Systems
The Mutual Influence
of Materials and
Thermal-hydraulics
on Design of SCWR –
Review of Results of
the Project HPLWR
Phase 2
J. Starflinger,
T. Schulenberg
• Design Target Data:
• Operational pressure: 25 MPa
• Core mass flow: 1160 kg/s
• Power output: 1000 MWe
• Constraints:
• Average core exit temp.: 500°C
• Max. cladding surface temp.: 625°C
• Max. linear heat rate: 39 kW/m
10/10/2016
University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 2
5th Framework Programme of the EU
HPLWR – High Performance Light Water Reactor
AREVA NP, 2005
• „Hot Channel“ by definition is the channel, in which all uncertainties, non-
homogeneities and allowances sum up, leading to the highest enthalpy
rise of the entire core under normal operation conditions!
• Very conservative, provides a very high safety margin
10/10/2016
University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 3
Definition
„Hot Channel“ Form Factor Analysis of the Core
av
av
h
h
h
hF
max
max
Maximum enthalpy rise in the „Hot Channel“
Average enthalpy rise in the core
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 5
Design Targets of Hot Channel Factors
2.01.6Total
Power control, flow control, pressure control, inlet
temperature control
1.15Allowances
Material properties of coolant and claddings,
physical modelling, hydraulic modelling, heat
transfer coefficient, geometry tolerances
1.2Uncertainties
1.6Axial power factor
1.15Local peaking
factor inside FA
1.25Radial peaking
factor
Fuel enrichment and distribution, water density
distribution, reflector design and properties, fuel
and control rod pattern, burn-up, burnable
poisons, …
Form factors for
power profiles
Key ParametersradialaxialHot Channel Factor
2.01.6Total
Power control, flow control, pressure control, inlet
temperature control
1.15Allowances
Material properties of coolant and claddings,
physical modelling, hydraulic modelling, heat
transfer coefficient, geometry tolerances
1.2Uncertainties
1.6Axial power factor
1.15Local peaking
factor inside FA
1.25Radial peaking
factor
Fuel enrichment and distribution, water density
distribution, reflector design and properties, fuel
and control rod pattern, burn-up, burnable
poisons, …
Form factors for
power profiles
Key ParametersradialaxialHot Channel Factor
Schulenberg, KIT, 2010
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 6
One-Pass Core
„Hot Channel“ Form Factor Analysis of the Core
• Designed for
500°C core
outlet
temperature
• Coolant
average
conditions
Heinecke, AREVA, 2010
Average
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IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 7
One-Pass Core
„Hot Channel“ Form Factor Analysis of the Core
• Hot fuel
assembly
(∙ 1.25)
Heinecke, AREVA, 2010
Average
+ Assembly Power
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 8
One-Pass Core
„Hot Channel“ Form Factor Analysis of the Core
• Hot rod
(∙ 1.25
∙ 1.15
= 1.44 )
Heinecke, AREVA, 2010
Average
+ Assembly Power
+ Rod Power
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 9
One-Pass Core
„Hot Channel“ Form Factor Analysis of the Core
• Hot rod +
uncertainty
(∙ 1.25
∙ 1.15
∙ 1.2
= 1.73 )
Average
+ Assembly Power
+ Rod Power
+ Uncertainty
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10
One-Pass Core
„Hot Channel“ Form Factor Analysis of the Core
Heinecke, AREVA, 2010
• Hot rod + uncertainty + operation (∙ 1.25 ∙ 1.15 ∙ 1.2 ∙ 1.15 = 1.98 )
• Coolant temperature ≈ 1200°C
• Cladding surface temperature > 1200°C
Average
+ Assembly Power
+ Rod Power
+ Uncertainty
+ Operation
• Simple „Hot-Channel“ analysis revealed the unfeasibility of single-pass
core design. No material available.
• Idea from T. Schulenberg, KIT:
Propose a “Three-pass core” with intermediate mixing in special mixing
chambers.
• One key-issue of a feasible core design is mixing!
• not to overheat the cladding surface temperature
• avoid hot streaks from one assembly to another and hot-spots on the
cladding surface
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 11
Consequences
„Hot Channel“ Form Factor Analysis of the Core
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
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Three Pass Core Design Proposal for a HPLWR
1000
1500
2000
2500
3000
3500
4000
Evaporator Superheater 1 Superheater 2
En
tha
lpy
[k
J/k
g]
hot channel
average
Strategy to overcome hot-channel issue: Heat-up in steps with Intermediate mixing of the coolant
Mixing
Mixing
Schulenberg, 2006
4 : 2 : 1
Power ratio of the core zones
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Three Pass Core Design Proposal for a HPLWR
200
250
300
350
400
450
500
550
600
650
Evaporator Superheater 1 Superheater 2T
em
pe
ratu
res
[°C
]
cladding
hot channel
average
Schulenberg, 2006
• A 3-Pass coolant flow in the core allows 500°C average core exit
temperature with 625°C cladding temperature
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Heat transfer, Thermal-hydraulics, System Design for SCWR 14
Core Arrangement
Evaporator:
52 Clusters
Upward Flow
Superheater 1:
52 Clusters
Downward Flow
Superheater 2:
52 Clusters,
Upward Flow
Köhly, 2010
Downcomer flow
(50%)
Moderator flow
(50%)
Inlet flow:
280°C
25 MPa
1179 kg/s
Core flow
(100%)
Upper dome
Downcomer
Area
University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Heat transfer, Thermal-hydraulics, System Design for SCWR
HPLWR Flow Path
3/11/2016 15
Köhly, 2010
• Analyze the core, whether the power peaking factors will be met
• Neutronics:
• Simulate neutronics (BOC, EOC) for core and assembly-wise power
distribution (input from materials and TH needed)
• Thermal-hydraulics:
• Suitable heat transfer correlation with an uncertainty of less than 25%,
especially for fuel rod bundles with wire wraps as spacers.
• Materials & Water Chemistry
• Identify suitable materials for thick wall and thin wall components,
especially for cladding.
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 16
Tasks for the HPLWR Partners
Design Support
• Partners:
VTT, Finland, NRI Czech Republic, CEA, France, EKI, Hungary, JRC-IE,
Petten (supporting)
• Test of 16 materials in autoclaves at different temperatures
• Investigation of general corrosion, stress-corrosion cracking and creep
• Special focus on thin wall materials
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 17
Materials
• 40 fuel pins with 8mm diameter
• cladding thickness 0.5 mm
• wire wraps as grid spacers
• wire diameter 1.34 mm
• assembly box with 3 mm
thickness incl. thermal insulation
• moderator box with 2 mm
thickness incl. thermal insulation
• Insulation material: ZrO2
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 18
Details of the Assembly Design Concept
Moderator
box
Assembly
box
Wire
wrap
spacers
Himmel, Köhly 2008
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 19
Local Temperature Distribution
Fuel Rods with Wire Wrap Spacers
Temperature [°C]
670
590
510
470
390
310
bare rod with wire
Temperatures > 670°C
• Bending (stresses, torque)
• Hot streaks (local corrosion) Lycklama, 2009
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 20
General Corrosion Results
Materials
400°C
500°C
650°C
650°C, HWC
Sample Holder,
VTT
Heikinheimo, 2009
• Alloy 316L tube samples after 1000 h exposure under SCW conditions at
650oC:
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 21
Strong impact of cold work on corrosion
Effect of Surface Treatment
“as received”
surface finish
with #1200
emery paper
surface finish
with #600
emery paper
Machined with blunt
edge hard metal
cutting tool Heikinheimo, 2009
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 22
Influence of high Cr content on oxidation
Materials
after 600h at 650°C
0,1
1
10
100
1000
0 5 10 15 20 25
Cr(%)
Ox
ide
Th
ick
ne
ss
(m
m)
P91
P92
ODS (FZK)
ODS (EU)
PM2000
316NG
1.4970
BGA4
800H
IN 625
Data from VTT, JRC,
UJV Rez
• Oxide thickness on AISI 316NG vs. exposure time after exposure to
supercritical water at 650°C.
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
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Extrapolation of oxide thickness
Materials
50% cladding
thickness
10% cladding
thickness
Toivonen, Pentillä, 2009
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 24
Initial temperature distribution
Deformation Analysis of the Box
Inlet:
286°C Outlet:
518°C
3 mm stainless
steel plate filled
with ZrO2
Reis, 2008
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 25
Deformation Analysis of the Box
Max. deflection 4 mm
(high strain)
High stress
@ high temperatures
Reis, 2008
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IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 26
at 500°C
Stress-strain curves
High yield
strength
(good for
designers)
Toivonen, 2010
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
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at 650°C
Stress-strain curves
Reduced
yield
strength
(concern for
designers)
Toivonen, 2010
Stress-Corrosion cracking
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 28
Alloy Maximum
stress (MPa)
Strain to
failure (%)
TGSCC
(y/n)
IGSCC
(y/n)
Side cracks at the
gauge surface (y/n)
347H
@ 500°C
465 45 No
No Yes, morphology not
identified
347H
@ 650°C
NA NA NA NA NA
316NG
@ 500°C
Interrupted at
330
Interrupted
at 33
NA NA Yes, IG and TG
316NG
@ 650 °C
195 38 Yes Yes No
1.4970
@ 500°C
675 26 No No Yes, morphology not
identified
Toivonen, 2010
Stress-Corrosion cracking (cont‘d)
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
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Alloy Maximum
stress (MPa)
Strain to
failure (%)
TGSCC
(y/n)
IGSCC
(y/n)
Side cracks at the
gauge surface (y/n)
1.4970
@ 650°C
360 28 Badly
oxidized
Badly
oxidized
Yes, morphology
not identified
BGA4
@ 500°C
425 41 Yes Yes Yes, IG
BGA4
@ 650°C
NA NA NA NA NA
PM2000
@ 500°C
325 Interrupted
at 50
No
No No
PM2000 #8
@ 650°C
100 40 No No No
Toivonen, 2010
• Design rules for high temperature applications
• Creep strength
• Rupture
• Tensile yield strength
@ 1% strain
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 30
From the designers point of view…
Creep
1.4970 (650°C, 250MPa)
SCW: 25 MPa, 100 ppb
O2, deionized water
k<0,1µS/cm (inlet), water
flow rate 2-3 ml/min
Gas: Helium Toivonen, 2010
• SCW environment increases strain rate compared to He environment for
316NG and 347H (short duration tests, usually > 1000 hrs)
• Strong impact on
design expected!
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 31
SCW vs. He-atmosphere
Creep
• SCW environment increases strain rate compared to He environment for
316NG and 347H (short duration tests, usually > 1000 hrs)
• Strong impact on
design expected!
10/10/2016
University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 32
SCW vs. He-atmosphere
Creep
The reasons for the increased primary
strain rate:
“Give me more money and I will find out why ”
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 33
Thick walled component
Example from Stress Analysis
• Design pressure: 28.75 MPa
• Design temperature: 280 to 500 °C
• Material: 1.4970
• Wall thickness
• Core base plate: 0.300 m
• Bottom plate: 0.050 m
• Separation wall: 0.020 m
• Outer wall: 0.025 m
• Weight: 24.9 t
• 3-D 10-Node Tetrahedral structural + thermal solids:
• 215131 nodes
• 118591 elements
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 34
Thick walled component
Example from Stress Analysis
• Max.: 0.7 mm
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 35
Results: Deformation (cold state)
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
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Results: Mises stress distribution (cold state)
• Max.: 54 MPa
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 37
Results: Temperature distribution (operational/steady state)
433°C
309°C
• Max.: 12.5 mm
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 38
Results: Deformation (operational/steady state)
• Max.: 603 MPa
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University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 39
Results: Mises stress distribution (operational/steady state)
• Max.: 603 MPa
10/10/2016
University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 40
Results: Mises stress distribution (operational/steady state)
Allowed stress peaks: s < 2 x Rp0,2
1.4970: YS (600°C) = 350 N/mm² => 700 N/mm²
Calculated values: 603 N/mm² => below 700 N/mm² => ok!
Materials, thermal hydraulics have a strong mutual interaction on design of
SCWR.
Thick walled components operating at max. 500°C
• No major structural problems with respect to corrosion (fossil plant
technology).
Thin walled components operating at max. 500°C:
• Corrosion problems to be avoided (high Cr steels ?)
at above 600°C:
• High corrosion rate with licensed low Ni alloys (especially fuel cladding)
• High impact on core design! Redesign necessary if no suitable
materials will be found.
10/10/2016
University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 41
Main findings
Summary
• Closer collaboration between the scientific fields should be established:
• Theory of McElligott and Laurien: Surface roughness plays a role in heat
transfer (disturbing boundary layers)
• Wire wrap helpful to homogenize the surface temperature. However, hot
streaks are visible (also close to the wire) -> Local oxidation?
• Neutronic aspects must be taken into account
• Water chemistry may decide, whether to build a BWR or a PWR-type
SCWR
• In-pile experiments are the next step for SCWR materials selection!
• (for heat transfer: bundle experiments to be performed)
10/10/2016
University of Stuttgart – Institute of Nuclear Technology and Energy Systems (IKE)
IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 42
Some personal thoughts
Conclusions
Thank you!
phone +49 (0) 711 685-
fax +49 (0) 711 685-
Universität Stuttgart
Pfaffenwaldring 31 • 70569 Stuttgart • Germany
Prof. Dr.-Ing. Jörg Starflinger
62116
62008
Institute of Nuclear Technology and Energy Systems
Institute of Nuclear Technology
and Energy Systems