Status of US ITER Neutronics Activities
Outline
 Examples of US activities during EDA
 US ITER neutronics activities in the past year
 Possible future US contribution to ITER
neutronics activities
US Neutronics Computation Capabilities
 The US has been developing state-of-the-art computational tools
and data bases for nuclear analyses
 Most recent versions of codes and data will be used in ITER
nuclear analysis:
Transport codes: MCNP5 (Monte Carlo), DANTSYS3.0
(ONEDANT, TWODANT, THREEDANT), DOORS3.2
(ANISN, DORT, TORT)
Activation Codes: ALARA, DKR-Pulsar, REAC
Data Processing Codes: NJOY99.0, TNANSX2.15, AMPX-77
Nuclear Data: ENDF/B-VI, FENDL-2
Sensitivity/Uncertainty analyses: FORSS, UNCER
U.S. has been Active Participant in ITER
Nuclear Analysis During CDA and EDA
Examples of US Contribution:
• Nuclear assessment of breeding blanket options and
magnet shield optimization during CDA
• Contributed to development of complete integrated
MCNP ITER (EDA) model that includes details of
shielding blanket modules, divertor cassettes, VV
with ports, TF coils, PF coils, CS coils
• Modified MCNP allowing it to sample from actual
pointwise neutron source distribution in ITER
plasma
• Calculated poloidal neutron wall loading distribution
at FW and divertor cassette plasma facing surface
3-D Nuclear Analysis for Divertor Cassette
 Determined nuclear heating (W/cm3) profiles in divertor cassette
0.01 0.1 0.5 1
1 0.5 0.1 0.01
3
7
5
3
5
10
1
7
0.5
0.1
3
0.001
1
0.5
5
0.01
1
3
0.001
 Calculated radiation damage in cassette components
 Evaluated adequacy of VV and TF coil shielding by calculating hot spot
damage, gas production and insulator dose
 Performed 3-D divertor cassette pulsed activation calculations to determine
radioactive inventory, decay heat, and radwaste level
 Assessed streaming effects
Neutronics Assessment of Divertor Diagnostics Cassettes
 Nuclear parameters in
waveguides
 Streaming through
viewing slots
 Nuclear parameters in
mirror assemblies
2-D R-q Modeling of Two Test Modules Placed in Test Port
Distance from plasma center
Earlier Work -1996
Two Test modules of the
EU Li4SiO4 helium –
cooled design are
placed in the test model
Cos q
Edge-on Arrangement
A lattice consists of FS
layer (0.8 cm), Be bed
layer (4.5 cm) and SB
bed layer (1.1 cm)
Buffer zone between
modules (10 cm)
Upper module has 75%
Li-6 (19 lattices), lower
module 25% Li-6 (16
lattices)
Lower attenuation
(higher flux) behind
beryllium layers
Dose Rate (mS/h) in ITER Building During
Operation and After Shut Down
2-D Model of ITER Machine
Coupled 2-D and 3-D Calculations
using Deterministic method (DORT
and TORT codes) for neutron and
gamma flux calculations and DKRPulsar code for dose calculations
One Week after Shutdown
2-D Model of ITER Building
Tritium
Building
3-D
modeling
Verification of ITER Shielding Capability with Various Codes and Nuclear
Data
Participants
in this Task:
US/JAERI:
for thick
shield
experiments
EU/RF: for
thin shield
experiments
R-Z model of SS316/Water
assembly
R-Z Model of the simulated Super
conducting magnet in
SS316/Water Assembly
SS=Shelf-shielded data- SS316/water
assembly
Data
Verified:
Various
reaction
rates, neutron
and gamma
spectra,
heating rates
US Nuclear Support for ITER Restarted
Following US Rejoining ITER in 2003
Major effort has been in support of ITER TBM program
A study was initiated to select two blanket options for US ITER-TBM in
light of new R&D results
Initial conclusion of US community is to select two blanket concepts
• Helium-cooled solid breeder concept with ferritic steel structure
• Dual-Coolant liquid breeder blanket concepts with ultimate potential
for self-cooling
– a helium-cooled ferritic structure with self-cooled LiPb breeder
zone that uses SiC insert as MHD and thermal insulator
– a helium-cooled ferritic structure with low melting-point molten salt
DEMO Blanket Testing in ITER
1. Detailed design of realistic
DEMO Blanket
2. DEMO relevant TBM
designed
ITER
3. TBM inserted for testing in
ITER port
Blanket
DEMO
Assessment of Dual Coolant Liquid Breeder
Blankets in Support of ITER TBM
SW/FW He
Inlet/Outlet
Grid Plates
Manifold
First Wall
DC-Molten Salt
Back Plate
He Inlet Pipe
LL Flow Outlet
Distribution Baffle
He Outlet Pipe
He Flow Inlet
Distribution Baffle
DC-PbLi
3-D Calculation for DC MS
Be zone
Perforated
plates
Front Flibe
Zone, Flow
Direction
Poloidal Up
FW Coolant
channels
He Manifold
Separator plate,
Back FLiBe Zone,
Flow Direction
Poloidal Down
Cross section
in OB blanket
at mid-plane
• Total TBR is 1.07 (0.85 OB, 0.22 IB). This is conservative estimate (no
breeding in double null divertor covering 12%)
• 3-D modeling and heterogeneity effects resulted in ~6% lower TBR
compared to estimate based on 1-D calculations
• Peaking factor of ~3 in damage behind He manifold
Preliminary Neutronics for DC PbLi Blanket
Blanket thickness
OB 75 cm (three PbLi channels)
IB 52.5 cm (two PbLi channels)
Local TBR is 1.328
OB contribution 0.995
IB contribution 0.333
If neutron coverage for double
null divertor is 12% overall TBR
will be ~1.17 excluding breeding
in divertor region. To be
confirmed by 3-D neutronics
Shield is lifetime component
Manifold and VV are reweldable
Magnet well shielded
Nuclear energy multiplication is
1.136
Peak nuclear heating values in OB blanket
o FS
36 W/cm3
o LL
33 W/cm3
o SiC
29 W/cm3
Two Types of Helium-Cooled SB PB modules under consideration for Testing in ITER
Type 1: Parallel Breeder and Multiplier:
Type 2: Edge-On configuration
NWL 0.78 MW/m2
75% Li-6 enrichment
Packing factor for Be
and Li4SiO4 Pebble
beds 60%
Local TBR: 1.2
In Demo
Configuration- 1-D
Local TBR: 1.04
In Demo
Configuration- 1-D
Nuclear Heating Rate in the Li4SiO4 Breeder
Wall Load= 0.78 MW/m2
10
Heating Rate in the Breeding Unit
NWL= 0.78 MW/m2
Nuclear Heating Rate in the Beryllium Multiplier
Wall Load= 0.78 MW/m2
Breeder1 100%
10
zone 4 Fe 100
zone 4 be 100
zone 5 fe 100
zone 5 be 100
zone 6 fe 100
zone 6 be 100
zone 7 fe 100
zone 7 be 100
Br fe 100
Br be 100
br br 100
10
Breeder2 100
Li4SiO4 Breeder
1
BE Be1 100
Be Be2 100
Be Be3
Be Be4 100
1
1
Structure
DCPB TBM with Parallel Breeding zones
Beryllium
0.1
0
10
20
30
40
50
Distance from Front Edge, cm
DCPB TBM with Parallel Breeding zones
0.1
0
10
20
30
Distance from Front Edge, cm
40
50
0.1
0
10
20
30
Distance from front edge (cm)
40
50
Neutronics module under evaluation to ensure that design goals are met
Goals
• Determine geometrical size requirements such that high spatial resolution for
any specific measurement can be achieved in scaled modules
• Allow for complexity, to maximize data for code validation
Proposed scheme is to evaluate two
design configurations simultaneously;
however 2-D (3-D) neutronics analysis
must be performed to ensure that
design goals are met.
Demo Act-alike versus ITERoptimized designs
• Structural fraction:
• ~ 23% Demo Vs. ~21% ITER
• Total number of breeder
layers/layer thickness:
• 10 layers/13.5 cm Vs. 8
layers/14.9 cm
• Beryllium layer thickness:
• 19.1 cm vs. 17.9 cm
2-D R-theta Model developed for DORT discrete ordinates calculations to analyze the
nuclear performance of the US two sub-modules with actual surroundings
Close-up radial details
Details of Theta
variation at the Port
Top View
Nuclear Heating in FW of The Two
U.S. Test Blanket Configurations in the
Toroidal Direction
9.5
Toroidal Profile of Tritium Production
Rate in each Breeder Layer of the Two
Test Blanket Configurations
5 10
-5
Right Configuration
Left Configuration
d (distance from front edge, mm)
9
4 10
Left Configuration
8.5
Layer#
-5
Layer#
Right Configuration
1
3 10
d= 0 mm
1
2
3
-5
8
2
3
2 10
d= 10.5 mm
7.5
-5
5
6
3
4
7
4
7
1 10
d= 21 mm
10
20
30
8
5
6
6.5
0
-5
40
50
60
Toroidal Distance from Frame, cm
70
0
0
10
20
30
9
40
50
Distance from Frame, cm
60
• Profiles are nearly flat over a reasonable distance in the toroidal direction
where measurements can be performed with no concern for error due to
uncertainty in location definition
• Steepness in profiles near the edges is due to presence of Be layer and
reflection from structure in the vertical coolant panels
70
US will Contribute to ITER Nuclear Analysis
US Contribution to ITER Nuclear Analysis will be in
the following areas:
 Nuclear analysis for ITER TBM
 Nuclear support for basic ITER Machine
 Development of CAD/MCNP interface
Level of effort will depend on availability of funding
Nuclear Analysis for TBM
TBM designs will be developed and
modeled for 3-D neutronics
calculations with all design details
The TBM 3-D model will be
integrated in the complete basic ITER
machine 3-D model
Perform 3-D neutronics calculations
using the integrated model
Neutronics calculations will provide
important nuclear environment
parameters (e.g., radiation damage,
tritium production, transmutations,
radioactivity, decay heat, and nuclear
heating profiles in the TBM) that
help in analyzing TBM testing results
Detailed Nuclear Analysis is Needed for ITER Basic
Machine During Design and Construction Phases
Cryo pump
duct
Cryostat
IVV
channel
Cryo pump
 ITER is still undergoing major
design changes
 As ITER moves toward
construction, more accurate
nuclear analysis becomes
essential part of final design
process
 Experience shows that neutronics
and radiation environment
assessments continue through
final design and construction
phases of nuclear facilities
– Examples include
• TFTR and JET
• Spallation Neutron Source
Nuclear Analysis for ITER Basic Machine
 This will include computation of radiation field, radiation
shielding, nuclear heating, materials radiation damage, and
absorbed dose to insulators and other sensitive components
 Three-dimensional neutronics calculations will be performed
using MCNP5 and FENDL/MC-2.0
 Activation analysis will be planned to support safety assessment
of the site-specific issues as needed. This includes calculating
radioactive inventory, decay heat, and maintenance dose
 Activation calculations will be performed using the state-of-theart ALARA pulsed activation code along with the FENDL/A-2.0
activation data
 Radiation leakage through holes and other penetrations must be
fully assessed to establish activation levels for personnel access
Neutronics Support for Module 18 of FW/Shield
(Baffle)
 We will provide neutronics
support for design and
construction of module 18
 3-D neutronics calculations
using the full ITER model will
be performed to determine
nuclear heating and radiation
damage in components of
module 18
Module 18
ITER diagnostics landscape
Nuclear Analysis for Diagnostics Ports
 Many diagnostics systems will be employed in ITER at upper,
equatorial and lower ports
 Neutron and gamma fluxes affect diagnostics performance
 Determination of radiation environment is essential for estimating
shielding requirements for diagnostic components such as
insulated cables, windows, fiberoptics and transducers, as well as
detectors and their associated electronics
 Radiation leakage through penetrations in these diagnostic
systems must be fully assessed to establish activation levels in and
near diagnostic equipment where frequent access will be
necessary
 We will coordinate with diagnostics group to provide needed
nuclear support
Neutronics Support for Heating and CD Systems
Will support Ion Cyclotron and
Electron Cyclotron heating and
current drive systems
These systems have sensitive
components (antennas, RF sources,
gyrotrons,
insulators,
and
transmission lines) and neutronics
support will be essential to address
radiation damage and streaming
issues
We will coordinate with plasma
heating group to provide
neutronics support as needed
HV DC
Supplies
RF Sources
Transmission Lines/
Decoupler/Tuning
Eight-strap
antenna
ITER ion cyclotron system block diagram
Neutronics Support for ITER Central Solenoid
 In FY04, we provided neutronics support to US effort on ITER CS
 Concern was that helium embrittlement could be a problem for
proposed JK2LB steel conduits (employed in place of Incoloy
908) with deliberate boron additives at some grain boundaries
 Concluded that helium embrittlement of JK2LB will not be a
problem, even with a 50% B concentration at the grain boundaries
 Will continue providing neutronics support as needed to ITER CS
Fusion Technology
Institute
CAD-Based MCNP
Parallel Computing Sciences Department
 Use Sandia’s CGM interface to evaluate CAD directly
from MCNP
» CGM provides common interface to multiple
CAD engines, including voxel-based models
 Benefits:
MCNP
MCNP
Native
Geometry
» Dramatically reduce turnaround time from
CAD-based design changes
– Identified as key element of ITER Neutronics analysis
strategy
» No translation to MCNP geometry commands
CGM
ACIS
OB
Pro/E
Voxels
ARIES-CS Plasma
IB
θT
– Removes limitation on surface types
– Robustness improved by using same engine for CAD and MCNP
– Provides 3rd alternative for CAD-MCNP link
» Can handle 3D models not supported in MCNP
θT
 Status: prototype using direct CAD query from MCNP
 Issues/plans:
» (Lack of) speed: 10-30x slower than unmodified MCNP
» Key research issue: ray-tracing accelerations (lots of acceleration
techniques possible)
» Support for parallel execution (CGM already works in parallel)
» Goal: speed comparable to MCNP, but using direct CAD
evaluation
IB
OB