Use of SCALE to Generate Neutron Cross Sections Analysis Activities

RIC 2012
Session TH28 - Thermal-Hydraulic and Severe
Accident Research
Use of SCALE to Generate Neutron Cross Sections
Libraries in Support of PARCS/TRACE Reactor
Analysis Activities
Mourad Aissa
U.S. NRC – Office of Nuclear Regulatory Research
March 15, 2012
Introduction
- The SCALE computer code was developed at Oak Ridge
National Laboratory (ORNL) for NRC beginning in 1976
- It is a multi-purpose computational system for analyses of
various nuclear systems and phenomena, including:
•
•
•
•
Criticality safety
Radiation shielding
Spent nuclear fuel and high level waste characterization
Reactor physics
– The scope of this presentation deals with the use of SCALE in
generating the neutron cross section libraries needed by the
PARCS/TRACE reactor analysis codes.
2
Reactor Core Analysis
ENDF
Data
Basic Data
Parameters
Cross Section
Library
Processing
AMPX 2000
}
2D Lattice Physics → 3D Nodal Simulation
3D Nodal Simulation
PARCS/TRACE
Nuclear
Data
Generation
SCALE
Converter
Code
GENPMAXS
2D Lattice Physics NEWT
Calculation
Libraries:
Continuous (point) data
Multigroup: 101-103 groups
Few Group
Database
3
1
TRITON Control Module
Lattice Physics Sequence
Scale Driver
And TRITON
Input
BONAMI / NITAWL
or
BONAMI / CENTRM / PMC
NO
All branches
complete?
Energy Detail
(10 orders)
NEWT
YES
COUPLE
Spatial/Angular Detail
(5 orders)
NO
All mixtures
complete?
ORIGEN-S
Es
YES
NO
All time
steps done?
Cf
Output Edits
(repeat for all requested
mixtures)
U
1
1
3
Pa
Th
OPUS
Ac
End
Rn
At
Po
Bi
Pb
1
1
Tl
1
1
3
Bk
Am
1
1
3
3
1
3
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1 210 211 212 213 214
1
1
1
1
1
3
2
3
2
3
3
1
1
1
3
2
3
3
2
1
1
2
1
1
3
3
1
1
1
1
1
1
1
1
1
253 254 255
1 247
243 244 245 246
1 242
1
1
1
1 236 237 238 239 240 241
1
3
1
1 235
Isotopic Details
(2227 nuclides)
230 231 232 233 234
1 219
3
1
1 252
248 249 250 251
1 229
220 221 222
1
1
1
1
3
1 223
1
3
1
1
1
1
3
3
1
1 224 225 226 227 228
1
3
3
1
1
1
3
3
1
1
1
3
1
Pu
Ra
1
1
1
1
Np
Fr
1
1
Cm
3
Explicit Yields (new)
2
Explicit Yields (old)
216 217 218
1 215
206 207 208 209
4
TRITON
Reactor Physics Analysis
• 2D Lattice physics analysis
– few-group cross-section preparation
– branch states
• Key features:
– continuous-energy resonance
self-shielding
– flexible
fl ibl geometry
t modeling
d li
– Depletion with ORIGEN, with most recent
data for more than 2000 nuclides
• Additional capabilities
– 3D Monte Carlo depletion
– 1D depletion using XSDRN
– 2D sensitivity and uncertainty analysis
5
Resonance Self-Shielding of
Multigroup Cross Sections
• CENTRM (Continuous Energy TRansport Module) calculates
1-D discrete ordinates continuous energy flux spectrum for a unit cell.
• PMC (Pointwise Multigroup Converter) uses continuous
energy flux and cross sections from CENTRM to generate
problem-dependent multigroup cross sections.
• CENTRM/PMC explicitly handles effects from the following:
– fissile material in the fuel and surrounding moderator
– overlapping resonances
CE Resonance Processing
– anisotropic scattering
– inelastic level scattering
M acroscopic A bsorption C ross Sections
Σ a 0 - 10 ev
1 ×1 0
2
1 ×10
1
1 ×1 0
0
235
U
U
Pu
Pu
241
Pu
Σ abs
Σ total
238
239
• BONAMI
Bondarenko factors for unresolved
resonance region
1
-
240
)m
c
-1
(1 ×1 0
Σ
1 ×10
-2
1 ×1 0
-3
-4
1 ×10 0
2 35
-4
U = 1.57 x10
2 38
-2
U = 2.16 x10
-3
2 39
Pu = 1.03 x10
2 40
2 41
2
-5
Pu = 6.43 x10
-6
Pu = 4.34 x10
4
6
8
10
Energy (ev)
6
2
NEWT
2-D Multigroup Transport Solver
• NEWT (NEW Transport Algorithm) is a two-dimensional transport
solver
– uses discrete ordinates (solves for angular fluxes)
– Is based on the Extended Step Characteristics (ESC) method
– uses arbitrary geometry with automated grid generation
– calculates a steady-state flux solution
– calculates lattice physics data (few-group cross-sections,
assembly discontinuity factors, kinetic parameters, etc.)
BWR 8x8
VVER 440
ACR 700
7
BWR Lattice Physics Analysis using
SCALE and PARCS
• Recent evaluation of Peach Bottom Unit 2
Cycles 1 and 2 using SCALE 6.1:
– SCALE evaluated10 lattice designs.
– Each lattice design depleted at 6
different in-channel void fractions and
Control Rod (CR) conditions (0%,
40%, 80% with CR in and out).
– Each depletion calculation modeled at
~40 depletion steps and 12 branch
cases.
– The analysis did over 4,800
evaluations to generate few-group
cross-section database.
Peach Bottom Unit 2 Cycle 1 Lattice
BOC Cycle 2 Power Distribution
8
Evolution of SCALE Lattice Physics
– SCALE 5.0 (2004)
• initial release of TRITON/NEWT
– SCALE 5.1 (2006)
• double-heterogeneous XS processing capability
• coarse Mesh Finite Difference (CMFD) acceleration
• generation of database file for GENPMAXS
– SCALE 6.0 (2008)
• prismatic assemblies
– SCALE 6.1 (2011)
• CMFD for prismatic assemblies
• parallel computation of branch cases
• support for user-input Dancoff factors
9
3
Future Lattice Physics Enhancements
– Run time improvements
• Reduce run time by approximately a factor of 5.
• Modernization of cross-section processing, transport, and
depletion modules.
• Improved
I
d parallelism
ll li
– New capability
• Embedded Self-Shielding Method (ESSM) for cross-section
processing.
• New Method of Characteristics flux (MoC) solver similar to
industry codes.
– Simplified user interface
• Simplified LWR-specific input will be ~100 lines.
• Simplified lattice physics output.
10
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