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1、This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may not be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an agency of the Unite
2、d States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third partys use, or the results of such use, of any information, apparatus, product or proc
3、ess disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the United States Government or the sponsoring agency. INL/CON-05-00662 PREPRINT Comparison Of The 3-D Deterministic
4、Neutron Transport Code Attila To Measure Data, MCNP And MCNPX For The Advanced Test Reactor M&C 2005 International Topical D. S. Lucas H. D. Gougar T. Wareing G. Failla J. McGhee D. A. Barnett I. Davis Douglas Lucas September 2005 Comparison of the 3-D Deterministic Neutron Transport Code Attila To
5、Measured Data, MCNP and MCNPX For the Advanced Test Reactor D. S. Lucas1,H. D. Gougar1, T. Wareing2, G. Failla2, J. McGhee2, D.A. Barnett2and I. Davis2 1Idaho National Laboratory, P. O. Box 1625, Idaho Falls, Idaho 83415-3885 Douglas.Lucasinl.gov 2Transpire Technologies, 6659 Kimball Dr., Suite D-40
6、4, Gig Harbor, WA 98335 Abstract An LDRD (Laboratory Directed Research and Development) project is underway at the Idaho National Laboratory (INL) to apply the three-dimensionalmulti-groupdeterministicneutrontransportcode (Attila) to criticality, flux and depletion calculations of the Advanced Test
7、Reactor (ATR). This paper discusses the development of Attila models for ATR, capabilities of Attila, the generation and use of different cross-section libraries, and comparisons to ATR data, MCNP, MCNPX and future applications. 1.0 Introduction This report discusses the Advanced Test Reactor (ATR),
8、 a brief overview of the Attila 1 three-dimensional deterministic neutron transport code, the model development for ATR with Attila and the SolidWorks CAD tool, the results of the comparisons to fresh core data and depletion benchmarks. Additional discussion is given on future plans for the Attila c
9、ode at INL. The Advanced Test Reactor is operated and maintained by the Idaho National Laboratory (INL) for the Department of Energy (DOE). ATR tests and experiments are responsible for much of the worlds data on material response to reactor environments. The ATR has nine flux traps in its core and
10、achieves a close integration of flux traps and fuel by means of the serpentine fuel arrangement shown in Figure 1.0. Figure 1.0 ATR Reactor and Core The nine flux traps within the four corner lobes of the reactor core are almost entirely surrounded by fuel, as is the center flux trap position. The r
11、emaining four flux trap positions have fuel on three sides. Experiments can be performed using test loops installed in some flux traps with individual flow and temperature control, or in reflector irradiation positions using the primary fluid as coolant. Five of the flux traps are equipped with inde
12、pendent test loops and four are used for drop-in capsules. The ATR also uses a combination of rotational control cylinders (shims), and neck shim rods that withdraw vertically to adjust power while maintaining a constant axial flux profile. The power level (or neutron flux) of the flux trap position
13、s in ATR can be adjusted for irradiation requirements. Maximum total power is 250 MW (thermal) in ATR. Balancing maximum ATR full power distribution results in as much as 50 MW produced in each lobe. Power shifting allows for a maximum and minimum lobe powers of 60 and 17 MW. 2.0 Attila Problem Solv
14、ing Capabilities Attila uses the standard first order steady state form of the linear Boltzmann Transport Equation (BTE) 1: ()() ()()()() , tSf d r Er Er EQr EQr Eq r E ds + = + + ? (1) where ()() () 04 , Ss Qr Er EEr EddE = ? ?(2) and () ( ) ()() 04 , ff E Qr Er Er EddE k = ? (3) wheredenotes the a
15、ngular flux,/d dsis the directional derivative along the particle flight path, is a unit vector denoting the particle direction, t denotes the total macroscopic interaction cross section (absorption plus scattering), s denotes the differential macroscopic scattering cross section,is the fission spec
16、trum, f denotes the fission macroscopic cross section,is the mean number of fission neutrons produced in a fission andqdenotes a fixed source. This is the basic form of the transport equation solved by Attila. Attila uses multi-group energy, discrete-ordinate angular discretization and linear discon
17、tinuous finite-element spatial differencing (LDFEM). The LDFEM spatial discretization is third-order accurate for integral quantities and provides a rigorously defined solution at every point in the computational domain. The general solution technique within Attila is source iteration. Both k-eigenv
18、alue (Keff) and fixed source modes are supported, including coupled neutron-gamma calculations. Attila also has depletion capability. Attila uses a built in code called Fornax. Fornax solves the fully coupled equations for the production, depletion, and decay of nuclides using a series expansion app
19、roximation to the matrix exponential solution. Short time constant products are treated separately using the same algorithm as in the ORIGEN code. Fornax supports an arbitrary number of fissile species. Separate data for up to 99 metastable states are supported for a given nuclide. Default data for
20、1307 nuclides, including half lives, three group reaction cross sections, and fission product yields are provided in an XML data file (fornax.xml, 30,000 lines) based on an ORIGEN-S data set. Special DTF cross sections files were developed to support the burn, including detailed KERMA values for pow
21、er normalization and cross sections for the individual capture reactions. For representative problems Fornax typically solves 20-75 burn zones per second on a single CPU 2.0 GHz Athlon Linux system. 3.0 Cross-Section Libraries The COMBINE 2 code was used and modified to develop a four group ENDF-5 a
22、nd ENDF-6 set of cross section (XS) libraries for Attila that gives the cross section libraries in Data Table Format (DTF) for fresh fuel configurations. This avoids having to use translation programs written in C or Fortran for ANISN to DTF data table structures. All data processing with COMBINE us
23、ed an ATR energy spectrum combining the fast and thermal regions in COMBINE. Resonance treatment was used for those materials that have resonance data in the ENDF-5 and ENDF-6 cross-section sets. Testing was performed on the cross-section libraries for assurance of reasonable values. The Combine XS
24、set was compared to the Hansen-Roach cross-section library using the Venus Reactor test provided with Attila. Transpire also supplied cross section sets that are being used for comparisons. The Transpire cross-section sets are based on the ENDF-6 formulation with NJOY for neutrons and gammas with de
25、pletion (burn). Additional work is being undertaken by Transpire using SCALE to develop a collapsed eleven group burn cross section set in AMPX format. This work is based on a two-dimensional ATR core regional model for separate 2D cross section sets for the fuel, reflector, shims and other elements
26、 of the ATR reactor core. Transpire has an SBIR with DOE for fiscal year 2006 to couple the ENDF-6/7 libraries to Attila directly for the generation of 2D/3D regional cross section sets. 4.0 Model Development This section discusses some of the model development efforts for non-depleted fuel and depl
27、eted fuel problem comparisons. 4.1 Attila ATR 3D Model Geometric and material information for the Attila 1 ATR model, which includes atom mixture densities and atom fractions, were obtained from ATR core calculations using the ATR MCNP 3 model. Additional models discussed in this report also used in
28、put from MCNP and MCNPX. Geometry parameters for the Attila calculations were generated using SolidWorks, (SW), a computer aided CAD design system. The CAD assembly allowed test section modifications and control drum (shim) rotations. The ATR Attila model included the structure of the reactor on the
29、 top, bottom and perimeter of the reactor core. In order to compare the Attila ATR model with MCNP, the 19 radial plate fuel elements were homogenized into 3 radial sections. The CAD assembly was exported to Attila through the Parasolid format. Attila preserves the original CAD component names in th
30、e translation, aiding the assignment of region-wise material properties. Attilas graphical user interface (GUI) was used for the full analysis, including mesh generation, material assignments, boundary conditions and the creation of post processing edits. The code can be executed in the GUI setup or
31、 separately as a solver. The computational model for the Attila ATR model included approximately six hundred thousand tetrahedral elements with 5 axial layers for the core region and one layer each for the top and bottom. The top and bottom layers were a mixture of aluminum and water, based on the M
32、CNP geometry. The outer regions of the Attila ATR model use an unstructured mesh while the core, loops and fuel can be modeled using specified layers for the mesh. This recent capability allows finer detail for the fuel, absorber regions (shims) and experiments. Figures 1.0, 2.0 and 3.0 provide VisI
33、t 4 illustrations of the solid geometry for the ATR, the fuel and structured portion of the mesh for the Core Internals Changeout (CIC) configuration used in the data comparison of the Attila ATR 3D model. VisIt has been coupled to Attila for plots by the High Perfromance Computer (HPC) group of the
34、 INL. Figure 2.0 3D ATR Core SectionFigure 3.0 ATR Core Layered Mesh 4.2 Attila and MCNP Toy ATR Models Additional model development was performed for a simplified 3D ATR MCNP model, referred to herein as the Toy ATR model developed by Bruce Schnitzler of the INL. It is also being used for compariso
35、ns of the Attila, MCNP-MOCUP 5 and MCNPX 6 codes for depletion analysis. In constructing the Attila Toy ATR model, the same approach was used with Solidworks and the Attila mesher. The Toy ATR model geometry is illustrated in Figure 5.0. The Toy model consists of an aluminum barrel, a Beryllium core
36、 containing the six fuel elements of 93% enriched U-235, shim absorbers, the lower three shims having Hafnium pointed inward, shown in yellow and the upper three shims pointed outward. There are seven interior targets and six outer targets with the same geometrical configurations consisting of Neptu
37、nium 237, Np-237. The Toy mesh used in this study consisted of 140 K (thousand) terahedrons (tets). Attila allows fission and depletion of the U-235 and NP-237 in the calculations. It should be noted that for deterministic codes such as Attila which use the Finite Element Method (FEM) it is importan
38、t to use a large number of elements in the fuel and absorber regions to approximate the volume correctly. Since the mesh generator places points on the solid model geometric surface the polygons are inscribed. To obtain an exact volume points for the polygons would have to be placed outside the soli
39、d geometry surface. 4.3 Attila and MCNPX Godiva Models A more elementary model used to benchmark the depletion capability was that of the well known Godiva 7 problem. The Godiva facility, shown in Figure 6.0, was one of the experiments performed at Los Alamos in the 1950s to determine the critical m
40、ass of a bare 94% enriched U-235 sphere which consisted of two identical sets of nested hemispheres. Figure 7.0 illustrates the solid geometry by VisIt for a half-section of the Godiva model. The Figure 5.0 Toy ATR Model MCNPX modified Godiva model problem has a fuel region which consists of two con
41、centric spheres, a larger water region for neutron moderation and a thin iron outer shell. The number of tets used in this model was 70 K. For models such as this which consist of concentric spheres CAD tools normally allow “mating” the surfaces together. However, in some instances the Parasolid or
42、interface file between the CAD program and the mesher utility allows extra numerical “slop” that results in some of the mesh not appearing. This is overcome during the assembly process of the CAD code by placing the concentric bodies at the origin. Figure6.0GodivaMultiplicationConfigurationFigure7.0
43、GodivaSolidGeometry 4.4 Attila and MCNPX 7 Can HEU Models The last model discussed in this report is entitled the 7-Can HEU Test Problem by the authors of the MCNPX depletion code, shown in Figure 8.0 and a solid geometry section view in Figure 9.0 for Attila. 5 It consists of seven aluminum cans wi
44、th 5% enriched U-235 in the lower portions of the can and a void in the upper part of the can. The cans are surrounded by air. The model consists of approximately 50 K mesh elements. Figure8.0Figure9.0 7-CanHEUSolidGeometry 5.0 Calculations and Results This section provides the highlights of the cal
45、culations and results obtained for the models discussed in Section 4.0 of this report. 5.1 Comparison of 3D Attila Model to MCNP and Test Data The first calculation discussed in this section will be for the 3D Attila ATR model, shown in Figure 1.0, compared to the measured data from the 1994 Core In
46、ternals Changeout (CIC) 8 performed for the ATR. After the Beryllium reflector block was taken out and replaced with a new reflector block and fresh fuel, measurements were taken using flux wands in the water gaps of the fuel elements. The forty fuel elements are arranged in a serpentine pattern as
47、shown in Figure 1.0. The flux wands were placed in the water gaps between the eighteenth and nineteenth fuel plates. The experiments were also repeated in the ATRC (Advanced Test Reactor Critical Facility), a miniature low-power version of the ATR. MCNP models were compared to the data taken from th
48、ese tests. Edits were used in the Attila code for fission (n,f) reactions for the power. These edits are also available for plotting using VisIt. This calculation was performed using the original Radion cross-section libraries. The model used was that of Figure 1.0 with 622 K tetrahedral elements. T
49、he calculation was performed on a 2CPU Opteron with a locally parallel version of Attila. The run time was approximately 24 hours. The Kefffor Attila was 1.015 compared to a Keffcomputed by MCNP of 1.0012. The results shown in Figure 10.0 are for the ATR Attila model compared to the test data from ATR and ATRC along with comparisons to MCNP. The MCNP results compare well with the ATR data while Attila compares favorably with the ATRC data. The cross section set used a default fission spectrum from NJOY given by 1.036 ( )0.453sinh2