studsvik’s next generation nodal code · pdf filethe basic geometry unit in simulate-3...

Advances in Nuclear Fuel Management IV (ANFM 2009) Hilton Head Island, South Carolina, USA, April 12-15, 2009, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2009)
ANS 2009, Topical Meeting ANFM 2009, p. 1/12
STUDSVIKS NEXT GENERATION NODAL CODE SIMULATE-5
Tamer Bahadir Studsvik Scandpower, Inc.
1087 Beacon St. Suite #301, Newton, MA 02459-1700, USA [email protected]
Sten-rjan Lindahl
Studsvik Scandpower AB Hantverkargatan 2A, SE-722 12 Vsters, Sweden
[email protected]
Keywords: SIMULATE-5, Nodal Method, Core-Tracking, Gamma Scan ABSTRACT
SIMULATE-5 is Studsviks next generation nodal code which has been developed along with CASMO-5, Studsviks next generation lattice physics code, to address the deficiencies of existing reactor physics tools for todays aggressive core designs with more heterogeneous fuel, extended cycle lengths, and operation strategies. This paper discusses the new neutronic and thermal-hydraulics models employed in SIMULATE-5 as well as benchmarking activities against theoretical problems, critical experiments, core follow calculations, and gamma scan evaluations.
1. INTRODUCTION
Studsviks nodal code SIMULATE-31 has been in use for LWR reactor analysis for nearly 25 years. The model has been proven to be accurate for typical LWR applications. Todays aggressive core designs with more heterogeneous fuel and extended cycle lengths and operation strategies require more comprehensive reactor physics tools. New emerging issues in the industry also demand improved accuracy in pin powers and hence, thermal margins related to pin loads. To address these concerns Studsvik has developed the next generation lattice physics code CASMO-52 and the nodal code SIMULATE-53,4,5. (Note that the development of the new nodal code was started under the product name SIMULATE-4, and following the completion of the development, the new code was renamed SIMULATE-5.)
SIMULATE-5 is largely based on true physics and true geometry and avoids ad hoc models. In addition to improving accuracy, SIMULATE-5 provides more detailed information about the reactor core and its components than SIMULATE-3.

T. Bahadir, et al, Studsviks Next Generation Nodal Code SIMULATE-5
In the first part of this paper, the geometry, cross section, neutronic, and thermal-hydraulics models employed in SIMULATE-5 are summarized. Various validation results are presented in the second part.
2. MODELS
2.1 Geometry
The basic geometry unit in SIMULATE-3 is the node, defined as 1/N:th of an assembly (typically, N=24). The conventional node may contain several material zones and hence, be strongly heterogeneous.
In SIMULATE-5 the basic unit is the subnode, defined such that it is materially homogeneous in the axial direction. The complexities of spacers, control rods and their zonings, enrichment/BA zoning, and reflector material at the assembly end points are thus all taken into account.
The subnode layout may differ from one assembly design to another. Also, as control rods move, the boundaries move.
Burnup and nuclide data are stored subnode-wise. Thermal margins are evaluated in this geometry. Hence, the uncertainty of the meaning of burnup or thermal margin in a heterogeneous, conventional node is avoided.
In the x/y direction, an assembly is divided into NN submeshes (typically N=5) where the submeshes follow pin cell boundaries. For BWRs, the outer level of submeshes is made up of the water gap region. For PWRs, the outer submesh layer is chosen to capture intra-assembly mismatch effects. Submesh cross sections and discontinuity factors are generated from CASMO-5 parallel to assembly average data.
The conventional node concept (sometimes also called 3D nodes below) is used also for SIMULATE-5. The node is used for coupling the axial subnode and the radial submesh calculations as will be described later.
2.2 Cross section model
The cross sections are evaluated by a hybrid macroscopic/microscopic model:
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Approximately 50 isotopes (17 actinides and 30+ fission products/burnable absorbers) are tracked for each node. Nuclides have been chosen based on their impact on reactivity and on their interest from a safeguard point of view (special nuclear materials).
In addition to these 50 isotopes, five important actinides (U235, U238, Pu239, Pu240, Pu241) are tracked for each submesh in order to accurately describe the internal radial skewness of the nuclide distribution of an assembly. The cross section model faithfully reproduces history effects. The shutdown cooling phenomenon is automatically taken

into account and as-built enrichment and fuel weight appear naturally in the hybrid model.
SIMULATE-5 includes a new control rod depletion module for BWRs. The new model keeps track of the depletion of active absorber material (either B10 and/or Hf) and the fast fluence for each wing of the control rod with detailed axial zoning. The reactivity effect of the depletion of an active absorber is considered via cross section feedback to the nodal solution.
2.3 Neutronic Model
SIMULATE-5 solves the multi-group nodal diffusion equation as described below3. Macroscopic and microscopic cross sections and assembly discontinuity factors are generated from single-assembly CASMO-5 calculations with the assumption of a zero current boundary condition. The number of energy groups can be any subset of that used in the two-dimensional transport solver of CASMO-5 (typically 19 for UO2 or 35 for MOX fuel).
In the case of cores loaded with MOX fuel, it is known that methods based on P1 or diffusion theory are not capable of capturing the strong transport effects at the MOX-UO2 interface. Accordingly, the higher order SP3 (simplified P3) method is implemented in SIMULATE-5 as an option.
The core flux is found by a three-step approach:
1) The 1D diffusion equation is solved, one assembly at a time using the subnode geometry3. The radial leakage, obtained from the 3D solution (step 3) is converted into an equivalent absorption. The obtained detailed 1D flux is employed to compute flux weighting factors needed when computing homogenized cross sections for the conventional nodes. Also, axial discontinuity factors are computed.
2) The 2D diffusion equation is solved, one axial plane at a time, relying on the submesh geometry3. The axial leakage, known from the 3D solution, is converted into an equivalent absorption. The resulting flux solution is then used to compute homogenized cross sections and radial discontinuity factors for the 3D nodes. The 2D submesh approach overcomes the simplifications of the original CASMO-5 solution where homogenized cross sections and discontinuity factors are based on zero net current boundary conditions.
3) The 3D diffusion equation is solved using conventional 3D nodes with homogenized cross sections and discontinuity factors obtained as indicated above. The analytic nodal method (ANM) solution technique is employed. The only approximation made in ANM (within diffusion theory) is for the estimate of the shape of the so-called transverse leakage. However, the shape is known in some detail from the 2D submesh solution, and the traditional quadratic shape approximation need not be made.

Note, that the 3D solution is needed to tie the 1D axial and the 2D radial solutions together.
2.4 Pin Power Reconstruction Model
The flux and power of a pin in a 3D node is computed by synthesizing the 1D, 2D, 3D, and CASMO-5 solutions4.
The 2D solver provides a detailed so-called homogeneous flux in the x/y directions of each submesh. This flux takes into account radial variations of burnup, isotopes, fuel temperature, xenon and density inside the assembly. From the 1D and 3D solvers the homogeneous axial flux shape is known for each pin and subnode of the full 3D node. Finally, CASMO-5 form functions are superimposed subnode-wise on the homogeneous flux to produce pin powers. Note, that the pin power may be discontinuous from one subnode to another if there are material discontinuities.
Pin burnups are found by time integration of pin powers and taking into account the pin wise loadings as opposed to modulating the node homogenous burnups with exposure form functions.
The neutron detector response is computed using the reconstructed flux at the detector position. The gamma detector response in a BWR is found using pin powers of the four surrounding assemblies and detector response functions known from CASMO-5:
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The BWR assembly may be divided into four radial sub-channels, which communicate via cross flow (or are closed for SVEA type fuel). The cross flows are determined by solving lateral momentum equations.
The PWR core treats assembly cross flow by solving the axial and lateral momentum equations as in COBRA IIIC6.
Thermodynamic quantities are evaluated by using the NIST/ASME7 steam/water function library.
3. BENCHMARKING
3.1 Critical experiments
The widely used B&W 18108 series of critical experiments which provide high quality pin-by-pin fission rate (power) measurements are used for validatin

studsvik’s next generation nodal code · pdf filethe basic geometry unit in simulate-3...

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