Evaluation of the DRAGON code for VHTR design analysis.

Taiwo, T. A.; Kim, T. K.

doi:10.2172/929210

Cited by 8 publications

(3 citation statements)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The DRAGON 4 code has been successfully used in high temperature reactor (HTR) applications [15], and it is a well established nuclear analysis tool.…”

Section: Reactor Core Configurationmentioning

confidence: 99%

Deterministic Modeling of the High Temperature Test Reactor

Ortensi¹,

Cogliati²,

Pope³

et al. 2010

View full text Add to dashboard Cite

Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability for the Next Generation Nuclear Power (NGNP) project. In order to examine INL's current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19-fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal-z full-core solver used in this study and is based on the Green's Function solution of the transverse-integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, as well as the deterministic transport code INSTANT, provide benchmarking capability for the DRAGON and HEXPEDITE. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2-3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The uncertainty in the graphite impurity appears to be the main source of the error, whereas inaccuracies in the ENDF/B-VII graphite and U 235 cross-sections have a secondary effect. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross-section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross-sections. vii viii CONTENTS Abstract

show abstract

“…The DRAGON 4 code has been successfully used in high temperature reactor (HTR) applications [15], and it is a well established nuclear analysis tool.…”

Section: Reactor Core Configurationmentioning

confidence: 99%

Deterministic Modeling of the High Temperature Test Reactor

Ortensi¹,

Cogliati²,

Pope³

et al. 2010

View full text Add to dashboard Cite

show abstract

“…This density has been determined to conserve the total number of graphite atoms in the whole core calculation (see Appendix C). This solution has been also used by other authors confronted by the same problem 22,23,24,25 and it appears to be accurate.…”

Section: Description Of a Fuel Block And Boundary Approximationmentioning

confidence: 96%

Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

Descotes¹

2011

View full text Add to dashboard Cite

In Very High Temperature Reactors (VHTRs), the long mean-free-path and large migration area of neutrons leads to spectral influences between fuel and reflector zones over long distances. This presents significant challenges to the validity of the classic two-step approach of cross section preparation wherein infinite lattice transport calculations are performed on relatively small physical domains (e.g. single assembly) in order to compute homogenized few-group cross sections for whole core analysis.Effects of the inner and outer reflectors challenge the classical two-step approach, while burnable poison locations affect neighboring assemblies as well. Use of transuranics-only (TRU) Deep Burn fuel in a prismatic VHTR (DB-VHTR) presents the additional challenge of producing vastly different neutron spectra between fresh and burned fuel.Modeling the VHTR also requires the ability to treat in the code the hexagonal geometry, the presence of TRISO particles, whose distribution in the fuel compacts is stochastic, and large lattice domains. Currently, DRAGON can treat these features: the hexagonal geometry is in place, the large domains may be treated by the method of characteristics (MOC) or collision probability (CP), and the doubleheterogeneity treatment can be used for TRISO particles.Therefore, in order to complement the analysis and design activities for Deep Burn in prismatic High Temperature Reactors (HTRs), this study was performed in order to provide some initial guidance on the necessary procedures to model the DB-VHTR with sufficient accuracy. The purpose of this study is to provide evaluation of the tradeoffs in accuracy between domain size in the lattice calculation and number of energy groups used in the whole core calculation. The lattice calculations were performed using the DRAGON-4 code and the whole core calculations using the Idaho National Laboratory code INSTANT, both of which are described in the following sections. The accuracy of the whole-core calculations was judged by comparison of Eigenvalues and ¿ssion rate shapes to results from MCNP calculations. The number of groups used in the lattice calculations was fixed at 295 and two different domain sizes were used, a single block and a supercell. Primarily of interest in this study were the effects of the reflector on the spectrum and, in turn, the adequacy of the cross section sets.Also included in this report are results from a preliminary study aimed at determining the effect of the presence of reflector on the burnup characteristics in a peripheral block. Depletion was performed in a supercell in order to observe the differences across a block of interest.The results showed that using the Method of Characteristics (MOC) in DRAGON, both a single block and a supercell leads to good predictions of the Eigenvalue and of the relative fission rates of each fuel block in the core if enough groups are kept in INSTANT (e.g., 26 groups). The supercell path produced the best results compared to the equivalent calculations using the single-block path....

show abstract

“…The pre-transient rod configuration for transient #2855 had transient rods 66.25% withdrawn, control/shutdown 49.53% withdrawn, and compensation/shutdown rods 100% withdrawn. After changing the transient rod positions to 100% withdrawn, the relative power distribution was calculated with MCNP (using F6 tallies in neutron/photon mode assuming 193.7MeV/fission [18]) for a cold core (cross sections set at room temperature). Using the measured fuel-graphite heat capacity as a function of temperature [13] and the MCNP-calculated relative power distribution, the average temperature of every fuel assembly was calculated for total core energy steps of 100, 200, 300, 400, 500, 1000, 2000, 3000, 4000 and 5000 MJ.…”

Section: Core Average and Hot Spot Temperaturesmentioning

confidence: 99%