Fission Density Distribution Research Articles

A Novel Fission-Matrix Based Radial Boundary Correction Methodology and its Implementation into the RAPID Code System

The RAPID code system utilizes a novel combined fission matrix methodology to rapidly solve the whole-core eigenvalue problem. To represent axial and radial reflector regions, the fission density distribution in the boundary assemblies, a fission density correction factor methodology was devised and demonstrated. Although, accurate results were achieved, because of the need for a full core Monte Carlo calculation, this technique requires significant computing resources. To address this issue, this paper introduces a novel FM based methodology, which achieves accurate results, however at very small computing expense. The methodology has been implemented into the RAPID code system, and successfully demonstrated for the Watts Bar whole core OECD/NEA

High Fidelity Whole Core Reactor Eigenvalue Calculation using the Hybrid Monte Carlo and Deterministic RAPID Code System

This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for whole core reactor simulation. The RAPID code is an implementation of the Multistage Response function Transport (MRT) methodology resulting in a hybrid Monte Carlo and deterministic technique. This paper utilizes the Watts Bar Unit 1 benchmark as the computational model to exam the accuracy and efficiency of RAPID compared to Serpent. The RAPID calculated eigenvalue and 3-D fission density distribution are compared to the results from Serpent. This paper shows that RAPID is able to obtain pin-wise, axially-dependent fission density in 7 minutes using one computer core, as opposed to assembly-wise, axially dependent fission density in 8 days on 40 computer cores.

A Consistent Low-Order Acceleration Technique for Coarse Mesh Transport (COMET) Methods

The COarse MEsh Transport (COMET) method, a hybrid continuous energy stochastic and deterministic transport method/tool based on the incident flux response expansion theory, is capable of providing highly accurate and efficient continuous energy whole-core neutron solutions to various heterogeneous reactor cores. In this work, a novel low-order (zeroth-order) acceleration technique is developed to significantly improve COMET’s computational efficiency for core calculations. This new method is based on consistent coupled low-order and high-order calculations to obtain the COMET core solution. In the low-order calculations, COMET is used to converge the total partial current escaping from each coarse mesh and the core eigenvalue. The resulting fixed-source problem in which the off-diagonal terms (equivalent to the scattering and fission neutron sources) are constructed by the zeroth-order solution are efficiently solved by the high-order COMET calculations. The resulting high-order angular flux on each coarse mesh bounding surface is then used to update (collapse) the low-order response coefficients. The coupled low-order and high-order calculations are repeated until both the eigenvalue and the low-order response coefficients are converged. The new acceleration method is implemented into COMET and tested in a set of stylized Advanced High Temperature Reactor (AHTR) benchmark problems. It is found that the core eigenvalues and the local fission density distributions predicted by COMET with the low-order acceleration agree very well with those computed by the original COMET. The eigenvalue discrepancy varies from 0 to 1 pcm, and the average relative differences in the stripewise and assembly-average fission density distributions are in the range of 0.021% to 0.032% and 0.004% to 0.01%, respectively. The comparisons have shown that the new low-order acceleration method can maintain COMET’s accuracy while improving its computational efficiency for core calculations by 12 to 16 times.

Continuous-Energy Time-Dependent Coarse Mesh Transport (COMET) Method for Kinetics Calculations

In this paper, the novel continuous-energy coarse mesh transport (COMET) method is extended to perform time-dependent neutronics calculations in highly heterogeneous reactor core problems. In this method, the time-dependent transport equation is converted into a series of steady-state transport equations by estimating the time derivative term using implicit finite differencing. The resulting steady-state transport equations, having additional terms that are imbedded in the total collision term and in the volumetric source terms, are then solved by the steady-state COMET method, in which all the phase-space variables, including energy, are treated continuously. Finally, the fission density distribution constructed by the steady-state COMET is used to solve a set of ordinary differential equations to obtain the delayed neutron precursor concentrations. The time-dependent COMET method is benchmarked against a direct continuous-energy Monte Carlo method (i.e., MCNP) in a set of infinite homogeneous problems and a set of single-assembly benchmark problems consisting of identical pin cells. It is found that the COMET results agree very well with the Monte Carlo reference solutions while maintaining its formidable computational speed (orders of magnitude faster than the Monte Carlo method).

Additional Stylized 3-D Benchmark Problem Set Based on the Pin-Fueled SmAHTR

Previous work presented a set of stylized three-dimensional benchmark problems based on the Oak Ridge National Laboratory (ORNL) preconceptual design of a fluoride-salt-cooled small modular advanced high-temperature reactor, or SmAHTR, with prismatic assemblies fueled by tri-isotropic (TRISO) particles. That previous work created a detailed description of the benchmark problems by closing several outstanding design gaps from the ORNL preconceptual design report, notably by addressing the lack of active control mechanisms, for which control rod “bundles” were implemented. In this paper, the creation of two additional stylized benchmark problem sets based on that past work is detailed, offering two new control rod configurations. The fluoride salt, small size, and highly heterogeneous TRISO-fueled pins make these additional benchmark problem sets useful numerical validation references in benchmarking neutronics tools against continuous-energy stochastic Monte Carlo results. Detailed reference results, including the eigenvalue (keff) and 1/11th assembly-averaged relative fission density distributions, are provided for both control rod configurations in full-core cases with all control rods withdrawn and all control rods fully inserted. A near-critical core benchmark problem and results are provided for one configuration. The provided results are calculated using the continuous-energy Monte Carlo code MCNP.

A Stylized 3-D Benchmark Problem Set Based on the Pin-Fueled SmAHTR

In this paper, a set of stylized numerical benchmark problems is developed. These problems are based on the Oak Ridge National Laboratory preconceptual design of a fluoride-salt-cooled small modular advanced high-temperature reactor, or SmAHTR, that uses prismatic fuel assemblies with cylindrical pins/rods containing tri-isotropic fuel particles. A detailed description of the benchmark problems is achieved by closing several outstanding design gaps and modifying the coolant channel shape to reduce bypass flow for improved coolant and fuel temperature distributions. The benchmark problems, while stylized, retain the important thermal-hydraulic and reactor physics features (e.g., fuel particles) necessary for benchmarking tools for reactor core analysis. In addition to the full description, detailed reference results such as the eigenvalue (keff) and fuel pin and assembly-averaged fission density distributions are provided for five benchmark problems: full-length fuel assemblies with control rods fully withdrawn and inserted, and full core with all control rods withdrawn, all control rods fully inserted, and some control rods fully inserted (near-critical core). The provided results are calculated using the continuous-energy Monte Carlo code MCNP.

COMET neutronics solutions to the prismatic VHTR benchmark problem

Recently, the hybrid stochastic deterministic coarse mesh radiation transport method COMET for whole core criticality calculations was extended to treat the energy variable continuously consistent with its treatment of the rest of the phase space variables. In this paper, a surface-dependent asymptotic spectrum expansion scheme was used by the continuous-energy COMET to obtain the transport solutions to a stylized Very High Temperature Reactor (VHTR) benchmark problem in which all levels of heterogeneity that are of reactor physics importance are preserved. Because of its complex geometry and material distribution, this benchmark problem is challenging to solve with deterministic neutronic methods, even those based on transport theory. The eigenvalue and the fuel pin fission density distribution for all control rods fully inserted and fully withdrawn configurations were compared to the corresponding values calculated by the continuous-energy Monte Carlo code MCNP. The agreement between the two codes was found to be excellent. The eigenvalue difference was 66 pcm for the uncontrolled configuration and 69 pcm for the controlled configuration. The corresponding average pin fission density differences were 0.36% and 0.53% for the uncontrolled and controlled configurations, respectively. It was also found that COMET is 4–5 orders of magnitude faster than MCNP. This indicates that COMET has continuous-energy Monte Carlo like accuracy but with formidable computational speed for solving problems with multiple levels of heterogeneity such as the VHTR problem.

Consistent spatial expansion functions on hexagonal surfaces for use in radiation transport methods

Some of the radiation transport solvers that are based on expansion function methods in hexagonal geometry generally use projection techniques that transform the spatial dependence of the angular flux in hexagonal geometry to Cartesian geometry by mapping and then expanding in terms of expansion functions (e.g., Legendre polynomials). However, this approach usually requires a relatively higher-order expansion in space for accurate representation of the spatial distribution of the flux. To overcome this limitation, a set of consistent spatial expansion functions have been developed in hexagonal geometry in this paper. These functions satisfy the orthogonality condition on hexagonal surfaces and guarantee the conservation of currents after expansion regardless of the truncation order. The latter is an important property for preservation of neutron balance.The new spatial expansion method has been implemented into the coarse mesh radiation transport (COMET) code and tested in both the single assembly and the full core Advanced High Temperature Reactor (AHTR) benchmark problems. The eigenvalues and assembly-averaged and stripe-wise fission density distributions predicted by the new method with 2nd order spatial expansion functions were found to exhibit similar accuracy as those computed by the current mapping method with 4th order spatial expansion functions. This indicates that the angular fluxes in AHTRs can be predominantly represented by a low order expansion if the new spatial expansion functions are used. As a result, to maintain the same accuracy, the computational efficiency of COMET is improved significantly (2–5 times) by using the new spatial expansion method.

Continuous-energy COMET solution to the stylized AHTR benchmark problem

In this paper, the continuous energy coarse mesh transport code COMET was used to provide the whole core transport solution to a set of stylized Advanced High Temperature Reactor (AHTR) benchmark problems. This set includes two full length single assembly and three full core configurations. Solving this type of reactor core is a significant challenge to high and low order neutron transport methods because of its high heterogeneities due to TRISO fuel and burnable absorber particles imbedded in graphite plates with Y-shaped control blades and molten salt as coolant. The eigenvalue and fission density distributions calculated by COMET were compared with the direct Monte Carlo solutions. In the single assembly calculations, it was found that eight-bin expansion in energy and 2nd order expansion in the other phase space variables used by COMET are sufficient for calculation of the eigenvalue and the assembly averaged fission density distribution, while 4th order expansion in space is necessary to obtain the stripe-wise fission density distribution with accuracy comparable to that of the continuous energy Monte Carlo method. It was also found that it is impractical to use Monte Carlo to generate the detailed solutions (stripe-wise fission density distribution) to the full core benchmark problems because of the prohibitive computation resource requirement (estimated to be about 2200 days on an 84-CPU cluster). It took only about 4.5 h on a single CPU for COMET to calculate the detailed solution including the stripe-wise fission density distribution in the entire core. Given that COMET agreed very well with Monte Carlo in terms of assembly averaged and stripe-wise fission density distributions in the single assembly problems as well as in terms of the assembly averaged fission density distribution in the full core problems, it was concluded that the continuous energy COMET can provide the reference solution to the stylized AHTR benchmark problems.

Continuous energy coarse mesh transport (COMET) method

The coarse mesh radiation transport method COMET developed for reactor physics calculations originally treated the phase space variable continuously except for the energy variable. In this work, the method is extended to also treat the energy variable continuously. COMET uses incident flux response expansion coefficients that are generated from local (oarse mesh) fixed source calculations for obtaining the transport solution to whole core problems. In the new method, the energy dependence of the angular flux (spectrum) is factorized into a product of two functions: a strongly varying asymptotic function and a slowly varying (smooth) shape function representing the deviation from the asymptotic function. The method uses spectra from a single assembly or color set calculation as the asymptotic functions and the shape function is expanded by the 0th order B-splines in a set of coarse energy bins. This new factorization allows the use of continuous energy stochastic method for generating the incident flux response expansion coefficients for solving whole-core problems. The new COMET method was tested in a PWR benchmark problem in whole core configurations using UO2 and MOX fuels. It was found that the COMET results agree very well with the corresponding continuous energy Monte Carlo reference solutions. The difference between the COMET and MCNP eigenvalues varied from 29 to 65 pcm, while the mean relative difference in the pin fission density distributions computed by COMET and MCNP was in the range of 0.25%–0.63% for three (namely, controlled, uncontrolled and some-rods-in) core configurations. This difference is less than 2 standard deviations of the MCNP uncertainty. The computational efficiency of the continuous energy COMET was found to be consistent with the multigroup COMET (i.e., more than 4 orders of magnitude faster than MNCP). Based on the benchmark results, it can be concluded that the new continuous energy expansion theory allows COMET to perform high fidelity and highly efficient whole-core transport calculations in heterogeneous reactors.

High-Fidelity Modeling and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies.The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuel meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. Calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.

Monte Carlo solution convergence analysis via COMET in a stylized integral inherently safe LWR benchmark problem

In this paper, the coarse mesh radiation transport (COMET) method is used to study the convergence of the Monte Carlo solution in a stylized benchmark problem based on the beginning of the equilibrium cycle of the Integral Inherently Safe Light Water Reactor concept (I2S-LWR) with silicide U3Si2 fuel. The benchmark problem is comprised of 121 fuel assemblies with radial, top and bottom reflectors. COMET consists of a stochastic local solver (response function generator) for generating a library of incident flux response expansion coefficients for the unique coarse meshes (e.g., all fuel assembly types and reflector/non-fuel blocks) and a deterministic iterative solver for whole-core calculations using this response library. The Monte Carlo method is regarded as the gold standard for benchmarking other numerical transport methods. However, for eigenvalue problems, Monte Carlo results require careful analysis to ensure the convergence of the fission source distribution. It is known that this issue can be of concern, particularly for large eigenvalue problems typical of realistic (i.e., operating) reactors. This became evident when the radially-integrated axial fission density distributions predicted by 50 independent MCNP runs in the I2S-LWR benchmark problem were compared to each other and were found to be significantly different, with the difference ranging from −9% to +9%. However, the average of the fission density distributions from these 50 runs agreed very well with that predicted by COMET, with a relative difference varying from −0.18% to 0.13%, being within one standard deviation of the average. This agreement confirms that the averaged Monte Carlo solution is converged.

Incorporation of interface current method based on 2D CP approach in VISWAM code system for hexagonal geometry

The design and operation of nuclear power reactors has seen a significant improvement since the operation of the first reactor in 1950s. Most of the operating reactors around the world today belong to the Generation II or III design versions. Generation IV reactors which are being evolved, have improved thermal efficiencies. Very High Temperature Reactor (VHTR) is one of the reactor concepts being considered under the scope of Generation IV reactors. The development of next generation reactors requires advanced reactor physics methods with greater accuracy and fidelity than traditional diffusion methods. A comprehensive code system VISWAM for physics analysis of current and future power reactors is being developed (Jagannathan et al., 2013a). The lattice analysis method initially incorporated in VISWAM code is based on a combination of 1D multigroup transport and a 2D few group diffusion theory. This methodology is computational efficient. However for analyzing problems with complexities like Gd or control cells some limitations were noted (Jagannathan et al., 2013b). We have applied the 2D collision probability (CP) method at single pincell level and linked the cells using interface currents with double P2 (DP2) expansion of angular flux at cell boundaries. We have implemented this method in VISWAM to analyze hexagonal fuel assembly (FA). The use of DP2 expansion for hexagonal geometry is not reported in the literature to the best knowledge of the authors. The FA cell in hexagonal geometry with irregular lattice structure at boundaries is modeled exactly. Two benchmark problems, a heterogeneous benchmark problem that is typical of a high temperature reactor (HTTR) proposed by Zhang et al. (2011) and VVER – 1000 OECD computational benchmark (NEA/NSC/DOC, 2002), are studied using VISWAM. The present paper describes the advanced methodology incorporated in VISWAM and the comparison of results of benchmark analysis with published results. The k∞ for HTTR pincell/FA calculation matches within 0.01%/0.18% respectively with the benchmark results. The RMS error in the fission density distribution is seen as 0.13%. For VVER – 1000 OECD benchmark, the k∞ compares well with benchmark results. The % RMS deviation in fission densities using DP2 is seen to be 0.462%.

A novel extension of chord length sampling method for TRISO-type fueled reactor applications

A new chord length distribution model is proposed to characterize the stochastic distribution of TRISO fuel particles in nuclear reactor systems and is used in the chord length sampling (CLS) method to analyze the neutronic behavior of TRISO fuel systems. In this model, the coating layers of fuel particles are homogenized with the background matrix region. The probability density function (PDF) of the chord length between fuel kernels, instead of fuel particles, is developed and is used in the CLS method. We first apply the new CLS model to solving one-group eigenvalue problems in a simplified 3-D stochastic medium system. Good accuracy is obtained in predicting the multiplication factor and fission density distribution. The relative differences are within 1.0% for both the multiplication factor and the total fission density in all the studied scenarios. We then apply the new CLS model to analyzing three realistic reactor designs: two Very High Temperature Gas-cooled Reactor unit cells and one fuel pin unit cell of an innovative light water reactor design with accident tolerant fuel. Infinite multiplication factor and intra-Dancoff factor are evaluated for the three unit cells respectively. Compared with the reference results, predictions from CLS simulations show a relative difference of less than 0.26% for infinite multiplication factors and less than 1.0% for intra-Dancoff factors in all the studied cases. Meanwhile, a significant improvement in the computational efficiency has been observed for the CLS new model compared with the old model, with a speedup of at least 1.70. The CLS method with the new chord length distribution model is verified to be an efficient Monte Carlo method, superior over the old model and without sacrificing the accuracy.

A Stylized Three-Dimensional Pressurized Water Reactor Benchmark Problem with UO2 and MOX Fuel

This paper fully describes a whole-core benchmark problem based on a small two-loop pressurized water reactor composed of UO2 and MOX fuel assemblies. The specification includes heterogeneity at both the assembly and the core levels. The geometry and material compositions are fully described, and multigroup material cross-section libraries are provided in the two-group, four-group, and eight-group formats. Detailed Monte Carlo reference solutions including core eigenvalue, assembly-averaged fission distribution, and selected fuel pin fission density distributions are presented for benchmarking diffusion and transport methods. Three different core configurations are presented in this paper, namely, all-rods-out, all-rods-in, and some-rods-in.

A heterogeneous coarse mesh radiation transport method for neutronic analysis of prismatic reactors

A new whole-core transport method is described for 3-D hexagonal geometry. This is an extension of a stochastic–deterministic hybrid method which has previously been shown highly accurate and efficient for eigenvalue problems. Via Monte Carlo, it determines the solution to the transport equation in sub-regions of reactor cores, such as individual fuel elements or sections thereof, and uses those solutions to compose a library of response expansion coefficients. The information acquired allows the deterministic solution procedure to arrive at the whole core solution for the eigenvalue and the explicit fuel pin fission density distribution more quickly than other transport methods. Because it solves the transport equation stochastically, complicated geometry may be modeled exactly and therefore heterogeneity even at the most detailed level does not challenge the method.In this paper, the method is evaluated using comparisons with full core Monte Carlo reference solutions of benchmark problems based on gas-cooled, graphite-moderated reactor core designs. Solutions are given for core eigenvalue problems and the calculation of fuel pin fission densities throughout the core. Using a single processor, results are found in minutes for small cores, and in no more than a few hours for a realistically large core; attempts to optimize the computational speed by parallel computing or advanced acceleration schemes are left for future work. Typical eigenvalues calculated by the method differ from reference solutions by less than 0.1%, and pin fission density calculations have average accuracy of well within 1%, even for unrealistically challenging core configuration problems. This new method enables the accurate determination of core eigenvalues and flux shapes in hexagonal cores with efficiency far exceeding that of other transport methods.

Study of Using Zirconium as a Reflector for Light Water Reactors

Zirconium is studied in this paper as a reflector for light water reactors. An exploratory analysis of using zirconium as a reflector for two simple reactor core models was carried out. The study showed that use of zirconium as a reflector has a valuable impact on the core reactivity. The study also showed that zirconium-water reflector is more effective than water reflector or stainless steel-water reflector. A typical Westinghouse 1150-MW(electric) pressurized water reactor was simulated using the Monte Carlo code MCNP5 as a case study. The simulation was carried out at the beginning of the core cycle of three batch cores with 235U enrichments of 2.25, 2.8, and 3.3 wt%. The simulation showed that use of Zircaloy-4 reflector between the fuel assemblies and the core barrel adds a positive reactivity Δkeff of 0.00686, while use of stainless steel reflector adds a positive reactivity Δkeff of 0.0037.Use of Zircaloy-4 reflector increases the relative power density in the peripheral assemblies by ˜38%. The power peaking factor is shifted from the center toward the periphery, and the assembly power peaking factor is reduced by ˜13%. The use of Zircaloy-4 reflector with this increase of the reactivity of the peripheral assemblies increases the fast neutron current (E > 0.5 MeV) that reaches the reactor pressure vessel (RPV) by 70%, while the use of stainless steel reflector reduces it by 44%.Adjusting the 235U enrichment in the peripheral assemblies batch to compensate for the excess reactivity caused by using Zircaloy-4 reflector reduces the 235U enrichment by 8.5% in this batch. This means a reduction of 3.35% of the core 235U average enrichment can be achieved by the use of Zircaloy-4 reflector. This reduction in the 235U enrichment reduces the increase of the fast neutron current that reaches the RPV to 23%. In this case, increasing the water gap between the core barrel and the RPV by 3 cm reduces the fast neutron current that reaches the RPV to 95% of that of the basic case. The use of Zircaloy-4 reflector has a good effect on flattening the fission density distribution in the peripheral assemblies batch both before and after reducing 235U enrichment.

A coarse mesh radiation transport method for 2-D hexagonal geometry

In this paper, a whole-core stochastic–deterministic hybrid coarse mesh transport method is extended to 2-D hexagonal geometry. This method may be used to calculate the eigenvalue and explicit pin fission density profile of hexagonal reactor cores. It models the exact detail within complex heterogeneous cores without homogenizing regions or materials, and neither block-level nor core-level asymmetry poses any limitations to the method. It solves eigenvalue problems by first splitting the core into a set of coarse meshes, and then using Monte Carlo methods to create a library of response expansion coefficients, found by expanding the angular current in phase-space distribution using a set of polynomials orthogonal on the angular half-space defined by mesh boundaries. The coarse meshes are coupled by the angular current at their interfaces. A deterministic sweeping procedure is then used to iteratively construct the solution.The method is evaluated using benchmark problems based on a gas-cooled, graphite-moderated high temperature reactor. The method quickly solves problems to any level of detail desired by the user. In this paper, it is used to explicitly calculate the fission density of individual fuel pins and determine the reactivity worth of individual control rods. In every case, results for the core multiplication factor and pin fission density distribution are found within several minutes. Results are highly accurate when compared to direct Monte Carlo reference solutions; errors in the eigenvalue calculations are on the order of 0.02%, and errors in the pin fission density average less than 0.1%.

Simplified two and three dimensional HTTR benchmark problems

To assess the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of whole core configurations. In this paper we have created two and three dimensional numerical benchmark problems typical of high temperature gas cooled prismatic cores. Additionally, a single cell and single block benchmark problems are also included. These problems were derived from the HTTR start-up experiment. Since the primary utility of the benchmark problems is in code-to-code verification, minor details regarding geometry and material specification of the original experiment have been simplified while retaining the heterogeneity and the major physics properties of the core from a neutronics viewpoint. A six-group material (macroscopic) cross section library has been generated for the benchmark problems using the lattice depletion code HELIOS. Using this library, Monte Carlo solutions are presented for three configurations (all-rods-in, partially-controlled and all-rods-out) for both the 2D and 3D problems. These solutions include the core eigenvalues, the block (assembly) averaged fission densities, local peaking factors, the absorption densities in the burnable poison and control rods, and pin fission density distribution for selected blocks. Also included are the solutions for the single cell and single block problems.

The experimental determination and evaluation of the three-dimensional fission density distribution of the IPEN/MB-01 research reactor facility for the IRPhE project

Experiments for the determination of the three-dimensional fission density distribution were successfully carried out at the IPEN/MB-01 research reactor facility. The experiment is of very good quality for utilization in the benchmark of computer codes and related nuclear data library commonly used for the calculation of fission density distribution in reactor cores. The complete experimental data set comprises a very massive set of measured fission density distribution. The experiment was evaluated and included in the IRPhE handbook. Two calculation methodologies were employed for the theoretical analysis of the proposed benchmark; one stochastic (MCNP-5) and the other one deterministic (NJOY/AMPX-II/TORT); both utilizing ENDF/B-VII.0 as the source of nuclear data. The theory/experiment comparison reveals in general a good agreement for both methodologies. However, MCNP-5 results are in a better shape than those of TORT. TORT underestimates the relative power distribution mainly in the axial upper part of the fuel rod. This effect is mostly credited to the over-prediction of the control bank worth.