Coolant Loop Research Articles

Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor.Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for accident mitigation.

Non-dimensional analysis of the transient flow rate and the coolant temperature in a reactor coolant system under different pump failures

ABSTRACTThe mathematical models are developed to solve the non-dimensional transient flow rates in two loops and a reactor core under different power failures of reactor coolant pumps. Comparison of the experimental results of the Qinshan Nuclear Power Plant and the test results of the nuclear ship reactor in Japan under one pump power failure shows an excellent agreement. The non-dimensional flow rates are determined by the established non-dimensional parameters λ, ϵ, and α. Under the sequential power failure of two pumps, the non-dimensional flow rates are determined by the established λ, ϵ, α, and ΔT parameters. λ, ϵ, α, and ΔT are four important non-dimensional parameters in the prediction of flow transients. λ indicates the resistance coefficient ratio of the single loop to the reactor core, ϵ indicates the fluid inertia ratio of the reactor core to the single loop, α indicates the ratio of the initial kinetic energy of the single loop coolant fluid to the effective initial kinetic energy of the reactor coolant pump, and ΔT means the non-dimensional time interval of the sequential power failure of two pumps. The effects of λ, ϵ, α, and ΔT on the non-dimensional flow rates and the temperature change are investigated.

A study on heat transfer enhancement in the radial direction of gas flow for thermoelectric power generation

The performance of thermoelectric generation (TEG) systems is significantly dependent on the hot side temperature of thermoelectric legs and the temperature difference between the hot side and cold side of the legs. To keep the TEG module working at an optimal condition, a high heat flux over 10W/cm2 through the TEG needs to be maintained. Due to the low heat transfer coefficient of the gas flow to the exhaust pipe wall surface, typically at a high temperature ranging from 400°C to 800°C, the actual heat flux into the TEG heat exchanger is limited significantly, resulting in relatively low efficiency of the TEG conversion. In the present study, an effective solution for enhancing the heat transfer of gas flow in the radial direction to the TEG is proposed by means of immersing high temperature heat pipes perpendicularly into the exhaust flow. Similarly, conventional heat pipes are radially inserted into a concentric coolant jacket in order to enhance heat transfer performance at the cold side of TEG modules. Overall, the TEG assembly is configured as a compact and scalable heat pipe heat exchanger. The simulation results show that the hot side temperature of the TEG can reach and be maintained as high as 300°C while the cold side temperature of the TEG can be maintained at approximately 85°C for a normal engine coolant loop. The results also show that the closer to the heat source in the pipeline the TEG system is located, the better the power generation that is expected. Moreover, better thermoelectric generation can be expected at a higher engine speed. By installing the TEG heat exchanger between catalytic converter and muffler, the best power output in the thermoelectric heat exchanger can be achieved at 450W and 5000rpm. If the TEG heat exchanger is adjacent to the outlet of a catalytic converter, the best-simulated performance at 6000rpm is 705W for a single sub pipeline. Therefore, a total power generation of 1410W is achievable since the existing exhaust pipe is a dual pipeline system.

Mechanical analysis of flying robot for nuclear safety and security control by radiological monitoring

The flying robot is investigated for the nuclear accident and security treatment. Several mechanics are introduced for the movement of the drone. The optimized motion of the drone should cover all areas of Nuclear Power Plants (NPPs) over the site where the circular and surmounting motions are needed with traverse of zigzag shapes. There is the Yaw motion in the circular moving and the Pitch motion in the climbing and downing against reactor facility. The fallout is calculated from the radiation concentration in the breaking part of the NPPs where the radioactive material leaks from the containment, coolant loop, plant facility and so on. The dose equivalents are obtained where the values are changeable following the random values of the y value, average wind speed, and dispersed concentration in the detection position. The simulation of new positions of x, y, and z are normalized from 0.0 to 1.0. The mechanics of flying robot produces the multidisciplinary converged technology incorporated with the aerial radiation monitoring information.

Prediction of the Nuclear Fuel Rod Abrasion in the Probability Sense

The paper deals with the upper and lower limits estimation of the friction work and fretting wear in the contact of nuclear fuel rods with fuel assembly (FA) spacer grid cells. The friction work is deciding factor for the prediction of the fuel rod cladding abrasion caused by FA vibration. Design and operational parameters of the FA components are understood as random variables defined by mean values and standard deviations. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of the friction work and fretting wear in particular contact surfaces between the fuel rod cladding and some of spacer grid cells. The fuel assembly vibration is excited by pressure pulsations of the cooling liquid generated by main circulation pumps in the coolant loops of the NPP primary circuit. The method is applied for hexagonal type nuclear fuel assembly in the VVER type reactors.

Qualitative and Quantitative Evaluation of Coupling Approaches for Coupling of RELAP5 and LabVIEW

In the study and design of a nuclear power plant, extensive system modeling is necessary to determine how the reactor will perform in any given situation, not only in the normal performance of the reactor, but also in transients including unanticipated transients without scram and hypothetical accidents. One type of nuclear power plant under study is the hybrid energy system, which uses nuclear power to generate both electricity and heat for facilities. Obviously, the second steam cycle in the nuclear power plant requires several design updates and experiments. Unfortunately, the current versions of the Reactor Excursion and Leak Analysis Program (RELAP) do not allow online data streams from experimental facilities to the computational model of the secondary steam loop. Therefore, this study develops a coupling between RELAP5 and Laboratory Virtual Instrument Engineering Workbench (LabVIEW) to model primary and secondary coolant loops. In this way, the LabVIEW model can easily be connected to an experimental apparatus to provide an online data stream and the online transient behavior of an entire nuclear power plant system. This study shows two different coupling approaches and makes qualitative and quantitative comparisons between these approaches.This paper demonstrates the results of different couplings between the primary and secondary systems of a typical pressurized water reactor (PWR). The primary loop model is a four-loop PWR. The model has been executed with steady state and transients (in this case, a loss-of-coolant accident). The results of both coupling methods have been compared with the typical RELAP5 results.

Integrated condition monitoring for plant-wide prognostics

The use of Condition Monitoring (CM) in the power industry is well established, with traditional manual analysis increasingly giving way to intelligent, automated analysis. Further, there is a growing interest in the extension of such CM systems to derive prognostic information for asset management. It is common however for CM applications to be retrofitted to existing equipment, to monitor a particular component or subsystem, resulting in several isolated CM systems attached to a larger system from which it is not trivial to extract plant-wide prognostic information. The extension of CM applications to deliver prognostic information is an area of current research in the nuclear industry as operators seek to maximise the operational life of ageing plants. The interdependence of multiple safety critical systems, each of which is monitored individually, is a critical challenge in determining the health of a life-limiting component which may be part of a chain of monitored sub-systems. The generation of new information about plant-wide state, based on subsystem prognostics and monitoring, can help inform operation and maintenance. This paper describes two CM systems used to monitor key systems on the UK fleet of Advanced Gas-Cooled Reactors, at different ends of the coolant cycle in the reactor. The first system monitors vibrations of the gas circulators used to pump coolant into the core, while the second system monitors the thermal power of the fuel channels inside the core. The systems make diagnoses of anomalies in their respective data-sets, however a fault in the gas circulator would directly affect the cooling of the core, potentially affecting the fuel channels further along the coolant loop. The paper discusses how closer integration of the outputs of each system may better inform the prognostic information about the plant, by identifying related anomalies and providing corroborating information to better inform prognostic estimates.

European Project “Supercritical Water Reactor–Fuel Qualification Test”: Overview, Results, Lessons Learned, and Future Outlook

The supercritical water reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). One of the major challenges in the development of a SCWR is to develop materials for the fuel and core structures that will be sufficiently corrosion resistant to withstand supercritical water conditions and to gain thermal-hydraulic experimental data that could be used for further improvement of heat transfer predictions in the supercritical region by numerical codes. Previously, core, reactor, and plant design concepts of the European high-performance light water reactor (HPLWR) have been worked out in great detail. As the next step, it has been proposed to carry out a fuel qualification test (FQT) of a small-scale fuel assembly in a research reactor under typical prototype conditions. Design and licensing of an experimental facility for the FQT, including the small-scale fuel assembly, the required coolant loop with supercritical water, and safety and auxiliary systems, was the scope of the recently concluded project “Supercritical Water Reactor–Fuel Qualification Test” (SCWR-FQT) described here. This project was a collaborative project cofunded by the European Commission, which took advantage of a Chinese–European collaboration, in which China offered an electrically heated out-of-pile loop for testing of fuel bundles. The design of the facility, especially of the test section with the fuel assembly, and the most important results of steady-state and safety analyses are presented. Material test results of the stainless steels considered for the fuel cladding are briefly summarized. Finally, important outcomes and lessons learned in the “Education and Training” and “Management” work packages are presented.

Numerical analysis on inlet and outlet sections of a test fuel assembly for a Supercritical Water Reactor

The Supercritical Water Reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). One of the major challenges in the development of a SCWR is to develop materials for the fuel and core structures that will be sufficiently corrosion-resistant to withstand supercritical water conditions. Previously, core, reactor and plant design concept of the European High Performance Light Water Reactor (HPLWR) have been worked out in substantial detail. As the next step, it has been proposed to carry out a fuel qualification test of a small scale fuel assembly in a research reactor under typical prototype conditions. Therefore design and licensing of an experimental facility for the fuel qualification test, including the small scale fuel assembly with four fuel rods, the required coolant loop with supercritical water and safety and auxiliary systems, is the scope of the project “Supercritical Water Reactor—Fuel Qualification Test” (SCWR-FQT). This project is a collaborative project co-funded by the European Commission, which takes advantage of a Chinese—European collaboration.As a sub-task of the SCWR-FQT project, the geometry of inlet and outlet sections of the fuel assembly has to be investigated and optimized according to thermohydraulic considerations such as expected stable and uniform inflow pattern and uniform outflow temperature field conditions. To accomplish this task three dimensional CFD analysis has been performed. During the analysis two main problems were identified. On the one hand, generation of a huge eddy was discovered in the so-called flow direction changing chamber (upstream to the inlet section) which has to be avoided or at least reduced in order to ensure stable inflow conditions. On the other hand, low velocity regions were identified directly downstream to the inlet which could cause cooling deficiency at the beginning of the heated part of fuel rods. Both problems have been solved by geometrical optimization of the inlet section. The outlet section has been analysed and it graded suitable as it is.

Simulation of fuel dispersion in the MYRRHA-FASTEF primary coolant with CFD and SIMMER-IV

The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4.Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant and descendant flows in the core channels suppresses the rise in the core temperature instead of the primary coolant loop. However, the test result suggests that the present DHRS has insufficient performance to remove the decay heat in the case of siphon break. Repeated tests have shown that the results of the water test have sufficiently high reproducibility in terms of the flow rate of the primary loop to be used for the verification of numerical codes.

Preliminary tests of particle image velocimetry for the upper plenum of a scaled model of a very high temperature gas cooled reactor

The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. During the design process multiple studies have been performed to develop safety codes for the reactor. Two major accidents of interest are the Pressurized Conduction Cooldown (PCC), and the Depressurized Conduction Cooldown (DCC) scenario. Both involve loss of forced coolant to the core, except the latter involves a pressure loss in the main coolant loop. During normal operation a circulator pumps the coolant into the upper plenum and down through the core, but following a loss of forced coolant the natural convection causes the flow to reverse to go through the core into the upper plenum. Computer codes may be developed to simulate the phenomenon that occurs in a PCC or DCC scenario, but benchmark data is needed to validate the simulations; previously there were no experimental test facilities to provide this. This study will cover the design, construction, and preliminary testing of a 1/16th scaled model of a VHTR that uses Particle Image Velocimetry (PIV) for flow visualization in the upper plenum. Three tests were run for a partially heated core at statistically steady state, and PIV was used to generate the velocity field of three naturally convective adjacent jets; the turbulent mixing of the jets was observed. After performing a sensitivity analysis the flow rate of a single pipe was extracted from the PIV flow field, and compared with an ultrasonic flowmeter and analytic flow rate. All the values lied within the uncertainty ranges, validating the test results.

Modeling primary and secondary coolant of a nuclear power plant system with a unique framework (MCUF)

Within the study and design of a nuclear power plant extensive system modeling is necessary to determine how the reactor is going to perform in any given situation, not only in the normal performance of the reactor but also transients including anticipated transients without scram (ATWS) and hypothetical accidents. Primary and other loops in multiple coolant loop systems in nuclear power plants can be modeled in nuclear system codes, such as Reactor Excursion and Leak Analysis Program (RELAP5), COBRA-TRAC, TRACE, and ATHLET, for the interaction and feedback effects between the coolant loops. In the recent years, new components, such as turbine, coolant pump, advanced compact heat exchangers, valves, sensors, instrumentation and control systems, are being designed to use in the coolant systems. However, current system codes are not capable of integrating the computational model of the whole nuclear power plant system to an experimental apparatus. Therefore, this article proposes a new coupling between a system code and Laboratory Virtual Instrument Engineering Workbench (LABVIEW) in a unique framework (called as Modeling of Coolant Systems with a Unique Framework -MCUF-) that allows integrating an experimental apparatus or a facility to a whole power plant model. MCUF allows online or interactive data exchange by using the powerful and flexible tools of LABVIEW. This article demonstrates how coupling between the primary and secondary coolant system of a typical Pressurized Water Reactor (PWR) can be performed by using MCUF. The primary and secondary sides of the PWR are modeled with RELAP5 and LabVIEW computer simulators respectively. The coupling between RELAP5 and LabVIEW has been executed with steady state and transients, in this case a loss of coolant accident (LOCA) for a four loop PWR. The results of the coupling have been compared with the typical RELAP5 results which significantly depend on “hard coded” data for second coolant loop. Code-to-code benchmark results demonstrate that this unique coupling framework has a good agreement with the RELAP5 code results with “hard coded” data for the secondary coolant system.

The effect of corrosion product CrF3 on thermo-physical properties of FLiNaK

FLiNaK (LiF–NaF–KF: 46.5–11.5–42 mol %) is a promising candidate as the secondary loop coolant in molten salt reactor. The thermo-physical properties of pure FLiNaK and FLiNaK containing up to 6000 ppm (equivalent to mg/kg) corrosion product CrF3 were measured. The results indicate that the effects of CrF3 on melting point, enthalpy, specific heat capacity, density and thermal diffusivity of FLiNaK in liquid state are negligible within the allowable error range, meanwhile the change of thermal diffusivity is significant for FLiNaK in solid state. This work provides fundamental knowledge for the thermo-physical properties of coolant in molten salt reactor.

Parametric Study of Time-Dependent Corrosion Product Activity due to56Mn,58Co, and60Co in the Primary Coolant Circuit of a Typical Pressurized Water Reactor

Results of a detailed study, based on the parametric analysis of activated corrosion products, in primary coolant of a typical pressurized water reactor (PWR) are presented. The parameters influencing time dependent buildup of corrosion product activity (CPA) in primary coolant loop of PWR were identified. The computer program CPAIR was used to accommodate for time dependent corrosion rates. The behaviors of56Mn,58Co, and60Co were studied over the reactor operational time. During the course of normal operation of reactor, the CPA is dominated by56Mn, while58Co and60Co are the predominant radionuclides after reactor shutdown. Parametric study suggests that the total CPA is most sensitive to ion-exchanger removal rates. For a removal rate of 300 cm3-s−1, the specific activity due to56Mn has the maximum value of 3.552 × 104 Bq-m−3after 1,000 hours of reactor operation. This value decreases drastically to 8.325 × 103 Bq-m−3at removal rate of 900 cm3-s−1. Additionally, CPA due to56Mn,58Co, and60Co shows strong dependence on removal rates from the core material surfaces. Variations in the values of radionuclide removal rates from piping surface and radionuclide removal rate from deposition on pipes showed only very small effects on CPA buildup.

Smart cooling system of the double loop coolant structure with engine thermal management modeling

A thermal management analysis for the two double loop coolant circuit cases, which used separated cooling structures for the head and block, was performed in the current study. With the conventional configuration, the problem of over cooling problem in the specific components arose, because the cooling system was constrained by the thermal limitations of specific components requiring high cooling loads. An integrated engine thermal management model was developed to observe features of the proposed structure, which was validated with experimental data, and a semi-iteration process was conducted using a 1D high-frequency engine model. To maximize the advantages of the proposed structure, two electric devices were applied to the ETM model to facilitate “smart cooling.” The current analysis indicated that the 30% flow rate of the basic system satisfied the thermal limitations of the target engine under half load conditions; hence, the thermal efficiency increased because of the decreased auxiliary power, friction and heat transfer loss. In addition, it was noted that the thermal deviation in an engine and warm-up time diminished because of the structural changes and optimal coolant control.

DEVELOPMENT AND VALIDATION OF THE AEROSOL TRANSPORT MODULE GAMMA-FP FOR EVALUATING RADIOACTIVE FISSION PRODUCT SOURCE TERMS IN A VHTR

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

Modeling a Freezable Water-Based Heat Exchanger for Use in Spacecraft Thermal Control

A spacecraft thermal control system typically uses water coolant loops to keep the cabin comfortable. However, the relatively high freeze point of water limits its use, because freezing water within the loop may damage the thermal control system. This paper describes a novel freeze-tolerant heat exchanger and the thermal model developed to correlate data from experiments. The radial transport of heat within the heat exchanger was modeled as a thermal circuit. The heat flows from the water into the thermally conductive shell and fins. One flow channel within the heat exchanger was lined with insulation to keep it free of ice. When the heat sink temperature is below the freeze point of water, a layer of ice builds up in the other flow channels to greatly increase the thermal resistance to heat transport, since the thermal conductivity of ice is only about th that of Al-6061. In addition, the wetted surface area is reduced by a factor of six when only the insulated channel can still flow water (all of the other channels are frozen). Together, these two effects turn down the rate of heat rejection in proportion to the net thermal load and the external heat sink environment encountered.

Safety improvement of pool type FBR against loss of flow accident

Among the various loss of flow situations possible in a fast reactor, rupture of coolant pipe is very important due to the rapid reduction in coolant flow resulted under this condition. This event dictates the number of coolant loops in primary circuit. Safety demonstration of two loop medium size fast reactor design is a challenging task due to large reduction in core flow caused during this event. This difficulty can be overcome by increasing number of coolant pipes. The pros and cons of this option are investigated in this paper. Thermal hydraulic benefits of increasing the number of coolant pipes are presented through plant dynamic studies. The main difficulty of this option lies in layout of pipes due to the space constraints. An option to overcome this is by reducing the gird plate size. The associated hydraulic issues have been investigated to ensure core safety. Parametric studies covering pipe size, hydraulic resistance of various parts and cavitation characteristics of pumps have also been performed to identity cliff edge effects.

Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the SN part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high SN/PN parameters. Detailed voxelization (homogenization) process in Denovo together with high SN/PN parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR facility model which included only reactor core and adjacent concrete structures with the steam generator and the pump. Hybrid FW-CADIS methodology appears to be very effective for shielding calculations of complicated MC geometries. It was demonstrated that definition of air as a global adjoint source gave lower uncertainty of photon dose rates. Activation of the primary loop coolant, which becomes additional gamma source in operating reactor, was considered. Compared with analog MC simulations, the hybrid methodology drastically improved Monaco distributions in quality as well in space coverage.The SCALE6.0 hybrid stochastic methodology was thoroughly investigated and its versatile ability for deep penetration models was proved. A careful selection of parameters can relax high CPU time and hardware requirements.