Hesham Nasif

Development of CAD-MCNP Interface Program GEOMIT and Its Applicability for ITER Neutronics Design Calculations

GEOMIT is a computer-aided design (CAD)/MCNP conversion interface code. It was developed to automatically generate Monte Carlo geometrical data from CAD data due to the difference in the representation scheme. GEOMIT is capable of importing as well as exporting different CAD formats. GEOMIT has the capability to produce solid cells as well as void cells without using the complement operator. While loading the CAD shapes (solids), each shape is assigned a material number and density according to its color on the original CAD data. A shape fixing process has been applied to cure the errors in the CAD data. Vertex location correctness is evaluated first, and then a removal of free edges and removal of small faces processes. A binary space portioning tree technique is used to automatically split complicated solids into simpler cells to avoid excessively complicated cells to allow MCNP to run faster. MCNP surfaces are subjected to an automatic reduction before creating the model. CAD data of the ITER benchmark model have been converted successfully to MCNP geometrical input. MCNP input model validations have been carried out by checking lost particles and comparing volumes calculated by MCNP to those of the original CAD data. Different test cases have been evaluated for ITER, including blanket first wall heat loading calculations, surface fluxes, and volume fluxes at different divertor regions as well as toroidal field coil heating.

Wavelet Integrated System to Calculate Radionuclide Release from a Repository in Fractured Media

Abstract Radionuclides released from a vitrified waste package after overpack failure spread into the buffer material surrounding the waste package, then migrate through different pathways into the water-bearing fracture in the rock surrounding the high-level radioactive waste repository, and transport through the faults to the biosphere. The buffer material has low permeability and the solute is transported through the engineered barrier system by diffusion only. In the water-bearing fracture, the problem is of the convection diffusion type with highly varying parameters from one medium to the other due to the variability in length, transmissivity, and other transport-relevant properties of the transport paths. This complex geometry is modeled using the wavelet Galerkin approach. The Wavelet Integrated Repository System (WIRS) wavelet-based system is an integrated tool to calculate the transport of single or radionuclide chains in both near and far fields of the repository system. The model, which is a very coarsely discretized wavelet based, is devised to be very fast since the scaling functions, which are used as a basis function, are compactly supported. Only finite numbers of the connection coefficients are nonzero, and the resultant matrix has a block diagonal structure that can be inverted easily. One of the main problems encountered in solving the model for the radionuclide transport in the geospheric media is the treatment of the boundary and interface conditions. In order to maintain the integrity of the system, the boundaries of the wavelet series are shifted until the end is independent of any expansion coefficients of the scaling function that affect the solution within the real boundaries. WIRS agreed well with models using a very detailed discretization. Accuracy is gained with the proper selection of wavelet-dilation orders pair. WIRS has been applied to the Japanese high-level radioactive waste repository concept where the migration is through different barriers and pathways. Single and decay chain radionuclide release calculations have shown the capability of WIRS to handle different situations rapidly and easily.

Development of CAD-MCNP Interface Program “GEOMIT”

GEOMIT is a CAD/MCNP conversion interface code. It is developed to automatically generate Monte Carlo geometrical data from CAD data due to the difference in the representation scheme. GEOMIT is capable of importing different CAD format as well as exporting different CAD format. GEOMIT has a capability to produce solid cells as well as void cells without using complement operator. While loading the CAD shapes (Solids), each shape is assigning material number and density according to its color on the original CAD data. Shape fixing process has been applied to cure the errors in the CAD data. Vertices location correctness is evaluated first, and then a removal of free edges and removal of small faces processes. Binary Space Portioning (BSP) tree technique is used to automatically split complicated solids into simpler cells to avoid excessive complicated cells for MCNP to run faster. MCNP surfaces are subjected to an automatic reduction before creating the model.Copyright

Risk-Informed and Defense-in-Depth Oriented Plant Design Approach

Defense-in-Depth is the basis of safety design of nuclear plants, and refined and strengthened year by year. Nowadays, the importance of it has been further highlighted triggered by the accident at Fukushima Daiichi Nuclear Power Station.Accidents described previously have been shown it could not care enough to uncertainty related to design, construction, maintenance and operation has exposed. It’s the lessons learned of that is to say to reduce uncertainty of design is the use of risk informed is essential. Therefore, the establishment of the design approach that uses risk informed consideration of the Defense-in-Depth is an important theme.Defense-in-Depth is a measure which prevents the increase of the event frequency and core damage in consideration of the degree of safety margins, redundancy, diversity and consideration of radiation safety due to core damage and security.The plant designers and utilities have made efforts to ensure safety utilized conventional design technique, which means deterministic, and risk information, in order to incorporate Defense-in-Depth concepts.To consider Defense-in-Depth in the design phase, various requirements should be taken into consideration. It has been coming to be able to perform more rational and quantitative judgment by utilization of risk information.In that case, while the designers of various fields will work in cooperation to ensure safety, if there are common utilization schemes for risk information among designers, more efficient and rational design works can be advanced in consideration of Defense-in-Depth. However, conventionally, it is hard to say that such schemes have functioned in order that designers may advance design works in collaboration.This research intends to generate the schemes which advance design works in sharing the same risk information databases, which mean various risk indications etc), in other words, the schemes which will become lingua franca for utilizing risk information, in case of nuclear plants designs.Copyright

Optimization of Dead-Layer Thickness for a HPGe Detector Using UCODE-MCNP Codes

The use of modeling programs to predict the response of HPGe detectors is increasing in importance due to the extensive laboratory work, both in term of source preparations and measuring time. MCNP code is a powerful and useful tool for the simulation of Ge-detector efficiency calibration. The experimental efficiency data and MCNP calculations based only on the known physical measurements of the HPGe crystal do not agree well in some detectors. Detector construction materials and surface dead layers must be well specified. The dead layer of Ge detector is one of the most important factors that affect the calculations. In addition, and if provided by the manufacturer, the dead layer may changes with time. Consequently, it is necessary to optimize the thickness of the detector’s dead layer in order to obtain more accurate results for the efficiency of the detector using Monte Carlo calculations. Our approach consists of employing hybrid UCODE-MCNP codes to optimize the dead layer of the Ge-crystal aiming at decreasing discrepancies between experimental and simulated data of the Ge detector efficiency. UCODE has two attributes that are not jointly available in other inverse models: (1) the ability to work with any mathematically based model or pre- or post processor with ASCII or text only input and output files, and (2) the inclusion of more informative statistics.Copyright

Development of the CAD/MCNP Automatic Conversion Code GEOMIT

GEOMIT is the CAD/MCNP conversion interface code. The old version of GEOMIT had a limited capability from CAD model handling point of view. It is developed to automatically generate Monte Carlo geometrical data from CAD data due to the difference in the representation scheme. GEOMIT is capable of importing different CAD format as well as exporting different CAD format. GEOMIT has a capability to produce solid cells as well as void cells without using complement operator. While loading the CAD shapes (Solids), each shape is assigning material number and density according to its color. Shape fixing process is been applied to cure the errors in the CAD data. Vertices location correctness is evaluated first, then a removal of free edges and removal of small faces processes. Binary Space Portioning (BSP) tree technique is used to automatically split complicated solids into simpler cells to avoid excessive complicated cells for MCNP to run faster. MCNP surfaces are subjected to an automatic reduction before creating the model. CAD data of International Thermonuclear Experimental Reactor (ITER) benchmark model has been converted successfully to MCNP geometrical input. The first wall heat loading calculations agree very well with other countries results.Copyright

A New Developed Interface for CAD/MCNP Data Conversion

In a complex and huge system as in ITER fusion reactor, the creation of the geometrical input data of Monte Carlo (MC) codes such as MCNP is a highly exhausting task. Accordingly, it is a general approach to shift the geometric modeling into a computer aided design (CAD) system and to use an interface, which performs the exchange of CAD data into a representation appropriate for MC code. We have developed efficient algorithms and computer code, which are used to convert Parasolid format CAD files including solid and void data into MCNP input data. The CAD-MCNP conversion processes include creating surface equations; determining surface senses; constructing cell geometry and creating MCNP input file. This paper describes the basic algorithms used for the CAD/MCNP interface along with some applications for different geometries.Copyright

Integrated Multi Path Model to Calculate Radionuclide Release From a Repository Using Wavelet Galerkin Method

This work represents a WIRS code developed using wavelet Galerkin method to solve radionuclide transport model in near field and far field of a repository for high-level radioactive waste. After overpack failure, radionuclides diffuse through the bentonite buffer material to the water bearing fracture around the repository transport horizontally through this geosphere then transport vertically through the major water conducting fault (MWCF) reach the biosphere. The radionuclides transport barriers considered in this model are engineered barrier system (EBS), geosphere, and MWCF. Hydraulic conductivity of the bentonite is more than three orders of magnitude smaller than that of the surrounding host rock, so the only transport mechanism through EBS is diffusion. In the host rock, the problem is of advection-diffusion type with highly varying parameters from one medium to other due to the variability in length, transmissivity and other transport-relevant properties of the transport paths. Daubechies’ wavelet is used as a basis function to solve the nonlinear partial differential equations arising from the model formulation of the radionuclides transport. Since the scaling functions are compactly supported, only a finite number of the connection coefficients are nonzero. The resultant matrix has a block diagonal structure, which can be inverted easily. We tested our WGM algorithm with several problems to verify the model. The solutions are very accurate with a proper selection of Daubechies’ order and dilation order. The solution is very accurate at the interfaces where the radionuclide concentration exhibits very steep gradients.Copyright

Detailed Multi Canister Release Model of Radionuclides in High Level Radioactive Waste Repository Using Wavelet Galerkin Method

This work develops a model to calculate the radionuclides release from a repository for high level radioactive waste, taking into account multiple-canister interface. Once the overpack loses its integrity, the waste glass starts to dissolve by porewater in the bentonite buffer. Bentonite is expected to have hydraulic conductivity more than three orders of magnitude less than that of the surrounding rocks. The migrating nuclide from the buffer region is transported in the near field granite host rock, then releases to the far field of the repository. A mass concentration calculation in the far field of the repository is also included in the model. The model is diffusion-advection model. The model is solved using wavelet Galerkin method (WGM). The model is devised to be fast and compact due to the compactly supported property of the Daubechies’ wavelet. Since the scaling functions are compactly supported only a finite number of the connection coefficients are nonzero. The resultant matrix has block diagonal structure, which can be inverted easily. We tested our model for a try of canisters contains 200 canisters. The results show well agreements with the results obtained from the analytic solution with a proper selection of wavelet-dilation order pairs.Copyright