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.

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

Engineering Practices for Tokamak Window Assemblies

For fusion tokamak reactors the diagnostics and RF heating systems require the use of components with parts made of non-metallic materials. These can form part of the vacuum boundary of the tokamak which is the primary safety boundary and have a function of containing tritium fuel or activated gases and particulate debris. The engineering practices for such components and non-metallic materials are in an early state of preparation and require development to enable systems to be used in a safety and licensing context. Such developments will have to reflect the brittle nature of the materials, and are likely to be based on established arguments developed within the nuclear industry, such as containment and defence in depth. Given these requirements this task is a major challenge. The window systems fall broadly into two categories: • Transmission windows for the input of high-power microwaves to drive and heat the plasma; • Diagnostic windows to monitor the plasma. Currently there are no established fusion design codes that can be used to assure nuclear safety and a consistent engineering approach for either application. This paper reviews the progress made in developing such practices for transmission and diagnostic windows made from ceramic materials. The investigations undertaken and the engineering practices addressed for the tokamak windows generally fall into the following areas: • reviews of potential candidate materials along with a summary of the available property data; • definition of the function of torus window assemblies and an outline of the complexity and variety of design considerations (including historical failures, and statutory requirements); • development of the design methodology for technical ceramics; • definition of the design routes considered and selected (rule, analysis, experiment); • consideration of the material data available (or lack of) for technical ceramics and their failure criteria; • qualification and design of metallic / ceramic joints; • definition of the requirements with regard to quality control, from manufacture to in-service inspection; • development and formation of a draft code procedure. The practices and procedures developed are considered to be an important contribution and significant step forward in the development of a fusion tokamak windows code. Important contributions have been made to the design, procurement and installation philosophies for windows, especially the development of design criteria and the application of pressure proof-testing. This paper provides a review of key requirements and issues, with recommendations to allow development of the code for acceptance by nuclear regulators for tokamaks such as the International Tokamak Experimental Reactor (ITER) and future fusion reactor power plants.

Implementation of a Pressure Drop Model for the CFD Simulation of Clogged Containment Sump Strainers

The present study aims at modelling the pressure drop of flows across growing cakes of compressible, fibrous materials which may form on the upstream side of containment sump strainers after a loss-of-coolant accident (LOCA). The model developed is based on the coupled solution of a differential equation for the change of the pressure drop in terms of superficial liquid velocity and local porosity of the fibre cake and a material equation that accounts for the compaction pressure dependent cake porosity. Details of its implementation into a general-purpose three-dimensional computational fluid dynamics code (CFD) are given. An extension to this basic model is presented, which simulates the time dependent clogging of the fibre cake due to capturing of suspended particles as they pass trough the cake. The extended model relies on empirical relations which model the change of pressure drop and removal efficiency in terms of particle deposit in the fibre cake.

Analytical Investigation of Deaerator Steam Inlet Nozzle Corrosion Allowance in a CANDU Station

Flow-accelerated corrosion (FAC) is a phenomenon that results in metal loss from piping, vessels and equipment made of carbon steel. This metal loss can lead to stress to occur at the steam inlet nozzle side, where it is located at the side of the deaerator. This paper presents a method to find the thickness critical of the steam inlet nozzle. A Finite Element (FE) model of the pressure vessel head was created to perform a stress analysis using NX Nastran 5.0. By applying materials properties, loads and constraints to the model, the results obtained are required to satisfy the following criterion: vonMises≥SySy=YieldStrength The results obtained from the stress analysis were analyzed to obtain a corrosion allowance and it was compared to the recommended value from a normal deaerator design, which is roughly 0.25 inches. From the FE model, and by continuously reducing the thickness of the nozzle, it was determined that the corrosion allowance is 0.229 inches, and that the percentage error was 8.4%.

A Study for the Practicality of the 3D Full Model in the SSI Analysis on the Nuclear Power Plant

This study is a part of research project ‘the practicality of the 3-D Full Model in the SSI (Soil-Structure Interaction) Analysis on the Nuclear Power Plant’ from June 2006 to May 2009. The purpose of the project is getting advanced and more accurate results with the practicality of Full Model SSI analysis. The SSI analyses of the NPP were being performed for the APR1400 (Advanced Power Reactor 1400MWe) design because the APR1400 was developed as a Standard NPP concept enveloping suitable soil conditions. The soil properties of SSI analyses were from lower bound characteristics to firm condition. The Standard NPP that was developed in early 1990 had the rigid foundation assumption because rigid mat had many advantages, simple modeling and quick calculation. But this assumption may create licensing issue because the rigid mat could include uncertainties in the modeling and analysis procedure. Now, the advanced computing capacities can offer the calculating environment using Full Model in the SSI analysis, and through this study the Full Model SSI analysis may be applied to the NI (Nuclear Island) of APR1400. The full modeling concept can be distinguished into two methods. First one is the Full Model about excavated soil part below the ground surface only, the super structures can be beam sticks. Another modeling is the Full Model about excavated soil part below ground surface together with the super structure. Finally, the Full Model of excavated soil part with the super structure was built, each structure of the Full Model was verified.

Effects of Ruthenium Release in Oxidizing Conditions on a BWR Source Term

The release of ruthenium in oxygen-rich conditions from the reactor core during a severe accident may lead to formation of significantly more volatile ruthenium oxides than produced in steam atmosphere. The effect of volatile ruthenium release in a case a reference BWR nuclear plant was studied to get rough-estimates of the effects on the spreading of airborne ruthenium inside the containment and reactor building and the fission product source term. The selected accident scenario starting during shutdown conditions with pressure vessel upper head opened was a LOCA with a break in the bottom of the RPV. The results suggest that there is a remarkable amount of airborne Ru in the containment atmosphere, unlike with the standard MELCOR Ru release model which predicts no airborne Ru at all in the selected case. The total release of ruthenium from the core can be 5000 times the release predicted by the standard model. Based on the performed plant scoping studies it seems reasonable to take the release of volatile ruthenium oxides into account when assessing source terms for plants during shutdown states.

Sensitivity Investigation of Specimen Shape on Elastic Behaviour of Perforated Plate

There are various types of tube support plates installed in a steam generator according to the component designer’s preference. Most widely used types of tube support plates are BTSP (broached tube support plate), ATSG (advanced tube support grid), and the eggcrate. In this study, trefoil BTSP specimens made of ASME stainless steel are analyzed and tested. This study is to investigate the effect of specimen shape on an elastic behavior of trefoil BTSP through the compression and bending tests. Prior to the compression and bending tests of BTSP specimens, the equivalent elastic properties of BTSP unit cell are analyzed by the finite element analysis according to the different loading orientation as well as size of the model. Autodesk® Inventor™ software was used to make an analytical model and ANSYS® software was used for the finite element analysis and post-processing. Five and three different shapes of trefoil BTSP specimens are machined and utilized for the compression and bending (4-point and 3-point side bending) tests, respectively. Through the finite element analyses, compression, and bending tests, the equivalent elastic modulus of trefoil BTSP specimen is suggested to be 6,254MPa (907ksi) and the equivalent Poisson’s ratio as 0.64. Specifically the CS5 type specimen which has a ratio of one-fourth (= width/length) was revealed as an appropriate shape of specimen to show those elastic behavior.

Use of Shear Lugs for Anchorage to Concrete

Anchoring structures, systems and components to concrete is a significant activity in the design and construction of a nuclear power plant. Early in this decade the Concrete Capacity Design method (CCD) was adopted by the American Concrete Institute (ACI) for use in the structural design for both commercial and nuclear facilities. This design method and associated qualification tests brings new challenges to designing efficient means for anchoring to concrete structures. Although the CCD method provides guidance on many aspects of concrete anchorage there are a few areas, pertinent to nuclear power plant construction, that are not covered or require significant interpretation of the most recent codes. This paper will focus on the design of shear lugs used to resist significant lateral loads. Results from laboratory tests of shear lugs are presented. These full scale tests considered the interaction of tension and shear loads on the performance of shear lug assemblies. Recommendations for the efficient use of shear lugs are provided.

A Lifecycle Management Program for NPP Turbine Balance of Plant

A lifecycle management program, for turbine balance of plant of light water reactor units, was proposed and implemented from the view point of system and equipment supplier to secure the high availability factor throughout the long intended residual lifetime. The program consists of unit surveillance analyzing operation and inspection data, degradation assessment for the equipment and prospecting for the future by appropriate measures to address the issues over the units based on both technical and economical criteria. The surveillance revealed the issue that the generating power would be adversely affected by main steam pressure reduction due to the scale adhesion to SG tubes. The prospect for the unit future was presented as an alteration of the water treatment to HAVT accompanied with optimum design for the replacement of the auxiliary heat exchangers of MSR and feedwater heaters.