ANSYS—Reactor Design

Users can optimise reactor performance by better understanding the effects and impact of feed locations, vessel geometries and internals, vibrations, failures, dead spots, shear rates, resident time distributions, hot spots and particle size distributions. 

ANSYS also provides insight and a detailed understanding of the formation and dispersion of pollutants (Nox, Sox, mercury and other VOCs) for a range of flow problems, including turbulence, chemical reactions, heat and mass transfer and multiphase flows. 

ANSYS simulation enables engineers to study multiphase distribution, heat and mass transfer calculations, chemical kinetics and reaction of gas-liquid reactions. These include the design of plate columns, packed columns and bubble columns, a loop reactor and bioreactor development, gas-in-liquid dispersion studies and emulsion design. 

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Location and Design Criteria for Area Radiation Monitoring Systems for Light Water Nuclecu- Reactors, ANSI/ANS-HPSSC-6. 8.1, American National Standards Institute, New Yoik, 1981. 

ANSI/ANS-57.2-1983, "Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants."... 

ANSI/ANS-57.3-1983,"Design Requirements for New Fuel Storage Fac iIi t i es at Li ght Water Reactor PI ants."... 

The American Nuclear Society standard, "Design Bases for Facilities for LMFBR Spent Fuel Storage in Liquid Metal Outside the Primary Coolant Boundary," ANSI/ANS-54.2-1985, provides guidance for the storage of fuel from sodium cooled reactors. 

This section discusses the measures employed to provide and maintain the integrity of the Reactor Coolant Pressure Boundary (RCPB) throughout the facility s design lifetime. The RCPB is defined in accordance with ANSI/ANS 51.1-1983. Included are all pressure containing components such as pressure vessels, piping, pumps, and valves which are ... 

ANSI/ANS 2.8, "Standard for Determining Design Basis Flooding at Power Reactor Sites", American Nuclear Society. 

The System 80+ Standard Design incorporates a variety of methods to assure the integrity of the Reactor Coolant System (RCS). The RCS is designed in accordance with 10 CFR 50.55a (Reference 8) which further references the ASME B PV Code and other accepted industry codes and standards. In addition, the reactor coolant pressure boundary is defined in accordance with ANSI/ANS 51.1 (Reference 9). 

ANSI/ANS 51.1, "Nuclear Safety criteria for the Design of Stationary Pressurized Water Reactor Plants", 1983. 

For companies who rely heavily on simulation software for process examination, there will be a gradual transition to incorporate the capabilities of PI plant in these. As an example, ANSYS Huent is involved in a major project that is developing correlations for two-phase heat transfer at the micro-scale, suitable for micro-reactor and micro-heat exchanger design and sizing, see Web 1 (2007). 

The application ofthe N18.2 checklist has been reviewed against the APIOOO design and PRA, and been appropriately updated to reflect the plant s specific design features. On this basis, while potential Anticipated Transient Without Scram (ATWS) faults are included in ANSI N 18.2, for the APIOOO no causes for these have been identified within the Design Basis (i.e. the initiating event frequency is less than 10-5 per reactor year). These faults are addressed via PRA and Severe Accident Analysis (as described in Section 5.4 and 5.5 of this chapter). This issue is discussed in sub-section 4.4.1.3 ofthe APIOOO Fault Schedule (Reference 5.1). 

Nuclear safety criteria for the design of stationary pressurized water reactor plant, ANSI/ANS-51.1-1983, (R1988). 

The safety classification methodology was derived from the standard, ANSI/ANS-51.1, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants" (Reference 2). The plan describes three safety-related classes and one nonnuclear safety class. The definitions for Safety Classes 1, 2, and 3 were found to comply with the standard definitions contained in 10 CFR 50.55a and NRC Regulatory Guide 1.26 (References 3 and 4). The definition for the nonnuclear safety class was included, but is outside the scope of this evaluation. 

This section and Section 10.2, "Training," outline the minimum requirements for both the selection and training of selected reactor facility personnel. These criteria are based on commercial industry standards such as American National Standards Institute (ANSI)/American Nuclear Standards (ANS) 3.1, "Selection, Qualification, and Training of Personnel for Nuclear Power Plants" (Reference 1) and DOE Order 5480.6, "Safety of DOE-Owned Reactors" (Reference 2). This section also discusses the minimum staffing levels designed to assure that the reactor is operated with adequate numbers of qualified personnel to assure safe operations. 

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