NucE 502 Advanced Thermal Hydraulic Analysis

Course Description

This course discusses detailed thermal-hydraulic analysis of reactor systems with the emphasis on the application of conservation equations for single and two-phase in detailed modeling of reactor cores using three dimensional subchannel analysis methods. NucE 501 is not a requirement for this course. There will be some initial lectures on two-phase flow, boiling and critical heat flux which will provide the basis for understanding the calculated core limits from the subchannel analysis. The purpose of the course will be to examine the reactor core thermal-hydraulic design for core limit analysis such as critical heal flux or critical power for PWRs and BWRs as well as for temperature limited liquid metal and gas cooled reactor designs. The governing sets of equations which form the basis for the three-dimensional thermal-hydraulic methods commonly used in the nuclear industry will be derived and discussed as well as specific models which are used for closure. Thermal-hydraulic methods used to analyze steam generators will also be discussed.

Hot assembly analysis will be performed as well as core wide analysis to determine the hot assembly and resulting hot subchannels in the core. Engineering hot channel factors will also be determined and applied in the analysis to determine the limiting subchannels. Newer fuel designs, specifically for BWRs will be analyzed on subchannel basis and compared to the current operating limits. A similar approach will also be used for PWRs.

New statistical approaches, which are currently used by the vendors, to calculate the limits for the hot channel performance in the core will also be discussed. The hot channel factors will also be discussed and shown how they are factored into the three-dimensional core analysis. The students will use state-of-the-art three-dimensional computer codes to model fuel assemblies and the reactor core to find the most limiting fuel pin and hottest subchannel. The computer codes that will be discussed include: COBRA-IV, VIPRE-II, THERMIN-II, and COBRA-TF. The COBRA-TF code will also be discussed since it is the only subchannel code that can model situations where the critical heat flux has been exceeded. Other subchannel codes will also be discussed using recent papers from the literature. A project will be given using one of these codes. The course project will utilize the the COBRA-IV subchannel code. This is an .exe application running on PC under Windows environment. You must have access to a Window OS computer to complete the Course Project.

Current Course Offering Schedule via C&DE

Semester: Summer 2013
Time and Date: NA - Online
Location: NA - Online
Start Date: May 13, 2013
End Date: August 9, 2013

Course Instructor:

PSU Course Catalog Description

502 ADVANCED THERMAL HYDRAULIC ANALYSIS OF LWRs (3) In-depth analysis of the thermal hydraulic design in LWRs. Topics include: LWR design criteria, fuel rod design, subchannel analysis, uncertainties analysis, and system design.

Required Course Textbooks (2)

Class Notes and Handouts will be posted in ANGEL

“Nuclear Systems II, Elements of Thermal-Hydraulic Design” by Todreas and Kazimi

“Nuclear Systems I, Thermal-Hydraulic Fundamentals” also by Todreas and Kazimi

Computer and Technology Requirements:

The enrollment process is:

1. Contact the the Office of Continuing and Distance Education for admission into the course at 814-865-7643.

2. Complete and submit the Registration Form via mail, fax, or in person. See the Registration Page.

Office for Disability Services
Penn State welcomes students with disabilities into the University’s educational programs. If you have a disability‐related need for reasonable academic adjustments in this course, contact the Office for Disability Services (ODS) at 814‐863‐1807 (V/TTY). For further information regarding ODS, please visit the Office of Disability Services website at http://equity.psu.edu/ods