Purdue University, School of Nuclear Engineering NUCL 200: Introduction to Nuclear Engineering Spring Semester 2009 Homework 4 Due Wednesday, March 4, 2009 1. (10 pts) Problem 5.3 from L&B. 2. (10 pts) Calculate the diffusion length and diffusion area for two materials given in problem H4-01. 3. (10 pts) Neutrons travel on average 1.1 mm in a medium before being absorbed. Calculate the macroscopic absorption cross section if the diffusion coefficient is 0.15 cm. 4. (15 pts) Problem 5.4 (a) and (b) from L&B. Hint: You will need to use the L’Hospital’s rule to Fnd the maximum ±ux, which occurs at the reactor center. 5. (10 pts) An inFnite planar source emits 5 × 10 8 neutrons cm − 2 s − 1 in an inFnite diffusing medium. Calculate the ±ux at x=-5 cm and x=5 cm from the center of the reactor. Given mean transport free path and macroscopic absorption cross section are 3 cm and 0.0419 cm − 1 , respectively. Comment on the result. 6.
This is the end of the preview. Sign up
access the rest of the document.
This note was uploaded on 05/02/2009 for the course NUCL 200 taught by Professor Jovanoic during the Spring '08 term at Purdue.