Published December 1984 | Version v1
Report Metadata-only

Evaluations on power ramp data of PWR fuels by FROST and THERMOST codes

Creators

  • 1. Mitsubishi Atomic Power Industries, Inc., Omiya, Saitama (Japan)
  • 2. Mitsubishi Heavy Industries, Ltd, (Japan)

Description

Power ramp data of Mitsubishi's PWR fuel rods tested in R-2, Studsvik were analyzed by FROST and THERMOST codes. The predictions of both codes show relatively good agreement with many measurements during or after irradiation. Especially detailed behaviours of the fuel related to axial PCMI, such as rod elongation-relaxation during ramp test, are predicted successfully. Axial and radial deformations of claddings are simulated so well that calculated stresses in claddings are considered to be reliable. The ramp data used in present analyses were obtained in two joint programs with five Japanese PWR utilities (KEPCO, KYEPCO, SEPCO, HEPCO, and JAPCO). (author)

Additional details

Publishing Information

Imprint Title
Water reactor fuel element performance computer modelling
Imprint Pagination
457 p.
Journal Page Range
p. 369-381.
Report number
IWGFPT--19

Conference

Title
Specialists' meeting on water reactor fuel element performance computer modelling.
Dates
9-13 Apr 1984.
Place
Bowness-on-Windermere (UK).

INIS

Country of Publication
International Atomic Energy Agency (IAEA)
Country of Input or Organization
International Atomic Energy Agency (IAEA)
INIS RN
16058372
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference
Descriptors DEI
CLADDING; DEFORMATION; FUEL RODS; MATHEMATICAL MODELS; STRESSES; WATER MODERATED REACTORS
Descriptors DEC
DEPOSITION; FUEL ELEMENTS; REACTOR COMPONENTS; REACTORS; SURFACE COATING