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Fast reactor fuel pin behaviour modelling in the UK

Abstract

Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier  More>>
Authors:
Matthews, J R; [1]  Hughes, H [2] 
  1. UKAEA, Harwell, Didcot, Oxon (United Kingdom)
  2. Springfields Nuclear Power Development Laboratories, Springfields, Salwick, Preston (United Kingdom)
Publication Date:
Dec 01, 1979
Product Type:
Conference
Report Number:
IWGFR-31
Resource Relation:
Conference: IAEA-IWGFR specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour, Fontenay-aux-Roses (France), 28 May - 1 Jun 1979; Other Information: 36 refs; PBD: Dec 1979; Related Information: In: Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report, 183 pages.
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BURNUP; CALCULATION METHODS; F CODES; FBRF REACTOR; FISSION PRODUCTS; FUEL CYCLE; FUEL PINS; FUEL-CLADDING INTERACTIONS; RISK ASSESSMENT; S CODES; STEADY-STATE CONDITIONS; SWELLING; TEMPERATURE GRADIENTS
OSTI ID:
20243570
Research Organizations:
International Atomic Energy Agency, International Working Group on Fast Reactors, Vienna (Austria)
Country of Origin:
IAEA
Language:
English
Other Identifying Numbers:
TRN: XA0200919016335
Availability:
Available from INIS in electronic form
Submitting Site:
INIS
Size:
page(s) 112-122
Announcement Date:

Citation Formats

Matthews, J R, and Hughes, H. Fast reactor fuel pin behaviour modelling in the UK. IAEA: N. p., 1979. Web.
Matthews, J R, & Hughes, H. Fast reactor fuel pin behaviour modelling in the UK. IAEA.
Matthews, J R, and Hughes, H. 1979. "Fast reactor fuel pin behaviour modelling in the UK." IAEA.
@misc{etde_20243570,
title = {Fast reactor fuel pin behaviour modelling in the UK}
author = {Matthews, J R, and Hughes, H}
abstractNote = {Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients represent physical parameters but are fitted to experimental data. Hopefully this allows for more realistic interpolation of results and to some extent extrapolation. The physical models used In the codes are described in detail and in the following the various applications of the codes are given.}
place = {IAEA}
year = {1979}
month = {Dec}
}