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High burnup models in computer code fair

Abstract

An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power  More>>
Authors:
Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [1] 
  1. Bhabha Atomic Research Centre, Bombay (India)
Publication Date:
Aug 01, 1997
Product Type:
Conference
Report Number:
IAEA-TECDOC-957; CONF-9409411-
Reference Number:
SCA: 210400; 210200; 210100; PA: AIX-28:068598; EDB-97:130000; SN: 97001863134
Resource Relation:
Conference: IAEA technical committee meeting on water reactor fuel element modelling at high burnup and its experimental support, Windermere (United Kingdom), 19-23 Sep 1994; Other Information: PBD: Aug 1997; Related Information: Is Part Of Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting; PB: 559 p.
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; BURNUP; F CODES; FUEL ELEMENTS; BWR TYPE REACTORS; FUEL ELEMENT CLUSTERS; PHWR TYPE REACTORS; PWR TYPE REACTORS; R CODES
OSTI ID:
534371
Research Organizations:
International Atomic Energy Agency, Vienna (Austria)
Country of Origin:
IAEA
Language:
English
Other Identifying Numbers:
Journal ID: ISSN 1011-4289; Other: ON: DE98602336; TRN: XA9744791068598
Availability:
INIS; OSTI as DE98602336
Submitting Site:
INIS
Size:
pp. 91-101
Announcement Date:

Citation Formats

Dutta, B K, Swami Prasad, P, Kushwaha, H S, Mahajan, S C, and Kakodar, A. High burnup models in computer code fair. IAEA: N. p., 1997. Web.
Dutta, B K, Swami Prasad, P, Kushwaha, H S, Mahajan, S C, & Kakodar, A. High burnup models in computer code fair. IAEA.
Dutta, B K, Swami Prasad, P, Kushwaha, H S, Mahajan, S C, and Kakodar, A. 1997. "High burnup models in computer code fair." IAEA.
@misc{etde_534371,
title = {High burnup models in computer code fair}
author = {Dutta, B K, Swami Prasad, P, Kushwaha, H S, Mahajan, S C, and Kakodar, A}
abstractNote = {An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.}
place = {IAEA}
year = {1997}
month = {Aug}
}