Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behaviors in FLWR design

Akifumi Yamaji*, Motoe Suzuki, Tsutomu Okubo

*この研究の対応する著者

研究成果: Conference contribution

抄録

The concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed and being studied at Japan Atomic Energy Agency (JAEA) to achieve effective and flexible utilization of the uranium and plutonium resources based on the well-developed LWR technology. FLWR is a Boiling Water Reactor (BWR) type concept and it is planned to be introduced by two stages. In the first stage, the MOX fuels are irradiated in a similar condition to that of the current BWR but with a harder neutron spectrum. The core average discharge burnup is about 45 GWd/tHM. In the second stage, the neutron spectrum is further hardened to achieve a multiple recycling of plutonium with higher burnups. In order to design and evaluate the integrities of FLWR fuel rods, the uncertainties in FEMAXI-6 calculations and models for predicting LWR MOX fuel behaviors need to be evaluated. As an introduction to the evaluation process, the Test-Fuel-Data-Base (TFDB) obtained from the Halden reactor experiments (IFA- 514) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX. Based on the present investigation, the following models were found to be particularly important. Namely the fission gas release (FGR), pellet densification, swelling, and relocation models. These models of FEMAXI-6 have been developed and the parameters have been optimized based on the past UO2 irradiation test data. For predicting MOX fuel behavior, the FGR model has a relatively large uncertainty and causes a large uncertainty in the FGR calculations. On the other hand, the uncertainties in the other models are within the range expected by the property variations of typical UO2 fuels. Hence, the densification, swelling, and the relocation models of FEMAXI-6 can be applied to MOX fuel analyses provided the corresponding MOX property variations are taken into account in the input parameters of these models. When the FGR of the MOX fuels is reasonably predicted, the calculated pellet centerline temperature can be expected to be within about plus or minus 50 K of the measurements. These uncertainties need to be taken into account when designing and evaluating the integrities of FLWR fuel rods.

本文言語English
ホスト出版物のタイトルInternational Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009
出版社Atomic Energy Society of Japan
ページ49-57
ページ数9
ISBN(電子版)9781617386084
出版ステータスPublished - 2009 1月 1
外部発表はい
イベントInternational Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009 - Shinjuku, Tokyo, Japan
継続期間: 2009 5月 102009 5月 14

出版物シリーズ

名前International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009
1

Other

OtherInternational Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009
国/地域Japan
CityShinjuku, Tokyo
Period09/5/1009/5/14

ASJC Scopus subject areas

  • エネルギー工学および電力技術
  • 原子力エネルギーおよび原子力工学

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