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

Akifumi Yamaji, Motoe Suzuki, Tsutomu Okubo

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Abstract

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.

Original languageEnglish
Title of host publicationInternational Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009
PublisherAtomic Energy Society of Japan
Pages49-57
Number of pages9
Volume1
ISBN (Print)9781617386084
Publication statusPublished - 2009
Externally publishedYes
EventInternational Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009 - Shinjuku, Tokyo, Japan
Duration: 2009 May 102009 May 14

Other

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

Fingerprint

Boiling water reactors
Relocation
Plutonium
Gases
Densification
Swelling
Neutrons
Uncertainty
Nuclear energy
Uranium
Recycling
Irradiation
Water
Experiments
Temperature

ASJC Scopus subject areas

  • Energy Engineering and Power Technology
  • Nuclear Energy and Engineering

Cite this

Yamaji, A., Suzuki, M., & Okubo, T. (2009). Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behaviors in FLWR design. In International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009 (Vol. 1, pp. 49-57). Atomic Energy Society of Japan.

Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behaviors in FLWR design. / Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu.

International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009. Vol. 1 Atomic Energy Society of Japan, 2009. p. 49-57.

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Yamaji, A, Suzuki, M & Okubo, T 2009, Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behaviors in FLWR design. in International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009. vol. 1, Atomic Energy Society of Japan, pp. 49-57, International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009, Shinjuku, Tokyo, Japan, 09/5/10.
Yamaji A, Suzuki M, Okubo T. Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behaviors in FLWR design. In International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009. Vol. 1. Atomic Energy Society of Japan. 2009. p. 49-57
Yamaji, Akifumi ; Suzuki, Motoe ; Okubo, Tsutomu. / Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behaviors in FLWR design. International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009. Vol. 1 Atomic Energy Society of Japan, 2009. pp. 49-57
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N2 - 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.

AB - 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.

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