An innovative acoustic sensor for first in-pile fission gas release determination — REMORA 3 experiment

Among numerous research projects devoted to the improvement of the nuclear fuel behaviour knowledge, the development of advanced instrumentation for in-pile experiments in Material Testing Reactor is of great interest. In the frame of JHR reactor, new requirements have arisen creating new constraints. An acoustic method was tested with success during a first experiment called REMORA 3 in 2010 and 2011, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was leading at OSIRIS reactor (CEA Saclay, France). The maximal temperature during the irradiation test was about 150 °C. [1], [2]. We have developed thick film transducers produced by screen-printing process. They offered a wide range of possible application for the development of acoustic sensors and piezoelectric structure for harsh temperature environment measurements [3]. We proposed a screen-printed modified Bismuth Titanate piezoelectric element on alumina substrate allowing acoustic measurements [4] for JHR environment. In this paper we will focus on the mechanical design of the new sensor. This acoustic sensor is composed of an acoustic element for generation and detection of acoustic waves propagating into a cavity filled with gaz. We will detail the choice of piezoelectric materials, the thickness of the different layers, the cavity shapes, the electrical connections, the means of assembly of the different parts. Theoretical and experimental results will be given. All that point will be discussed in terms of acoustic sensor sensitivity versus dimensional constraints, in the case of a high temperature range working.

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Fission Gas Release from Irradiated Uranium-Plutonium Nitride Fuel

Atomic Energy ◽

10.1007/s10512-021-00712-z ◽

2020 ◽

Vol 129 (2) ◽

pp. 103-107

Author(s):

A. F. Grachev ◽

L. M. Zabud’ko ◽

M. V. Skupov ◽

F. N. Kryukov ◽

V. G. Teplov ◽

...

Keyword(s):

Gas Release ◽

Nitride Fuel ◽

Fission Gas ◽

Uranium Plutonium ◽

Fission Gas Release

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Fission Gas Release from Uranium Heated in Carbon Dioxide

Journal of Nuclear Science and Technology ◽

10.1080/18811248.1965.9732186 ◽

1965 ◽

Vol 2 (5) ◽

pp. 168-171

Author(s):

Kiyoaki TAKETANI ◽

Katsuichi IKAWA

Keyword(s):

Carbon Dioxide ◽

Gas Release ◽

Fission Gas ◽

Fission Gas Release

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Fission gas release and swelling in uranium–plutonium mixed nitride fuels

Journal of Nuclear Materials ◽

10.1016/j.jnucmat.2004.01.002 ◽

2004 ◽

Vol 327 (2-3) ◽

pp. 77-87 ◽

Cited By ~ 36

Author(s):

Kosuke Tanaka ◽

Koji Maeda ◽

Kozo Katsuyama ◽

Masaki Inoue ◽

Takashi Iwai ◽

...

Keyword(s):

Gas Release ◽

Fission Gas ◽

Nitride Fuels ◽

Uranium Plutonium ◽

Fission Gas Release

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Pellet-Cladding Interaction of LMFBR Fuel Elements at Unsteady State: An Introduction to Computer Code ISUNE-5

Journal of Engineering for Power ◽

10.1115/1.3230785 ◽

1981 ◽

Vol 103 (4) ◽

pp. 627-636 ◽

Cited By ~ 1

Author(s):

B. M. Ma

Keyword(s):

Reactor Core ◽

Fatigue Cracking ◽

Computer Code ◽

Fuel Burnup ◽

Fuel Pellet ◽

Gas Release ◽

Unsteady State ◽

Fission Gas ◽

Fuel Elements ◽

Fission Gas Release

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.

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