Experimental study on prevention of high cycle thermal fatigue at the core outlet of advanced sodium-cooled fast reactor; Characteristics of temperature fluctuations and countermeasures to mitigate temperature fluctuations at a bottom of upper internal structure

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Nagasawa, Kazuyoshi*; Kurihara, Akikazu; Tanaka, Masaaki

JAEA-Research 2022-009, 125 Pages, 2023/01

JAEA-Research-2022-009.pdf:29.22MB

The design studies of an advanced loop-type sodium-cooled fast reactor (Advanced- SFR) have been carried out by the Japan Atomic Energy Agency (JAEA). At the core outlet, temperature fluctuations occur due to mixing of hot sodium from the fuel assembly with cold sodium from the control rod channels and radial blanket assembly. These temperature fluctuations may cause high cycle thermal fatigue around a bottom of Upper Internal Structure (UIS) located above the core. Therefore, we conducted a water experiment using a 1/3 scale 60 degree sector model that simulated the upper plenum of the advanced loop-type sodium-cooled reactor. And we proposed some countermeasures against large temperature fluctuations that occur at the bottom of the UIS. In this report, we have summarized that the effect of the countermeasure structure to mitigate the temperature fluctuation generated at the bottom of UIS is confirmed, and the Reynolds number dependency of the countermeasure structure and the characteristics of the temperature fluctuation on the control rod surface.

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 1; Proposal of countermeasures to mitigate temperature fluctuations around control rods

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.89 - 96, 2021/10

Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.

Release behavior of radionuclides from MOX fuels irradiated in a fast reactor during heating tests

Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.

Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08

https://doi.org/10.1016/j.jnucmat.2020.152199

In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.

Measurement of Velocity Field in Five Jets Water Test (FIWAT) for thermal striping in sodium-cooled fast reactor

Aizawa, Kosuke; Kobayashi, Jun; Tanaka, Masaaki; Kurihara, Akikazu; Ishida, Katsuji*; Nagasawa, Kazuyoshi*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 10 Pages, 2018/11

A conceptual design of an advanced loop type sodium cooled reactor has been carried out in the Japan Atomic Energy Agency (JAEA). Temperature fluctuation is caused by mixing of fluids at different temperature from the control rod channels and the core fuel assemblies, high cycle thermal fatigue may arise on the Core Instrument Plane (CIP) at bottom of the Upper Internal Structure (UIS). In JAEA, 1/3-scaled five jets water tests (FIWAT) have been performed in order to investigate thermal striping phenomena around the CIP. In this study, the velocity field was measured in the mixing area between the jet outlet and the bottom of the structure by using particle image velocimetry (PIV) to compare with the temperature fluctuation characteristics.

Fabrication and short-term irradiation behaviour of Am-bearing MOX fuels

Kihara, Yoshiyuki; Tanaka, Kosuke; Koyama, Shinichi; Yoshimochi, Hiroshi; Seki, Takayuki; Katsuyama, Kozo

NEA/NSC/R(2017)3, p.341 - 350, 2017/11

https://www.oecd-nea.org/pt/iempt14/

In order to investigate the effect of the addition of americium to MOX fuels on the irradiation behaviour, the "Am-1" program is being conducted at the JAEA. The Am-1 program consists of two short-term irradiation tests of 10-min and 24-h irradiation periods, and a steady-state irradiation test. The short-term irradiation tests and their post irradiation examinations (PIEs) have been successfully completed. To date, the data for PIE of the Am-MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation have been obtained and reported. In this paper, the results obtained from the Am-1 program are reviewed and detailed descriptions of the fabrication and inspection techniques for the Am-MOX fuels prepared for the program are provided. PIE data for the Am-MOX fuels at the initial stage of irradiation have been accumulated. In this paper, unpublished PIE data for the Am-MOX fuels are also presented.

High temperature physicochemical properties of irradiated fuels

Ishikawa, Takashi; Onishi, Takashi; Hirosawa, Takashi; Tanaka, Kosuke; Katsuyama, Kozo

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 10 Pages, 2017/00

http://hotlab.sckcen.be/proceedings

Thermophysical properties of americium-containing barium plutonate

Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*

Journal of Nuclear Science and Technology, 52(10), p.1285 - 1289, 2015/10

https://doi.org/10.1080/00223131.2015.1029554

Polycrystalline specimens of americium-containing barium plutonate have been prepared by mixing the appropriate amounts of (PuAm)O and BaCO powders followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The sintered specimens had a single phase of orthorhombic perovskite structure and were crack-free. The elastic moduli were determined from the longitudinal and shear sound velocities. The Debye temperature was also determined from the sound velocities and lattice parameter measurements. The thermal conductivity was calculated from the measured density at room temperature, literature values of heat capacity, and thermal diffusivity measured by laser flash method in vacuum. The thermal conductivity of americium-containing barium plutonate was roughly independent of the temperature and was almost the same magnitude as that of BaPuO and BaUO.

Early-in-life fuel restructuring behavior of Am-bearing MOX fuels

Tanaka, Kosuke; Sasaki, Shinji; Katsuyama, Kozo; Koyama, Shinichi

Transactions of the American Nuclear Society, 113(1), p.619 - 621, 2015/10

In order to evaluate the microstructural change behavior of Am-MOX fuels at the initial stage of irradiation, detailed investigations using image analysis were performed on X-ray Computed Tomography (X-ray CT) images and on ceramographs from fuels irradiated in both B11 and B14.

Influence of boron vapor on transport behavior of deposited CsI during heating test simulating a BWR severe accident condition

Sato, Isamu; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shinichi

Journal of Nuclear Materials, 461, p.22 - 28, 2015/06

https://doi.org/10.1016/j.jnucmat.2015.02.032

We observed one of the simplified processes by conducting primitive experiments. CsI was heated at 1323 K to be vaporized and deposited on sampling parts with a temperature range of 1023 - 423 K and then BO was vaporized at 1973 K to be reacted with Cs/I there. After heating tests, each sampling part was soaked into alkali water to dissolve the surface-deposits for ICP-MS analysis. The results showed that CsI deposited at the sampling parts kept above approx. 850 K was striped by BO vapour. This behaviour will be thermodynamically discussed to study the Cs/I/B chemistry in the severe accidents.

Research program for the evaluation of fission product and actinide release behaviour, focusing on their chemical forms

Miwa, Shuhei; Yamashita, Shinichiro; Ishimi, Akihiro; Osaka, Masahiko; Amaya, Masaki; Tanaka, Kosuke; Nagase, Fumihisa

Energy Procedia, 71, p.168 - 181, 2015/05

BB2013-2241.pdf:0.88MB

https://doi.org/10.1016/j.egypro.2014.11.867

A basic study towards enhanced safety management of irradiated fuels and materials from a severe accident is underway utilizing JAEA's hot laboratory complex in Oarai. The present study that consists of three basic research programs is aimed at contributing to building enhanced safety management measures (including radioactive decontamination, evaluation measurements, safekeeping, treatment and disposal) of irradiated fuels and materials from the severe accident. In this paper, not only the overview of activities of individual research programs but also the several preliminary results were shown together with future plans.