Experiments of melt jet-breakup for agglomerated debris formation using a metallic melt

Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Abe, Yutaka*

Nuclear Engineering and Design, 386, p.111575_1 - 111575_17, 2022/01

https://doi.org/10.1016/j.nucengdes.2021.111575

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

https://doi.org/10.1080/00223131.2019.1637795

no abstracts in English

Improved-EDC tests on the Zircaloy-4 cladding tube with an outer surface pre-crack

Shinozaki, Takashi*; Udagawa, Yutaka; Mihara, Takeshi; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(9), p.1426 - 1434, 2016/09

https://doi.org/10.1080/00223131.2015.1123658

Behavior of high burnup advanced fuels for LWR during design-basis accidents

Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Sugiyama, Tomoyuki

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09

http://www.euronuclear.org/events/topfuel/topfuel2015/transactions/topfuel2015-transactions-oral-2.pdf#page=10

Advanced fuels which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has started a new extensive research program called ALPS-II program (Phase II of Advanced LWR Fuel Performance and Safety program). This program is primarily composed of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup advanced fuels shipped from European nuclear power plants. This paper describes an outline of this program and some experimental results with respect to RIA and LOCA which have been obtained in this program.

The Evaluation of the influence of hydride rim and biaxial stress condition on the cladding failure under a reactivity-initiated-accident by using EDC test method

Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

JAEA-Research 2014-025, 34 Pages, 2014/12

JAEA-Research-2014-025.pdf:6.05MB

EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.

Heat transfer from fuel rod surface under reactivity-initiated accident conditions; NSRR experiments under varied cooling conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

JAEA-Data/Code 2013-021, 43 Pages, 2014/02

JAEA-Data-Code-2013-021.pdf:2.74MB
JAEA-Data-Code-2013-021-appendix(CD-ROM).zip:126.49MB

In order to study the effects of cooling conditions on the boiling heat transfer from the fuel rod surface to the coolant water, RIA-simulating experiments with fresh fuels had been conducted in the nuclear safety research reactor (NSRR) under cooling conditions with subcoolings of 10 to 80 K, flow velocities of 0 to 3 m/s, pressures of 0.1 to 16 MPa. In addition, pre-irradiated fuels had been subjected to the NSRR experiments under cooling conditions with subcoolings of 80 K, stagnant water, and atmospheric pressure. Out of the NSRR experiments, this report presents the fuel specifications, the test conditions, and the transient records during the pulse operations for the cases that the cladding temperature had been successfully measured. Characteristic parameters such as cladding peak temperatures were extracted from the transient records for summarizing the effects of cooling conditions and pre-irradiation on the heat transfer from the cladding surface.

RCCA ejection fault studies; Application of a burnup threshold for the evaluation of the radiological releases

Udagawa, Yutaka; Sugiyama, Tomoyuki; Suzuki, Motoe; Nagase, Fumihisa

IAEA-TECDOC-CD-1709, p.153 - 160, 2013/06

Model development and verifications for fission gas inventory and release from high burnup PWR fuel during simulated reactivity-initiated accident experiment at NSRR

Suzuki, Motoe; Udagawa, Yutaka; Sugiyama, Tomoyuki; Nagase, Fumihisa

Proceedings of Annual Topical Meeting on Water Reactor Fuel Performance (TopFuel 2012) (USB Flash Drive), 6 Pages, 2012/09

Behavior of fission gas release (FGR) analysis is performed for the high burnup PWR fuels which are pulse-irradiated in the simulated Reactivity-Initiated Accident (RIA) experiment conducted at NSRR (Nuclear Safety Research Reactor) in Japan Atomic Energy Agency. The FGR model consists of two main parts: FEMAXI-7 calculates fission gas bubble growth in grain boundaries during base-irradiation, while RANNS performed the grain separation and burst release of gas from grain boundary inventory in the rapidly heated pellet in the RIA experiment. The calculated results are compared with the measured data, which resulted in a rough agreement for the amount of fission gas in pellet and burst FGR.

Influence of coolant temperature and power pulse width on fuel failure limit under reactivity-initiated accident conditions

Sugiyama, Tomoyuki; Udagawa, Yutaka; Suzuki, Motoe; Nagase, Fumihisa

Proceedings of 2011 Water Reactor Fuel Performance Meeting (WRFPM 2011) (CD-ROM), 6 Pages, 2011/09

The Japan Atomic Energy Agency has performed pulse irradiation tests using the NSRR to investigate fuel behavior under Reactivity-Initiated Accident (RIA) conditions. The NSRR tests have provided data of the pellet-cladding mechanical interaction (PCMI) failure of high burnup fuels up to 77 GWd/t. In particular, the PCMI failure limit is the important information which is needed in the reactor safety review. However, there are some differences between the NSRR tests and RIAs supposed in power reactors, such as the coolant temperature and the width of power pulse. Influence of these differences should be quantitatively evaluated in order to estimate the PCMI failure limit anticipated under the power reactor conditions from the NSRR data. This paper presents experimental results from a set of room and high temperature RIA tests, and discusses the evaluation procedure of the influence of coolant temperature and power pulse width on the failure limit on the basis of the experimental data.

Evaluation of initial temperature effect on transient fuel behavior under simulated reactivity-initiated accident conditions

Sugiyama, Tomoyuki; Udagawa, Yutaka; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 47(5), p.439 - 448, 2010/05

https://doi.org/10.3327/jnst.47.439