Transient response of LWR fuels (RIA)

宇田川 豊; 更田 豊志*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

https://doi.org/10.1016/B978-0-12-803581-8.11682-0

This article aims at providing a general outline of fuel behavior during a reactivity-initiated accident (RIA) postulated in light water reactors (LWRs) and at showing experimental data providing technical basis for the current RIA-related regulatory criteria in Japan.

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 宇田川 豊; 三原 武; 天谷 政樹; 垣内 一雄

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

A pulse-irradiation test CN-1 on a high-burnup MOX fuel with M5 cladding was conducted at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Agency (JAEA). Although the transient signals obtained during the pulse-irradiation test did not show any signs of the occurrence of PCMI failure, the failure of the test fuel rod was confirmed from the visual inspection carried out after test CN-1. Analyses using fuel performance codes FEMAXI-8 and RANNS were also performed in order to investigate the fuel behavior during normal operation and pulse-irradiation regarding the test fuel rod of CN-1, and the results were consistent with this observation result. These experimental and calculation results suggested that the failure of test fuel rod of CN-1 was not caused by hydride-assisted PCMI but high-temperature rupture following the increase in rod internal pressure. The occurrence of this failure mode might be related to the ductility remained in the M5 cladding owing to its low content of the hydrogen absorbed during normal operation.

Behaviors of high-burnup LWR fuels with improved materials under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

Fuels for light water reactors (LWRs) which consist of improved cladding materials and pellets have been developed by utilities and fuel vendors to acquire better 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 adequacy of the present regulatory criteria in Japan and safety margins regarding the fuel with improved materials, Japan Atomic Energy Agency (JAEA) has conducted ALPS-II program sponsored by Nuclear Regulation Authority (NRA), Japan. In this program, the tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) have been performed on the high burnup advanced fuels irradiated in commercial PWR or BWR in Europe. This paper presents recent results obtained in this program with respect to RIA, and main results of LOCA experiments, which have been obtained in the ALPS-II program, are summarized.

Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

JAEA has conducted a research program called ALPS-II program for advanced fuels of LWRs. In this program, the tests simulating a RIA and a LOCA have been performed on the high burnup advanced fuels irradiated in European commercial reactors. The failure limits of the high-burnup advanced fuels under RIA conditions have been obtained by the pulse irradiation tests at the NSRR in JAEA. The information about pellet fragmentation etc. during the pulse irradiations was also obtained from post-test examinations on the test rods after the pulse irradiation tests. As for the simulated LOCA test, integral thermal shock tests and high-temperature oxidation tests have been performed at the RFEF in JAEA. The fracture limits under LOCA and post-LOCA conditions etc. of the high-burnup advanced fuel cladding have been investigated, and it was found that in terms of these materials the fracture boundaries do not decrease and the oxidation does not significantly accelerate in the burnup level examined.

Recent research activities using NSRR on safety related issues

宇田川 豊; 杉山 智之*; 天谷 政樹

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04

JAEA launched ALPS-II program in 2010 in order to obtain regulatory data for advanced fuels. Five new reactivity-initiated accident (RIA) simulated tests on the advanced fuels have been performed. The first two fuels tested, VA-5 and VA-6, were 1717-PWR-type with stress-relieved and recrystallized M-MDA cladding tube, and irradiated to ~80 GWd/tU. The cladding failed due to the pellet-cladding mechanical interaction. Fission gas dynamics tests to promote a better understanding of the behavior of fission gas during an RIA are planned. A recent qualification test on a prototype pressure sensor demonstrated its ability to obtain history data of transient fission gas release. JAEA also launched a new experiment program using NSRR to investigate fuel degradation behaviors in the temperature region beyond-DBA LOCAs.

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

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 杉山 智之

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

高燃焼度領域での燃料性能を向上させるとともに既設の原子炉の安全性を向上させるため、高耐食性被覆管や核分裂生成ガス放出を抑えたペレットで構成された改良型燃料が事業者や燃料メーカによって開発されてきた。このような改良型燃料の現行の規制基準や安全裕度の妥当性を評価するため、またこれらに係る将来の規制のためのデータベースを提供するため、原子力機構はALPS-IIと呼ばれる新しい研究プロクラムを開始した。このプログラムは、欧州から輸送された高燃焼度改良型燃料を対象とした反応度事故(RIA)模擬試験及び冷却材喪失事故(LOCA)模擬試験から主に構成されている。本論文では、このプログラムの概要及び現在までに得られているRIA及びLOCA模擬試験結果について述べる。

Experimental analysis with RANNS code on boiling heat transfer from fuel rod surface to coolant water under reactivity-initiated accident conditions

宇田川 豊; 杉山 智之; 鈴木 元衛; 天谷 政樹

IAEA-TECDOC-CD-1775; Proceedings of Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents (CD-ROM), p.200 - 219, 2015/00

https://www-pub.iaea.org/MTCD/Publications/PDF/TE-1775_CD_web.pdf

In order to promote a better understanding of the temperature evolution of fuel rod under reactivity initiated accident (RIA) conditions, we have investigated the effects of coolant subcooling, flow velocity, pressure, and cladding pre-irradiation on the heat transfer from fuel rod surface to coolant water during RIA boiling transient, based on a computational analysis with the RANNS code on the transient data from RIA-simulating experiments in the NSRR. The analysis showed that the film boiling heat transfer coefficients during RIA boiling transient increase with coolant subcooling, flow velocity, and pressure as predicted by the model for stable film boiling. The estimated boiling heat transfer coefficients were significantly larger than those predicted by semi-empirical correlations for stable film boiling. The analysis also suggested that the heat transfers during both transition and film boiling phases are strongly enhanced by pre-irradiation of the cladding.

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

宇田川 豊; 杉山 智之; 天谷 政樹

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

軽水炉の反応度事故において、燃料ペレット温度の上昇に伴って生じる被覆管温度の上昇は、燃料の変形及び破損挙動を決定づける因子の一つとなる。冷却水の沸騰を伴う被覆管表面の伝熱挙動は、冷却水の圧力、サブクール度、流速等の影響を受けることから、これらのパラメータが反応度事故条件下の被覆管表面伝熱挙動に及ぼす影響を調べるため、大気圧から約16MPaの圧力、約10から80Kのサブクール度、静止水から約3m/sまでの流動水等、広範な冷却水条件にわたる反応度事故模擬実験が、原子炉安全性研究炉(NSRR)において実施された。また室温大気圧条件については、試験炉で予備照射された燃料についてもNSRR実験が行われ、データが取得されている。本報告は、反応度事故条件下における被覆管の伝熱挙動に関するデータ及び知見の取得を目的として行われたNSRR実験のうち、有効な被覆管温度データの取得に成功した実験ケースについて、これまでに公開されていないケースを含め体系的にとりまとめ、冷却水の圧力,サブクール,流速,予備照射の有無等諸パラメータの効果について整理したものである。

Stress biaxiality in high-burnup PWR fuel cladding under reactivity-initiated accident conditions

宇田川 豊; 杉山 智之; 鈴木 元衛; 永瀬 文久

Journal of Nuclear Science and Technology, 50(6), p.645 - 653, 2013/06

https://doi.org/10.1080/00223131.2013.787030

In order to promote a better understanding of failure mechanisms of high-burnup PWR fuels under reactivity-initiated accident (RIA) conditions, stress biaxiality in cladding has been estimated for the pellet-cladding mechanical interaction (PCMI) phase. The estimation was based on an analysis of the transient elongations of a pellet stack and a cladding tube measured in RIA-simulating experiments in the nuclear safety research reactor (NSRR) using the RANNS code. Stress biaxiality in the high-burnup PWR fuel cladding during the PCMI phase has been estimated to be 0.7-0.8 on average at the mid wall of the cladding. A comparison to fresh fuel test results and a sensitivity analysis showed that the effects of burnup and pulse width on cladding stress biaxiality are less than 10% for the investigated range. The present analysis also indicated that pellet-cladding friction is strong and that the cladding constraint on pellet stack elongation is significant irrespective of burnup.

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

杉山 智之; 宇田川 豊; 鈴木 元衛; 永瀬 文久

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

原子力機構は、反応度事故時の軽水炉燃料の挙動を明らかにするため、NSRRを用いたパルス照射試験を実施している。NSRR実験はRIA時のペレット被覆管機械的相互作用(PCMI)で生じる破損について燃焼度77GWd/tまでの燃料についてデータを提供してきた。特にPCMI破損限界は原子炉の安全審査において必要となる重要な情報である。しかし、NSRR実験では冷却材温度や出力パルス幅といった条件において実機で想定されるRIAとは異なる。したがって、NSRR実験データに基づいて実機条件下の破損限界を評価するためには、それらの条件の違いが破損限界に及ぼす影響を定量的に評価する必要がある。本論文では、一組の室温・高温実験で得られた実験データを示すとともに、このデータに基づいて行った冷却材温度と出力パルス幅の影響評価の方法について議論する。