Collection of strength characteristic data used for analysis evaluation in reactor pressure vessel and in-core structures in accident

Shimomura, Kenta; Yamashita, Takuya; Nagae, Yuji

JAEA-Data/Code 2022-012, 270 Pages, 2023/03

JAEA-Data-Code-2022-012.pdf:38.25MB

In a light water reactor, which is a commercial nuclear power plant, a severe accident such as loss of cooling function in the reactor pressure vessel (RPV) and exposure of fuel rods due to a drop in the water level in the reactor can occur when a trouble like loss of all AC power occurs. In the event of such a severe accident, the RPV may be damaged due to in-vessel conditions (temperature, molten materials, etc.) and leakage of radioactive materials from the reactor may occur. Verification and estimation of the process of RPV damage, molten fuel debris spillage and expansion, etc. during accident progression will provide important information for decommissioning work. Possible causes of RPV failure include failure due to loads and restraints applied to the RPV substructure (mechanical failure), failure due to the current eutectic state of low-melting metals and high-melting oxides with the RPV bottom members (failure due to inter-material reactions), and failure near the melting point of the structural members at the RPV bottom. Among the failure factors, mechanical failure is verified by numerical analysis (thermal hydraulics and structural analysis). When conducting such a numerical analysis, the heat transfer properties (thermal conductivity, specific heat, density) and material properties (thermal conductivity, Young's modulus, Poisson's ratio, tensile, creep) of the materials (zirconium, boron carbide, stainless steel, nickel-based alloy, low alloy steel, etc.) constituting the RPV and in-core structures to near the melting point are required to evaluate the creep failure of the RPV. In this document, we compiled data on the properties of base materials up to the melting point of each material constituting the RPV and in-core structures, based on published literature. In addition, because welds exist in the RPV and in-core structures, the data on welds are also included in this report, although they are limited.

Evaluation of multiaxial low cycle creep-fatigue life for Mod.9Cr-1Mo steel under non-proportional loading

Nakayama, Yuta*; Ogawa, Fumio*; Hiyoshi, Noritake*; Hashidate, Ryuta; Wakai, Takashi; Ito, Takamoto*

ISIJ International, 61(8), p.2299 - 2304, 2021/08

https://doi.org/10.2355/isijinternational.ISIJINT-2020-780

This study discusses the creep-fatigue strength for Mod.9Cr-1Mo steel at a high temperature under multiaxial loading. A low-cycle fatigue tests in various strain waveforms were performed with a hollow cylindrical specimen. The low cycle fatigue test was conducted under a proportional loading with a fixed axial strain and a non-proportional loading with a 90-degree phase difference between axial and shear strains. The low cycle fatigue tests at different strain rates and the creep-fatigue tests at different holding times were also conducted to discuss the effects of stress relaxation and strain holding on the failure life. In this study, two types of multiaxial creep-fatigue life evaluation methods were proposed: the first method is to calculate the strain range using Manson's universal slope method with considering a non-proportional loading factor and creep damage; the second method is to calculate the fatigue damage by considering the non-proportional loading factor using the linear damage law and to calculate the creep damage from the improved ductility exhaustion law. The accuracy of the evaluation methods is much better than that of the methods used in the evaluation of actual machines such as time fraction rule.

Relation between intergranular stress of austenite and martensitic transformation in TRIP steels revealed by neutron diffraction

Harjo, S.; Kawasaki, Takuro; Tsuchida, Noriyuki*; Morooka, Satoshi; Gong, W.*

ISIJ International, 61(2), p.648 - 656, 2021/02

https://doi.org/10.2355/isijinternational.ISIJINT-2020-503

The Effect of base irradiation on failure behaviors of UO and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

AA2019-0372.pdf:0.81MB

https://doi.org/10.1016/j.anucene.2019.107268

Creep deformation analysis of a pipe specimen based on creep damage evaluation method

Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07

It has become more important to develop methods for evaluating failure behavior of the nuclear components under severe conditions. We are researching on prediction methods of creep deformation and failure behavior of the nuclear components under elevated temperature conditions based on finite element analysis. In this study, as a part of a project called COSSAL, we performed failure analysis of a large scale pipe experiment to validate our prediction methods based on a creep damage evaluation method. We conclude that creep constitutive law that consider material damage can provide the highest accurate analysis.

Cladding tube burst experiment assumed MA fuel pin for transmutation physics experimental facility

Sugawara, Takanori; Tsujimoto, Kazufumi

JAEA-Research 2017-011, 35 Pages, 2017/10

JAEA-Research-2017-011.pdf:4.88MB

The construction of Transmutation Physics Experimental Facility (TEF-P) is planned in the J-PARC project. TEF-P is a critical assembly and it will treat minor actinide (MA) fuel in the experiment. The temperature when the air cooling for the TEF-P core would stop was estimated but there were no data to evaluate the soundness of the MA fuel pin. To set a tentative limit temperature for the TEF-P core, cladding tube burst experiment was performed. As the result, the cladding tube burst occurred at 660C as the severest case. Through these results and the estimation of creep rupture time, the tentative limit temperature for the TEF-P core was set to 600C.

Creep-fatigue tests of double-end notched bar made of Mod.9Cr-1Mo steel

Shimomura, Kenta; Kato, Shoichi; Wakai, Takashi; Ando, Masanori; Hirose, Yuichi*; Sato, Kenichiro*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

This paper describes experimental and analytical works to confirm that the design standard for SFR components sufficiently covers possible failure mechanisms. Creep-fatigue damage evaluation method in JSME design standard for SFR components has been constructed based on experiments and/or numerical analyses of conventional austenitic stainless steels, such as 304SS. Since the material characteristics of Mod.9Cr-1Mo steel are substantially different from those of austenitic stainless steels, it is required to verify the applicability of the design standards to the SFR components made of Mod.9Cr-1Mo steel. A series of uni-axial creep-fatigue tests were conducted using double-ended notch bar specimens made of Mod.9Cr-1Mo steel under displacement controlled condition with 30 minute holding. The curvature radii of the specimens were 1.6mm, 11.2mm and 40.0mm. The specimen having 1.6mm notch and 11.2mm notch failed from outer surface but the specimen having 40.0mm notch showed obvious internal crack nucleation. In addition, though total duration time of the creep-fatigue test was only 2,000 hours, a lot of creep voids and inter granular crack growth were observed. To clarify the cause of such peculiar failure, some additional experiments were performed, as well as some numerical analyses. We could point out that such a peculiar failure aspect might result from corresponding stress distribution in the cross section. As a result of a series of investigations, possible causes of such peculiar failure could be narrowed down. A future investigation plan was proposed to clarify the most significant cause.

Damage evaluation techniques for FBR and LWR structural materials based on magnetic and corrosion properties along grain boundaries

Hoshiya, Taiji*; Takaya, Shigeru*; Ueno, Fumiyoshi; Nemoto, Yoshiyuki; Nagae, Yuji*; Miwa, Yukio; Abe, Yasuhiro*; Omi, Masao; Tsukada, Takashi; Aoto, Kazumi*

Transactions of the Materials Research Society of Japan, 29(4), p.1687 - 1690, 2004/06

JAERI and JNC have begun the cooperative research of evaluation techniques of structural material degradation in FBR and LWR, which based on magnetic and corrosion properties along grain boundaries. Magnetic method has been proposed as the one of the non-destructive detection techniques on the early stage of creep-damage before crack initiation for aged structural materials of FBRs. The effects of applied stress on natural magnetization were investigated on paramagnetic stainless steels having creep-damages. On the other hand, corrosion properties and magneto-optical characteristics of ion-irradiated stainless steels in the vicinity of grain boundaries were estimated by AFM and Kerr effect microscope, respectively. These degradations were induced by changes in characteristics in the vicinity of grain boundaries. It is found that the initial level of progressing process of damage can detect changes in magnetic and corrosion properties along grain boundaries of aged and degraded nuclear plants structural materials.

Analysis of pressure- and temperature- induced steam generator tube rupture during PWR severe accident initiated from station blackout

Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00

Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.

Dosimetry plan at the first irradiation test in the HTTR

Shibata, Taiju; Kikuchi, Takayuki; Shimakawa, Satoshi

Reactor Dosimetry in the 21st Century, p.211 - 218, 2003/00

The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan with a maximum power of 30 MW. The construction of it was completed successfully in March 2002. The HTTR aims to perform irradiation studies at its very wide irradiation spaces at high temperatures. Although the creep behavior of materials is measured by the large standard size specimens at out-of-pile, small size ones are generally used for in-pile creep tests because of the irradiation capability of reactors. The I-I type irradiation equipment, the first rig for the HTTR, is to be used for the in-pile creep test on a stainless steel with the standard specimens. The rig can give big tensile loads of about 9.8 kN on them. The temperatures of 550 and 600C and the fast neutron fluence of 1.210n/m are the targets of the test. Prior to the in-pile creep test, the in-core irradiation properties at the irradiation region are to be obtained by the rig as the first irradiation test. This paper describes the dosimetry plan at the first irradiation test and the subsequent data assessment procedure.