Research and Development of Advanced Nuclear System

2-4

Codification of Materials for Demonstration Fast Reactor in JSME Code－ Establishment of Standards for Material Strength and Elevated Temperature Design for 316FR and Modified 9Cr-1Mo Steels －1000

Mod. 9Cr-1Mo SteelSodium

600 ℃

CracksStressσ(MPa)250 ℃
T= 500 ℃Specimen

T= 550 ℃

10050Temp
Plate Forging(℃)

500
550
600
650○△□◇

T= 600 ℃●▲■◆

φ162.3

20045°

Liquid Penetrant Test

T= 650 ℃― : JSME Fast Reactor Code, 2012 Edition (Average)
1010100100010000

1000001000000

Creep rupture time tR (h)Configuration (mm)The evaluatednumberof cycles forし数
theき裂深さが10 mmに達した解析の繰返
longest crack
to試験結果reach （約1870回）
a 10 mm depth（約1530回）
は、を(approximately1530 cycles) wasこと
less
下回り、解析による評価が妥当であるがthan the observed value (approximately確認できた
1870 cycles), which demonstrated theappropriateness of the evaluation.

Fig.2-9 Creep rupture time of Mod. 9Cr-1Mo steel
Creep rupture data in which the rupture time exceeded 100000 h(approximately 11 years) were obtained. The slope of a graphof rupture time and stress was found to vary below and above
half of the 0.2% yield strength at temperature. It was inferredthat the longer the rupture time, the smaller the effects of plasticdeformation.

Fig.2-10 Example of structural test of Mod. 9Cr-1Mo steel
using sodium
A thick cylindrical specimen was subjected to a sodium flow
with a temperature that varied cyclically between 600 °C and
250 °C that generated thermal loads. It was demonstrated thatthe analysis was appropriate. By performing more tests under
different conditions, the adequacy of the margins in the design
standard was confirmed.

Future fast reactors need to be more reliable, and theirdesign has to be more flexible. The development of structural
materials is one of the keys to achieving this goal, andextensive research and development have been conducted.
For the Japanese demonstration fast reactor, which is in thedesign study phase, structural materials have been selectedthat are most suitable for meeting the requirements forcomponents, such as higher strength at elevated temperatures
or lower thermal expansion rates; 316FR steel was selected
for the reactor vessels and internals, and Modified 9Cr-1Mo(Mod.9Cr-1Mo) steel was chosen for most parts of the coolant
system, including the primary piping system. The 316FR steel
was developed in Japan by the Japan Atomic Energy Agency
and partners for fast reactor use. Mod.9Cr-1Mo steel wasoriginally developed by the Oak Ridge National Laboratory
in the United States, and extensive R&D programs wereconducted in Japan in order to apply it to fast reactors.
The temperature during steady-state operation of thedemonstration fast reactor will be approximately 550 °C, and adesign life of 60 years is planned. To establish methodologies
to achieve these goals, material data were obtained andevaluated under various conditions such as creep, in which
tensile stress is continuously applied at elevated temperatures;
fatigue, in which thermal stresses are repeatedly imposed;and creep-fatigue, in which creep and fatigue loadings aresuperimposed. For example, long-term creep tests, in whicha heated specimen fails by tensile stress imposed by a deadweight, were conducted; the maximum rupture time so far
obtained exceeds 100000 h (11.4 years). A database has been
established, and the data were analyzed from mechanicaland metallurgical viewpoints to formulate the properties.
In the formulation, the focus for 316FR steel was on taking
advantage of its superior properties under fast reactor
conditions, and for Mod.9Cr-1Mo steel, it was on representing
the creep properties, which vary depending on the stress levels(Fig.2-9). The creep strength of 316FR steel at 300000 h
and 550 °C, evaluated using the formulated properties, was
1.7 times higher than that of the conventional material
SUS304, and the increase was 2.3 times for Mod.9Cr-1Mo
steel compared with 2 1/4Cr-1Mo steel. Moreover, structural
tests in a sodium environment that is closer to that in fastreactor applications confirmed that the design method
developed for conventional materials is applicable to these
two materials with sufficient conservatism (Fig.2-10). The
above results led to a significant increase in the freedom ofstructural design for the demonstration fast reactor.
The above achievements are codified in the Japan Societyof Mechanical Engineers Code for Nuclear Power Generation
Facilities, Rules on Design and Construction for Nuclear PowerPlants, Section II Fast Reactor Standards, 2012 (in Japanese).

Reference
Asayama, T. et al., Creep-Fatigue Evaluation Methodologies and Related Issues for Japan Sodium Cooled Fast Reactor (JSFR), Procedia Engineering,vol.55, 2013, p.309-313.
JAEA R&D Review 201337