Materials and Water Chemistry Research Group

Research on structural integrity of the reactor pressure vessels and core internals

 More than one-third of domestic nuclear power plants (NPPs) have already been in long-term operation for over 30 years. For safe utilization of these NPPs, evaluating the effects of aging degradation on the integrity of nuclear reactors is necessary. The objective of our research is to provide scientific and technical data as a regulatory basis for assessment of aging management and approval for the lifetime extension of light water reactors (LWRs). We are researching the degradation (irradiation embrittlement and stress corrosion cracking) of reactor component materials as well as water chemistry (radiolysis) of reactor coolant (Fig. 1) under high temperature and strong radiation.

Fig. 1 Schematic of an LWR.

(1) Irradiation embrittlement of reactor pressure vessel

 A reactor pressure vessel (RPV, Fig. 1, labeled ①) is one of the most important safety-related structural components of a nuclear reactor; it confines radioactive materials and high-temperature water. The RPV is made of a low-alloy steel with a thickness greater than 10 cm. The low-alloy steel becomes brittle at low temperature, and the embrittlement of the steel is promoted by neutron irradiation. The effect of irradiation embrittlement on the integrity of RPV steel is monitored by surveillance tests during plant operation, and then, a safety margin of the RPV fracture is confirmed. For long-term utilization of an RPV, it is necessary to predict the future degradation of the RPV accurately. In this research, conditions and the extent of brittleness of materials by exposure to radiation will be investigated using neutron-irradiated materials; the data necessary for the accurate prediction of the irradiation embrittlement of RPV steels will be obtained.

 It is known that microstructural changes, such as a formation of a solute atom cluster by neutron irradiation, cause material degradation. We carried out microstructure analysis of the RPV steels irradiated in the JMTR (Japan Materials Testing Reactor) using the three-dimensional atom-probe technique and have investigated the relation between the formation of the solute atom cluster and neutron fluence (Fig. 2). Furthermore, installation of machining apparatuses for the irradiated materials and the preparation for fracture toughness tests (Fig. 3) of RPV steels with used Charpy impact test specimens are progressed.

Fig. 2 Microstructural change of RPV steel and results of three-dimensional atom-probe analysis.
Fig. 3 Fracture toughness test with used Charpy impact test specimen.

(2) Stress corrosion cracking of reactor core internals

 In the integrity evaluation on stress corrosion cracking (SCC) of stainless steel components used in reactor core internals such as a core shroud (Fig. 1, labeled ②) of a boiling water reactor (BWR), it is important to evaluate crack growth rates accurately considering irradiation effects for proper inspection, repair, and replacement of the components. In this study, the behavior of irradiation-assisted stress corrosion cracking (IASCC) in stainless steels has been investigated. We have been performing tests to evaluate the crack growth rates of the irradiated stainless steels under simulated LWR conditions. We also have obtained data on a deformation and a corrosion (oxidation) of irradiated stainless steels (Fig. 4) to understand the IASCC phenomena through mechanistic investigation.

Fig. 4 Evaluation of effect of localized deformation in irradiated stainless steel on oxidation in high-temperature water.

(3) Water chemistry of reactor coolant

 To maintain the integrity of structural materials, the water chemistry of the primary coolant in LWRs (Fig. 1, labeled ③) is controlled by optimizing the dissolved hydrogen concentration to suppress oxidized species generated by water radiolysis. Water chemistry in the reactor core can be evaluated through a combination of theoretical analyses and in-situ electrochemical corrosion potential (ECP) measurements, supporting the validity of the theoretical analyses by ECP measurement. In this study, analytical methods to evaluate water chemistry and ECP of materials in coolant water in the reactor core and the highly-reliable and highly-durable ECP sensor for the water chemistry experiments are being developed. The codes for water radiolysis and ECP calculation have been verified by using the limited data obtained under the irradiation conditions (Fig. 5).

Fig. 5 Comparison between calculated and measured ECP values.

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