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This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry.

The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water chemistry of supercritical water used as a coolant in nuclear power reactors. It will also help readers to broaden their understanding of fundamental elements of light water cooled reactor technologies and the evolution of reactor concepts.



Chapter 1. Introduction and Overview

This chapter describes the characteristics of supercritical water, the concept of supercritical pressure water cooled reactors, the designs of a thermal spectrum reactor (Super LWR) and a fast reactor (Super FR), safety characteristics, and results of experiments on thermal hydraulics, materials and interactions with coolants. High breeding and transmutation studies are also summarized.
Yoshiaki Oka

Chapter 2. Reactor Design and Safety

Core designs and safety analyses of the Super LWR and Super FR are described in Sects. 2.1 and 2.2. The single-pass core of the Super LWR adopts the fuel assembly with large water rods having a thermal insulator. The upper core structure allows for simplified refueling procedures like light water reactors (LWRs). The single-pass core of a Super FR adopts the blanket fuel assembly with mixed oxide fuel rods in the lower part. Safety characteristics at abnormal transients, accidents and anticipated transients without scram (ATWS) of the Super LWR and Super FR are described. The total loss of flow accident and loss of coolant accidents (LOCAs) are important. Transient sub-channel analysis predicts a lower fuel cladding temperature for accidents and abnormal transients than the single-channel model for the Super FR where peaking is small. Transient subchannel analysis is described in Sect. 2.3.
New rod-type spacer was developed and is described in Sect. 2.4. Transmutation of long life radioactive fission products (LLFPs) is studied from the viewpoints of environmental risk and human risk. The high breeding core of the Super FR is developed with the tightly packed fuel rod fuel assembly. They are described in Sects. 2.5 and 2.6 respectively. Nuclear calculation of the fast and thermal neutron coupled core is described in Sect. 2.7.
The author of this chapter except Sect. 2.4 is Yoshiaki Oka with assistance of Qingjie Liu and Sutanto. Shinichi Morooka is the author of Sect. 2.4.
Yoshiaki Oka, Shinichi Morooka

Chapter 3. Thermal Hydraulics

This chapter describes thermal hydraulic studies.
Section 3.1 describes the heat transfer and fluid flow studies with surrogate fluids. It includes the heat transfer and pressure drop measurements in single tube, single rod, rod bundle geometries. Effect of grid spacer on heat transfer and critical heat flux at near critical-pressure are described. The results of the experiments on condensation, critical-flow and cross-flow between bundle channels are summarized. The author of sections from 3.1.1 to 3.1.4 and 3.1.7 is Hideo Mori. The sections of 3.1.5 and 3.1.6 are authored by Hideo Mori and Yoshinori Hamamoto.
Section 3.2 describes heat transfer experiments in single tube geometry with supercritical water for validation of the data of surrogate fluids. The author is Koichiro Ezato.
Section 3.3 describes computational fluid dynamics (CFD) analysis of the experiments. The authors are Takeharu Misawa and Kazuyuki Takase.
Hideo Mori, Yoshinori Hamamoto, Koichiro Ezato, Kazuyuki Takase, Takeharu Misawa

Chapter 4. Materials

This chapter deals with fuel cladding and thermal insulating materials. The current status regarding development of fuel cladding materials is described in Sect. 4.1. Zr-modified 15Cr20Ni austenitic alloy (1520Zr alloy) has been developed for fuel cladding application of the supercritical-pressure light-water cooled reactor by improving the austenitic alloy PNC1520 to provide the necessary high temperature strength and compatibility with high temperature water environments. Excellent creep strength was confirmed for the 1520Zr alloy. The oxidation property in supercritical water and SCC susceptibility in subcritical water of 15Cr20Ni based austenitic alloys have been preliminarily examined using PNC1520 alloy and the essential resistance to those degradation modes was demonstrated.
The oxidation kinetics of three types of 15Cr-20Ni austenitic stainless steels (1520 SSs) in supercritical water at 700 °C under 24 MPa are described in Sect. 4.2. The cladding tube-shaped 1520 SSs showed very low oxidation kinetics and no spalling. It was considered that the tube-shaped 1520 SSs show good performance for fuel cladding application in the SCWR from the viewpoint of oxidation kinetics. The high oxidation resistance of the tube is due to a protective Cr-rich oxide layer formation on the outside surface of the tubes. It was considered that Cr diffusion within the outside surface layer of the tubes is accelerated as a result of grain refinement and/or an increase of dislocation density due to a high degree of cold work. The authors of the Sects. 4.1 and 4.2 are Yutaka Watanabe and Hiroshi Abe.
Section 4.3 describes the thermal insulating materials that are used on the water rod walls. They are required because of the large temperature difference between the coolant inside and outside. Yttria-stabilized zirconia (YSZ) was developed for this purpose. Evaluations showed that 8 mol% YSZ with a density above 40 % is suitable for the thermal shielding material for the Super LWR and Super FR. Section 4.3 was written by Yoshiaki Oka based on the results of Kazuya Sasaki and Takayuki Terai.
Yutaka Watanabe, Hiroshi Abe, Yoshiaki Oka

Chapter 5. Material–Coolant Interactions

Basic understanding of material–coolant interactions in subcritical and supercritical water is of great importance for verifying the sustainability of supercritical water cooled reactors (SCWRs); these include such interactions as elution and corrosion of nuclear structural materials, and transportation and deposition processes along the cooling circuit. Although elution and corrosion properties in general light water reactors (at conditions from room temperature up to 300 °C) have been intensively investigated, those above 300 °C are still not well known. Thus, in the work described in this chapter, some structural materials such as type 304 and 316 stainless steels (SUS 304, SUS 316), Alloy 625 (Inconel 625) were used and model elution and corrosion experiments were conducted to obtain fundamental water chemistry data in subcritical and supercritical water (temperature range from 250 to 550 °C at 25 MPa, and in various atmospheres (deaerated and dissolved H2 or O2 conditions)).
In Sect. 5.1, an in situ measurement of the elution behavior in subcritical and supercritical water was conducted by employing a new supercritical water loop system combined with an elution detection system. Radioactivated specimens were used, and elution behavior was traced by measuring gamma-rays emitted from the eluted material (Co-60) collected at the outlet of an autoclave. As this technique allows detection of the eluted material with extremely high sensitivity, it is expected to quantify the amount of the elution. Development of the new method was discussed and the elution property at various temperatures and water chemistry conditions was described.
In Sect. 5.2, transportation and deposition processes along a cooling tube in heating and cooling stages were investigated by employing another new supercritical water loop system. The thickness of the oxide layer and its elemental composition, chemical composition of the oxide outer layer, and surface structure were observed by Auger electron spectroscopy, Raman spectroscopy and scanning electron microscopy measurements, respectively. Corrosion behavior of the specimens along the loop was discussed. It was determined that the corrosion behavior is temperature dependent, namely position dependent in supercritical cooled water reactors. The authors of the chapter are Yosuke Katsumura and Yusa Muroya.
Yosuke Katsumura, Yusa Muroya


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