Thermal hydraulic characteristics of helical coil once-through steam generator under ocean conditions

Owing to its advantages of high heat transfer efficiency and compactness, the helical coil once-through steam generator(HCOTSG) can be used in floating nuclear power plants and has been widely used in the design of small modular reactors. The helical tubular geometric structure of the HCOTSG allows heat transfer and local flow changes to occur under complex ocean conditions. In this study, theoretical models of ocean conditions are added to the RELAP5/MOD3.3 code and verified. Using the modified RELAP5 code, the thermal–hydraulic characteristics of the HCOTSG under ocean conditions are simulated. The results show that under rolling conditions, the flow oscillation amplitudes of the single liquid-phase, twophase flow, and single gas-phase regions are different. A circular change in the horizontal position of the helical tube causes the fluctuation of the parameters to change periodically. A phase difference of approximately 3.9 s at a flow rate of 23 kg/s is observed in the flow fluctuation along the axial direction. The driving force, period, and amplitude of rolling significantly affect the flow fluctuation in the HCOTSG. In natural circulation, the flow in the HCOTSG is complex, and the primary-side flow fluctuation can reduce the trough of the flow oscillation at the helical tube by approximately 24.3%.

Preliminary Design and Analysis of ITER In-Wall Shielding

ITER in-wall shielding （IIS） is situated between the doubled shells of the ITER Vacuum Vessel （IVV）. Its main functions are applied in shielding neutron, gamma-ray and toroidal field ripple reduction. The structure of IIS has been modelled according to the IVV design criteria which has been updated by the ITER team （IT）. Static analysis and thermal expansion analysis were performed for the structure. Thermal-hydraulic analysis verified the heat removal capability and resulting temperature, pressure, and velocity changes in the coolant flow. Consequently, our design work is possibly suitable as a reference for IT＇s updated or final design in its next step.

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal shock （PTS） events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States （US） rules： Operators may act after 10 min （Japanese rules） and 30 min （the US rules） after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module（TBM） program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R＆D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices（ITER relevant and DEMO）.The Indian Lead–Lithium Cooled Ceramic Breeder（LLCB） blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder（HCCB） is an alternative blanket concept that consists of lithium titanate（Li_2TiO_3） as ceramic breeder（CB） material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.