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Background Cardiac rehabilitation (CR) is an essential component of care for patients with cardiovascular diseases (CVD). We aimed to evaluate clinical outcomes in outpatients with CVD who did and did not complete a 5-month CR program. Methods Three hundred thirty-two outpatients with CVD who participated in a 5-month CR program and were followed-up for maximum 5 years were registered. We divided the patients into two groups: those who completed the CR program (success group, n = 175) and those who could not (non-success group, n = 157). Both long-term (5 years) and short-term (5 months) clinical outcomes were compared between the two groups. Results There were no significant differences in patient characteristics at baseline between the success and non-success groups. With regard to both long-term and short-term clinical outcomes, the rates of all-cause death and hospital admission in the success group were significantly lower than those in the non-success group by a Kaplan-Meier analysis. There was a significant difference in short-term CVD death and hospital admission between the groups, but not for long-term CVD death and hospital. In long-term period, all-cause death and hospital admission was independently associated with completion of the CR program in addition to the presence of peripheral artery disease and VE vs. VCO2 slope after adjusting for age, gender, body mass index, types of CVD and medications. Conclusions Completion of a 5-month CR program was associated with the prevention of all-cause death and hospital admission, but not CVD death and hospital admission in the long-term, which suggests that we need to reconsider this issue.

DRX is a very small integral type PWR (750 kW t ) for a scientific deep-sea research bathyscaph. The core having a small amount of steam is cooled by natural circulation and pressurized by self-pressurization. During operation of the bathyscaph in a deep sea or near the water surface, a ship inclination or ship motions will affect the reactor behavior. This paper describes the effect of a heeling or a heaving on the thermal hydraulic behavior of reactor system, which is analyzed by retran-02/grav improved so to simulate the effect of ship motions. The dynamics has a feature of nuclear power-natural circulation flow coupling under the condition of external forces. The analysis shows that ship inclination induces the core flow to decrease but reactor power recovers to the initial level without help of the reactor automatic control system. The heaving makes the core flow and the reactor power oscillate in phase with heaving, which are different from a density wave oscillation. Oscillation amplitudes of the flow and the power have peaks at the heaving period of 5 s. The peaks are due to resonance of the natural circulation flow and the heaving. An effective measure to suppress this oscillations due to heaving is to pressurize the primary loop by filling non-condensable gas. The density wave oscillation occurs when the reactor power increases over the rated power, and the boundary of its occurrence is analytically revealed. Under the condition of both density wave oscillation and heaving, the system shows to oscillate with the overlapped effect.

The basic concept of an innovative advanced marine reactor with a passive safety system, MRX (Marine Reactor X) has been established for the primary application to ship propulsion. The design goals of the reactor system, to be lightweight and compact, and to be enhanced in safety and reliability, are achieved with adoption of new technologies. The MRX is of the integral-type PWR with 100 MW of thermal output. Adoption of a water-filled containment makes the MRX extremely lightweight and compact. The total weight and volume of MRX is about 1600 tons and 1210 m3, which is half that of the first Japanese nuclear ship, ‘Mutsu’, reactor, although the reactor power of MRX is three times greater than that of the ‘Mutsu’ reactor. Numbers of active components in the reactor system are greatly reduced, compared with loop type PWRs, by adopting an integral type reactor and the passive system. Safety was evaluated by both experiments and analyses. Core damage occurrence frequency estimated by probability safety analysis (PSA) is of two orders smaller than those of existing PWRs. Feasibility study on economics is conducted by comparing the total operation costs of a nuclear container ship installing the MRXs with a diesel engine ship. The nuclear ship has the advantage for greater speed and larger amounts of cargo carried.

An experiment on transient single phase natural circulation in a closed loop was conducted to study the core cooling behaviour by natural circulation in the loop-type PWR for marine application. The system employed a loop that had two U-shaped, and three inverse U-shaped flow paths, and a small elevation difference between the heater and the cooler. The experiment was conducted at atmospheric pressure with the heater input held constant ( Q = 0.5 kW to 3.0 kW ), and the cooling water flow rate of the cooler constant during the transient following tripping of the pump. From the experiment, a flow stability map was obtained which correlated heater input versus elevation differences. Comparison between experimental and numerical results obtained by RETRAN-02 analyses yielded general agreement.

The residual mechanical strength and ductility of 304 stainless steel claddings were studied in tests of ring compression, ring tension and hardness. The tests were made after the specimens had been heated in steam and in argon gas to temperatures between 900° C and 1350° C. The effects of crystal grain growth and oxygen absorption in the metal on the residual properties were also studied. The experiments showed that the claddings retained their ductility well at temperatures below 1300° C and when the oxide scale thickness was below 40% of the initial wall thickness. The claddings maintained a constant level of residual strength when heated at temperatures over 1000° C or when the oxide thickness was over five percent. These results were used to support our proposals for limits on stainless steel cladding damage in a LOCA.

An experiment on oxidation of 304 stainless steel was performed in steam between 900°C and 1350°C, using the spare cladding of the reactor of the nuclear-powered ship Mutsu. The temperature range was appropriate for a postulated loss of coolant accident (LOCA) analysis of a LWR. The oxidation kinetics were found to obey the parabolic law during the first period of 8 min. After the first period, the parabolic reaction rate constant decreased in the case of heating temperatures between 1100°C and 1250°C. At 1250°C, especially, a marked decrease was observed in the oxide scale-forming kinetics when the surface treated initially by mechanical polishing and given a residual stress. This enhanced oxidation resistance was attributed to the presence of a chromium-enriched layer which was detected by use of an X-ray microanalyzer. The oxidation kinetics equation obtained for the first 8 min is applicable to the model calculation of a hypothetical LOCA in a LWR, employing 304 stainless steel cladding.