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A highly reliable control rod drive mechanism driven by an electric motor installed inside the reactor vessel (INV- CRDM) for a very small reactor has been designed. The INV-CRDM contributes to the compactness and simplicity of the reactor system, and can eliminate the possibility of a rod ejection accident. In the design, a new type of latch mechanism using an electromagnetic force to directly connect both of the shafts, one of which was the motor driven shaft and the other the control rod driving shaft, was applied so as to make the INV-CRDM very compact. The cable supplying current remained stationary, even when both of the shafts was moving. The required functions of the latch mechanism are to maintain an adequate latching force for the control rod shaft to move within a stroke of 370 mm, and to release the shafts in a shorter time than 0.2 s after a scram signal is received. A functional test with a model that approximately simulated the design was conducted to test the latching force and de-latching...

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 collective dose to the Japanese population has been estimated from a hypothetical accident of a nuclear submarine if it sinks in an offshore region around Japan. A computer code system DSOCEAN has been used for assessing the collective dose due to radionuclides released to the ocean from a sunken nuclear submarine. The estimated collective effective dose commitment from all of the radionuclides released after the break of the fuel pellets is estimated to be 2.5 x104 man-Sv. The contribution of 241Am to the total collective effective dose commitment is the highest, followed by 137Cs, 238Pu, 240Pu, 239Pu and 241 Pu. The maximum of the estimated collective effective dose by the annual intake of marine products after radionuclide releases for one year is approximately 0.3% of the annual average dose by the natural radiation that is reported by UNSCEAR.

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water—310°C and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch the driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been developed to be capable of working under the high temperature water for a long period. Under condition ...

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.

Compactness and light weight are of importance for a reactor plant used for ship propulsion. An advanced marine reactor, MRX, which has been designed by Japan Atomic Energy Research Institute, has accomplished a passively safe, compact and light weight design by adoption of an integral PWR and a water filled containment vessel. Dose equivalent rates in all areas inside and outside a ship are required to be reduced below design criteria without any bulk shield outside the containment vessel. In this paper, a shielding design for installation of the steam generator in the vicinity of the reactor core to achieve a compact plant by reducing fast neutrons in the steam generator and to enable the engine room, where pipes of the secondary loop are installed, to be classified as surveillance area, not controlled area, is described. The main purpose to classify the engine room as the surveillance area is the improvement of working efficiency of crew. To accomplish the design goal, a relatively thick core barrel an...

A simple method is presented to calculate radiation streaming through cylindrical ducts in a concrete shield. It includes neutron and gamma-ray calculations and can be applied to the problems of practical slant incidence. The method is essentially one of summing the axial dose rates from the different neutron and gamma components which are functions of energy and angle of incident radiation, duct diameter, bend angle, etc. Axial dose rates are calculated by the MCNP Monte Carlo code for straight and bent ducts having diameters of 5,10,20,30 and 40 cm. Radiation sources are plane mono-directional fission neutrons, intermediate neutrons, thermal neutrons and 1.25-MeV gamma rays. Incident angles are 0, 15, 30, 45 and 60 deg for straight ducts. For bent ducts, an isotropic plane source is placed inside the duct mouth to reduce computation time. Bend angles are 30,45,60 and 90 deg. Reliability of the Monte Carlo calculations are assessed by comparison with a 60Co experiment and calculations by the one-dimensio...

Japan Atomic Energy Research Institute has studied a highly compact reactor DRX for deep-sea research. The DRX has no purification system to achieve compact and light weight design by simplification of the system. The DRX is designed to operate for one month without purification of the primary coolant water. To quantitatively evaluate the validity of reactor operation without a purification system, a computer code CTAM-II has been developed to calculate accumulation of the activated corrosion products during and after reactor operation. The code is an improved and modified version of CTAM, which was developed for the shield modification project of the nuclear ship Mutsu. Validity of CTAM-II and parameters used in the code has been confirmed by comparison of the calculated data and experimental ones for measurement of the concentrations of radioactive materials in the primary coolant water. Estimation of buildup of the corrosion products for DRX using CTAM-II has been carried out and shielding calculations...

This paper describes the experiments of the first Japanese nuclear ship ‘Mutsu’, to investigate the effects of sea wave on the thermal hydraulics of marine reactor system while cruising through various sea conditions. The experimental data were analyzed in time-domain by RETRAN-02/GRAV code. This code was modified so as to simulate the ship motion effect on reactor thermal hydraulics. The data were also analyzed in frequency domain by Blackman Turkey method for the calculation of the spectrum and response function. The experiments involving ship maneuvering were performed by cruising on different wave heights, as well as wave directions in the northern Pacific ocean. From the experiments, vertical acceleration due to ship motion was found to induce direct variation of water levels in the SGs and the pressurizer. The water level variations were largest in the head wave, but smallest in the following wave. On the other hand, the following wave caused greater variation of the reactor power when the feed back...

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.

A number of operational transient analyses of the nuclear-powered ship Mutsu have been performed in response to Japanese nuclear safety regulatory concerns. The RETRAN and COBRA-IV computer codes w...

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.