104 results about "Fast-neutron reactor" patented technology

A new transmutation assembly permits an efficient transmutation of a long-lived radioactive material (long-lived FP nuclides such as technetium-99 or iodine-129) which was produced in the nuclear reactor. Wire-type members of a long-lived radioactive material comprised of metals, alloys or compounds including long-lived FP nuclides are surrounded by a moderator material and installed in cladding tubes to form FP pins. The FP pins, and nothing else, are housed in a wrapper tube to form a transmutation assembly. The wire-type members can be replaced by thin ring-type members. The transmutation assemblies can be selectively and at least partly loaded into a core region, a blanket region or a shield region of a reactor core in a fast reactor. From a viewpoint of reducing the influence on the reactor core characteristics, it is optimal to load the transmutation assemblies into the blanket region.

A fast reactor having a reflector control system is provided which decreases the change in reactivity of the reactor core with time without performing control of a reflector lifting speed and that of a water flow rate. The above fast reactor has a liquid metal coolant, a reactor core immersed therein, and a neutron reflector which is provided outside the reactor core and which is moved in a vertical direction for adjusting leakage of neutrons therefrom for controlling the reactivity of the reactor core. The neutron reflector described above is gradually moved in an upward direction with the change in reactivity caused by fuel burn-up, and at least a part of a lower region of the neutron reflector is a high reflection region having a high neutron reflection ability as compared to that of the other region. The high reflection region is located from the bottom to a place between one fourth and one half of the height of the neutron reflector from the bottom end thereof.

The invention discloses a method for obtaining three-dimensional neutron flux density distribution in a reactor core transient process of a fast neutron reactor. A polygonal prism grid is adopted for the anisotropic neutron flux density of the reactor core of the fast neutron reactor to carry out fully-three-dimensional transportation space dispersion, a semi-spherical surface of a sixty-degree region is taken as a unit to carry out alternate scanning, and iterative format degradation in an angle parallel process is weakened; and the characteristics of the weak local effect and the global space coupling of the neutron flux density of the reactor core of the fast neutron reactor are considered, an estimated correction quasi-static strategy is adopted for neutron flux density change in a transient process to carry out time dispersion, ingredients of different change rates along with time in the neutron flux density can be decomposed, separated solution is carried out on different time scales, meanwhile, nonlinear iteration among above ingredients is avoided, and calculation efficiency is improved. The method for obtaining the three-dimensional neutron flux density distribution in the reactor core transient process of the fast neutron reactor has the advantages of being high in calculation accuracy and reasonable in calculated amount.

The invention discloses a method and a device for fast multiplicating and transforming nuclear fuel with high inherent security and low cost, and the method and the device adopt 'moving solid nuclear fuel' which can flow and is similar to fluid, can be applied in a fast neutron reactor, and produce more nuclear fuel with less fissile fuel initial installation quantity and lower cost in a shorter time, thus satisfying the nuclear fuel demand at the current world nuclear power big development stage. The method and the device for fast multiplicating and transforming nuclear fuel are real nuclear reactors with inherent security, and can thoroughly eliminate the serious accident of massive radioactive release caused by element melting of the nuclear fuel which still possibly occurs in the current nuclear power station.

The invention relates to a high-temperature supercritical nuclear reactor. A spherical solid is used as a coolant of the nuclear reactor, which is different from common nuclear reactors that adopt light water, heavy water, gas and liquid sodium. The spherical solid coolant can be graphite spheres coated with silicon carbide, stainless steel spheres, graphite spheres coated with stainless steel and the like. The spherical solid has the characteristic of good rolling ability, so that the spherical solid can roll to a steam generator under the action of gravity after heating in the reactor, and the purpose of transferring heat from the reactor to the steam generator is achieved. The reactor is applicable to both thermal neutron reactor and fast neutron reactor, and natural uranium, low-enriched uranium, plutonium and thorium can be used as nuclear fuel. The temperature of high-temperature steam / water outputted by the steam generator can reach 900 DEG C, the same parameters to a thermal power station can be reached, the requirement of supercritical even ultra-supercritical can be satisfied, and the high-temperature supercritical nuclear reactor has high safety and reliability.

A fast reactor having a reactivity control reflector has a reactor vessel in which a coolant is accommodated, a reactor core which is installed in the reactor vessel and dipped with the coolant, and a reflector installed outside of the reactor core so as to be movable in a vertical direction for controlling the reactivity of the reactor core. The reflector of the fast reactor has a lower neutron reflecting portion having a neutron reflection capability higher than that of the coolant and an upper cavity portion located above the neutron reflecting portion and having a neutron reflection capability lower than that of the coolant. The cavity portion is composed of a plurality of cylindrical hermetically-sealed vessels.

Disclosed herein are a nuclear fuel rod for fast reactors, which includes an oxide coating layer formed on the inner surface of a cladding, and a manufacturing method thereof. The nuclear fuel rod for fast reactors, which includes the oxide coating layer formed on the inner surface of the cladding, can increase the maximum permissible burnup and maximum permissible temperature of the metallic fuel slug for fast reactors so as to prolong the its lifecycle in the fast reactors, thus increasing economic efficiency. Also, the fuel rod is manufactured in a simpler manner compared to the existing method, in which a metal liner is formed, and the disclosed method enables the cladding of the fuel rod to be manufactured in an easy and cost-effective way.

Provided is a high-accuracy fast neutron reactor assembly few-group cross section obtaining method. The fine-group cross section of a fast neutron reactor assembly is calculated in the mode that point cross section data and multi-group data are combined, online calculation of an elastic scattering matrix is performed, a neutron transport equation is solved by using the fine-group cross section to obtain fine-group neutron-flux density, energy groups are merged on the basis, and accordingly few-group parameters of the fast reactor assembly are obtained. Due to the fact that the method directly uses the point cross section data, the elastic scattering resonance effect of a medium mass nuclide and the resonance interference effects of all nuclides are accurately considered. The neutron-flux density of a real system is adopted in the energy group merging process, and merging results are more accurate. The overall errors of fast reactor assembly few-group parameters calculated by adopting the method are within 1% compared with a reference value, and the method has higher accuracy.

A liquid-metal cooled fast reactor core having a nuclear fuel assembly constituted of nuclear fuel rods with varying cladding thicknesses in reactor core regions, in which: the nuclear fuel assembly (1) of a liquid-metal cooled fast reactor includes nuclear fuel assemblies (1a, 1b and 1c) in inner, middle and outer reactor core regions, respectively, and is installed in a hexagonal duct (3) with nuclear fuel materials (2-2a, 2-2b and 2-2c) surrounded by respective claddings (2-1a, 2-1b and 2-1c), and the claddings (2-1a, 2-1b and 2-1c) of a nuclear fuel rod (2a) in the inner reactor core region, a nuclear fuel rod (2b) in the middle reactor core region and a nuclear fuel rod (2c) in the outer reactor core region are formed at different thicknesses. The reactor core can flatten power distribution using a single-enrichment nuclear fuel in the liquid-metal cooled fast reactor.

The invention discloses a metallic fast reactor fuel element irradiation test device. The device can be used for carrying out neutron irradiation test on metallic fast reactor fuel in a thermal spectrum research reactor and has the functions of neutron spectrum hardening, metallic fuel irradiation temperature control, fuel element heat release power measurement and the like. The device comprises aprotection tube, a cadmium tube, an irradiation test piece, a related gas pipeline, a thermocouple and the like. According to the device, a cadmium metal layer is adopted to absorb thermal neutrons to reduce the thermal neutron flux rate in the device, so that the neutron energy spectrum of a test fuel element is close to that of a fast neutron reactor; and the temperature difference is established between air gaps and a liquid metal layer outside a fuel rod, and the gas components in the air gaps in each irradiation part are independently adjusted, so that the temperature of fuel core bodiesin different irradiation parts are controlled within a required temperature range.

A nuclear fuel rod for a fast reactor includes tubular fuel materials comprising a hollow portion formed therein, a tubular inner pipe inserted into the hollow portion of the tubular fuel materials to prevent collapse of the tubular fuel materials due to combustion of nuclear fuel, a tubular cladding pipe which surrounds the tubular fuel materials, and a liquid metal, or He gas or vacuum applied in a gap between the tubular fuel materials and the tubular cladding pipe.

A fast reactor having a reflector control system is provided which decreases the change in reactivity of the reactor core with time without performing control of a reflector lifting speed and that of a water flow rate. The above fast reactor has a liquid metal coolant, a reactor core immersed therein, and a neutron reflector which is provided outside the reactor core and which is moved in a vertical direction for adjusting leakage of neutrons therefrom for controlling the reactivity of the reactor core. The neutron reflector described above is gradually moved in an upward direction with the change in reactivity caused by fuel burn-up, and at least a part of a lower region of the neutron reflector is a high reflection region having a high neutron reflection ability as compared to that of the other region. The high reflection region is located from the bottom to a place between one fourth and one half of the height of the neutron reflector from the bottom end thereof.

The invention relates to a hybrid computation method for acquiring few-group cross section parameters of a fast neutron reactor. The hybrid computation method comprises the following steps: combining a Monte Carlo method with a certain theory method, finely considering the resonance effect of a fast reactor assembly by utilizing the Monte Carlo method, calculating precise multi-group microcosmic total cross section, fission cross section and elastic scattering cross section of each nuclide, solving various orders of elastic scattering cross sections and various orders of neutron flux moments of the fast reactor assembly by utilizing the certain theory method, and aggregating the multi-group cross section into the few-group cross section. The hybrid computation method provided by the invention is strong in universality and wide in application range, a high-precision few-group cross section of the fast reactor assembly can be produced, and precise and reliable cross section parameters are provided for a task of designing a nucleus of a reactor core.

The invention discloses a fast neutron reactor and alkali metal thermoelectric converter integrated device for an exoplanet, which belongs to the technical field of nuclear power resources. According to the complete device disclosed by the invention, an alkali metal fast neutron reactor and an alkali metal thermoelectric converter are combined; the fast neutron reactor and alkali metal thermoelectric converter integrated device is formed by modularized equipment consisting of two barrels which are mutually inserted, has a simple structure, does not need to be connected by pipelines, depends on the natural circulation and does not require a pump or various valves and switches; the reliability of equipment is greatly improved; and various accidents can be avoided. A thermoswitch heat pipe is used as a waste heat discharge system and cannot be started under the normal working condition, so that the heat loss of a reactor core is avoided; and the heat pipe has excellent thermal conductivity, so that residual heat of the reactor core is timely conducted and the safety of the reactor core is ensured. The device can be placed on the moon or other exoplanets to provide daily domestic electricity or electricity for large-scale equipment and can also be used for providing continuous power for sparsely populated areas.

The invention discloses a three-dimensional distribution measurement system for neutrons in a reactor based on measurement data outside the reactor. A reactor power inversion subsystem is used for obtaining neutron measurement values in areas around the reactor, and the out-of-reactor detector subsystem changes the number, positions and detection directions of detectors according to real-time monitoring needs; an online spectrum unfolding subsystem receives the measurement data from the out-of-reactor detector subsystem, calculates neutron energy spectra at the detector positions by an onlinespectrum unfolding method, and then transmits energy spectrum data at each detection position to the reactor power inversion subsystem; and the reactor power inversion subsystem calculates the neutronenergy spectra at different spatial positions in the reactor through the energy spectrum data at the out-of-reactor detectors. The three-dimensional distribution measurement system can rely on the out-of-reactor detectors only to quickly obtain the distribution of the neutrons in the reactor in different space and energy zones, and can be used in monitoring, diagnosis and analysis of advanced reactors such as critical driven reactors, fusion driven reactors and fast neutron reactors.

The invention relates to a grid type control drum device of a small fast neutron reactor. The grid type control drum device comprises an inner reflecting layer, a grid type control drum and an outer reflecting layer which sequentially sleeve the periphery of a reactor core from inside to outside, and a driving motor for driving the grid type control drum; the inner reflecting layer is in a cylindrical grid shape and is fixed in position; the outer reflecting layer is in a complete cylinder shape and is fixed in position; and the grid type control drum is in a cylindrical grid shape and can bedriven by the driving motor to rotate by a certain angle so as to adjust the overlapping area between the absorption grids and the reflection grid of the inner reflecting layer and change the number of leaked neutrons reflected back to the reactor core, so reactivity control is achieved. The device disclosed by the invention is small in occupied axial space, and meets the miniaturization design requirement; the absorption grids have no axial displacement, so that uneven axial power distribution of the reactor core cannot be caused; and the absorption grids are in central symmetry about the central axis of the reactor core, have good adaptability to material expansion caused by radial temperature gradient, and are not easy to block due to deformation.

Provided are a nuclear fuel rod for fast reactors that includes a metallic fuel slug coated with a protective coating layer and a fabrication method thereof. The nuclear fuel rod for fast reactors that includes a surface treated metallic fuel slug and a cladding tube according to the present invention has an excellent effect of stabilizing components of the metallic fuel slug and fission products or impurities, because the interdiffusion between the metallic fuel slug and the cladding tube does not occur. Also, since the uniform coating on the surface of the metallic fuel slug may be facilitated and fabrication costs may be significantly reduced in comparison to a typical technique of using a functional material for preventing the interdiffusion at an inner surface of the cladding tube, it may be suitable for fabricating the nuclear fuel rod for fast reactors.

The invention belongs to the technical field of nuclear power plant design, and relates to a fast reactor nuclear island main plant group arrangement structure. The fast reactor nuclear island main plant group arrangement structure comprises a reactor plant located in the center and further comprises a steam generator plant, a first electric plant, a second electric plant, a spent fuel plant, an operation service plant, a corridor and a steam turbine plant which are located on the periphery; the steam generator plant and the spent fuel plant are arranged in the mutually symmetrical direction;the first electrical plant and the second electrical plant are arranged in the mutually symmetrical direction; the operation service plant and the steam generator plant are connected with the first electric plant or the second electric plant; and the corridor is located on the periphery of the steam generator plant and connected with the steam generator plant, and the other end of the corridor isconnected with the steam turbine plant. By utilizing the fast reactor nuclear island main plant group arrangement structure, the requirements of design and operation modes of a fast reactor system canbe efficiently and reasonably met, and safety and economical efficiency of a nuclear island main plant group are guaranteed from the angle of arrangement design, so that the technical and safety requirements of the four-generation nuclear energy system are met.

The invention discloses a method for calculating an axial swelling effect of a fast neutron reactor assembly. The method comprises the following steps of 1, separating a transport computation grid from a burnup computation grid, wherein the transport computation grid is kept unchanged in the whole computation, and the burnup computation grid keeps growth of a same proportion along with axial swelling of a fuel assembly; 2, based on a variational nodal method, establishing a weak form of a neutron transport equation solution, wherein a section in a node is a function related to a spatial position, so that appearance of various materials is allowed in a transport computation node; 3, expanding a flux by utilizing a spherical harmonic function and a spatial orthogonal polynomial, and solving a flux expansion moment by adopting a response matrix method so as to obtain flux distribution of each node; and 4, calculating a burnup level of the node according to average power of a burnup node, obtaining a homogenized section of the burnup node according to burnup interpolation, and adjusting the homogenized section in the burnup node according to axial elongation of the assembly to obtain a required homogenized section in the computation node.

The invention discloses a method for establishing a fast neutron reactor fuel assembly grid model under a flow blocking condition. The method comprises the following steps: establishing a three-dimensional geometric model of a fast neutron reactor fuel assembly according to real design parameters; establishing a fluid domain model of the unblocked fast neutron reactor fuel assembly; carrying out unstructured tetrahedral mesh generation on the fluid domain model of the unblocked fast neutron reactor fuel assembly by using mesh generation software; performing grid marking operation on the flow blocking position according to calculation requirements, and adding a resistance source item which is three orders of magnitudes larger than the resistance source item in the other two directions in the flowing direction of the coolant. According to the method, the position, shape and size of the blocked area can be flexibly adjusted, and repeated geometric modeling work is avoided.

The invention belongs to fast fission reactor technology field and discloses a design transient determination method for sodium-cooled fast reactor nuclear power station cooling agent system and member. The method comprises three steps: defining design standard of cooling agent system and member, determining classification rule of design transient, determining design transient status and circulation times. The invention adopts American mechanical engineering boiler and pressure container standard (simply ASME) and France RCC-M standard. The design transient status of sodium-cooled fast reactor is classified to five conditions: normal operation condition, middle frequency accident condition, sparse accident condition, limit accident condition, experiment condition. The circulation times of a condition in service life of a reactor is target value of transient impact to be experienced by cooling agent system and member in service life of a reactor. The method satisfies design demand for sodium-cooled fast reactor.

The invention relates to nuclear engineering and can be used in the neutron irradiation of different materials (targets) to produce radionuclides. The essence of the invention lies in a method for producing radioactive isotopes in a fast neutron reactor which includes the following steps. Targets for the production of radioisotopes are positioned between sleeves and rods in an irradiation assembly, the rods being made of a neutron-moderating material, and the irradiation assembly is placed in a side shield of a fast neutron reactor. The irradiation assembly is surrounded with assemblies whichdo not contain nuclear fuel. In the irradiation assembly fast neutrons are passed through the neutron-moderating material, and the moderated neutrons are then passed through the material to be irradiated (the target) in the irradiation assembly. The characterizing features of the present invention are that radioisotopes are produced simultaneously in the irradiation assembly and the surrounding assemblies, in which the steel content is not greater than 50%. The targets in the irradiation assembly have an absorption cross section of more than 1 barn at a neutron energy of less than 0.1 MeV. Thetargets in the assemblies surrounding the irradiation assembly have a neutron absorption cross section of less than 1 barn at a neutron energy of more than 0.1 MeV.

The invention belongs to the technical field of shielding of fast-neutron reactors, and particularly relates to design of a shielding assembly structure with high-efficiency shielding properties. On the basis of maintaining the hexagonal outer sleeve, the inside of the shielding assembly is filled with shielding element balls from bottom to top, thereby simplifying the original design of the end plugs, cavity, holddown spring and wrapping wire of the shielding element bar, and enhancing the neutron and gamma-ray radiation shielding properties of the shielding assembly. The two-layer ceramic material design of the shielding element balls ensures the completeness and reliability of the shadowing assembly in the operation period; the loose pyrolytic carbon ceramic material of the inner layer provides storage space for the radiation activation product, absorbs the radiation-induced swelling of the boron carbide shielding material, and buffers the stress caused by the temperature and radiation; and the compact and isotropic pyrolytic carbon ceramic material of the outer layer provides a second protection layer, thereby preventing the loose pyrolytic carbon ceramic material layer of the inner layer from mechanical damage, and further enhancing the safety of the shielding element balls.

The invention provides an austenitic stainless steel plate steel plate for a fast neutron reactor and a manufacturing method of the austenitic stainless steel plate. The steel plate comprises the following components in percentage by weight: 0.04 to 0.10 percent of C, 0.10 to 0.60 percent of Si, 1.00 to 2.0 percent of Mn, less than or equal to 0.035 percent of P, less than or equal to 0.010 percent of S, 11.50 to 12.50 percent of Ni, 17.00 to 18.00 percent of Cr, 2.50 to 2.70 percent of Mo, less than or equal to 0.10 percent of Cu, 0.05 to 0.15 percent of N, less than or equal to 0.0015 percent of B, less than or equal to 0.03 percent of Al, less than or equal to 0.002 percent of Sb, less than or equal to 0.001 percent of Pb, less than or equal to 0.015 percent of Se, less than or equal to 0.005 percent of Sn, less than or equal to 0.10 percent of V, less than or equal to 0.01 percent of Zn, less than or equal to 0.01 percent of As, less than or equal to 0.06 percent of Co, less than or equal to 0.015 percent of Zr, less than or equal to 1.50 percent of of Nb,less than or equal to 1.00 percent of of Ti,less than or equal to 30ppm of [O],less than or equal to 5.0ppm of [H], and the balance of Fe and inevitable impurities. By adding a certain content of nitrogen element, the normal-temperature and high-temperature strength of austenitic stainless steel is greatly improved; and meanwhile, the ferrite content in the finished steel plate is smaller than 1%, and the steel plate has the uniform grain size, the grain size reaches the level 3-6, and meanwhile the steel plate has the excellent intergranular corrosion resistance.

The invention discloses a method for accurately obtaining reactivity feedback changes in the transient process of a fast neutron reactor. By considering spatial distribution of reactivity feedback, accurate fuel temperature reactivity feedback change, coolant density reactivity feedback change, axial expansion reactivity feedback change, radial expansion reactivity feedback change, assembly bending reactivity feedback change and control rod driving mechanism expansion reactivity feedback change in the transient process are obtained, and finally the three-dimensional reactivity feedback quantity in the transient process of the fast neutron reactor is obtained. The method is high in universality and wide in application range, the transient process of the reactor can be accurately simulated,calculation time is shortened, and accurate and reliable reactivity feedback parameters are provided for a design task of the reactor core.

A steam generator, used in a helical coil type steam generator for a sodium-cooled fast reactor which has heat transfer tubes of a double-wall tube structure, with high heat transfer efficiency and a heat transfer tube damage detection unit that can detect on-line in real-time whether the heat transfer tube is damaged or not. The heat transfer tube of a steam generator for a sodium-cooled fast reactor, includes an inner tube formed with a first material; an outer tube formed with a second material that is in close contact with the inner tube and which has a thermal expansion coefficient less than that of the first material; and a plurality of helium flow grooves formed between the inner tube and the outer tube along a lengthwise direction of the heat transfer tube for flowing helium gas. The inner tube and the outer tube are identically due to heat for the temperature difference generated during normal operation of the steam generator, so the degree of close contact between the inner tube and the outer tube does not decrease, and the decrease of heat transfer efficiency can be prevented.

The invention discloses a calculation method used for searching for a balanced cycle of a fast neutron reactor. The method comprises the steps of 1, representing a fuel management scheme as multiple fuel management paths; 2, making a fuel cycle process equivalent to an approximate balanced cycle; 3, for the approximate balanced cycle, performing neutron transport and burnup coupling calculation of an in-reactor cycle to obtain a transmutation matrix of each stage of each fuel management path; 4, repeating the steps 2 and 3 until a nuclear density vector of each stage of each fuel management path is converged, thereby obtaining an in-reactor cycle mode; 5, performing linear interpolation or extrapolation on cycle length, and performing a search to obtain the in-reactor cycle mode meeting discharge burnup level requirements; 6, calculating spent fuel reprocessing recovery and new fuel reproducing processes, performing burnup calculation of the in-reactor cycle on a nuclear density vector of newly loaded fuel according to the transmutation matrix, and repeating the processes until the nuclear density vector of the newly loaded fuel of each fuel management path is converged; and 7, adjusting the enrichment degree of the newly loaded fuel to realize a target effective multiplication factor of a specified time point.

The invention discloses a supercritical carbon dioxide circulation switchable vessel power system of a lead-cooled fast reactor. The supercritical carbon dioxide circulation switchable vessel power system comprises the lead-cooled fast reactor, a reactor cabin composed of a primary loop system of the lead-cooled fast reactor, a cabin composed of a switchable supercritical carbon dioxide power circulation secondary loop system and a propelling device. A lead-cooled fast neutron reactor is adopted as a heat source, and ship propulsion power and equipment electric power are output in a supercritical carbon dioxide power cycle mode combining primary reheating, intermediate cooling compression and shunt recompression processes. Effective switching of supercritical carbon dioxide circulation ofan inner cooling mode and a partial cooling mode under different operation requirements is achieved through gate valve opening and closing operation, and the direct propelling device is combined to drive the vessel to sail. The energy conversion efficiency is remarkably improved, the compactness of the power device and the effective load of the vessel are greatly improved, low-energy-consumption and high-performance operation of the vessel is satisfied; according to the invention, a new thought is provided for application of supercritical carbon dioxide circulation in the field of nuclear power vessels.

The invention discloses a nuclear reactor system, particularly an accelerator driven ceramic fast reactor (ADCFR) system. The system comprises a high-current superconducting linear accelerator, a gravity-driven high-power dense particle spallation target and a full-ceramic fast neutron reactor. The system is subjected to external source drive by the accelerator. The spallation target generates neutrons, so that the reactor carries out a nuclear reaction in a subcritical state. After preset time, the external source drive of the accelerator is stopped, so that the reactor is transformed to operate in a critical state. The structures in the reactor core of the reactor are made of ceramic materials.

A fast reactor has a reactivity control assembly including a reactor shutdown rod of a backup reactor shutdown system and neutron absorbers to suppress the initial surplus reactivity, a reactor shutdown rod drive mechanism for releasing the reactor shutdown rod and units of neutron absorber drive mechanism capable of moving the respective neutron absorbers up and down. The reactor shutdown rod and the neutron absorbers are arranged in a wrapper tube. The reactor shutdown rod drive mechanism causes an inner extension tube to fall and release the reactor shutdown rod by means of a gripper section at the lowermost end of an outer extension tube by turning off the power supply to a holding magnet at the time of scram. Each of the units of neutron absorber drive mechanism has a dual tube type drive shaft including an outer extension shaft and an inner extension shaft. When grasping the neutron absorbers, the outer extension shaft is pulled up to allow both of the extension shafts to be inserted. After the outer extension tube gets to the handling head section of the neutron absorber, the outer extension shaft is pushed down to grasp the neutron absorber externally by means of the latch fingers of the gripper section thereof so that the neutron absorber can be moved up and down.