100 results about "Heavy water reactors" patented technology

The present invention relates to a method for dynamically controlling corrosion in a steam- water circuit of a nuclear power plant, the method comprising: injecting the compounds ethanolamine and optionally ammonia, the system comprising: injection units, located at a dosing point and configured for injecting the compounds ethanolamine and ammonia in the fluid circulating in the circuit; and measurement units, configured for measuring a value of the parameters pH and at least one of the parameters selected from redox potential and corrosion potential in the circuit; the system being characterized in that it further comprises a control unit, the control unit comprising: a first acquisition module, connected to the measurement unit and configured for acquiring the measured values of the parameters and configured for acquiring at least one set-point value for said parameters; a command module, connected to the injection units and configured for commanding the injection of the compounds ethanolamine and ammonia by the injection unit as a function of the acquired at least one set-point value and measured values of the parameters, characterized in that ethanolamine is used as reducing agent. The present invention further relates to a system for dynamically controlling corrosion in a steam-water circuit of a nuclear power plant.

The invention provides a disassembling and assembling device for a heavy water reactor sealing plug. The disassembling and assembling device comprises a base and a disassembling and assembling mechanism. One end of the base is mounted on the truss manipulator, a rotating shaft is arranged on the base, and one end of the rotating shaft is rotationally matched with the base to drive the dismounting mechanism to grab a target sealing plug or drive the dismounting mechanism to dismount the target sealing plug; the dismounting mechanism comprises an outer cylinder, a positioning cylinder and at least two clamping mechanisms, the outer cylinder is used for axially positioning a target sealing plug, the positioning cylinder and the outer cylinder are coaxially arranged and are arranged on the inner side of the outer cylinder, and the at least two clamping mechanisms are arranged in the peripheral direction of the positioning cylinder; the clamping mechanism comprises a first driving piece and a clamping jaw rotationally matched with the first driving piece, and the clamping jaw is configured to be capable of being driven by the first driving piece to contract towards the central axis of the positioning barrel so as to release the target sealing plug or extend away from the central axis of the positioning barrel and be clamped to a clamping protrusion of an inner cavity of the target sealing plug so as to grab the target sealing plug.

There is provided a method of dismantling a dousing tank of a heavy water reactor structure. The method of dismantling a dousing tank of a heavy water reactor structure includes: selecting a dousing tank of a heavy water reactor structure for dismantlement; dismantling a 1st structure inside the dousing tank; fixing the dousing tank to a dome structure portion provided above the dousing tank through a mounting port; positioning a cutting unit on a 2nd structure including a concrete structure inside the dousing tank; and performing, by the cutting unit, cutting based on a forward or reverse rotation manner on an inner peripheral surface of the dousing tank.

The present disclosure belongs to the field of nuclear power technology, and particularly relates to a dry storage structure for spent fuel of a heavy water reactor. In the present disclosure, the storage module for storing spent fuel has a chamber inside, and a plurality of through holes are axially provided on the top plate of the chamber; each storage container is a cylindrical structure with an open top, each storage container is inserted into one of the through holes, and the bottom end of each storage container is fixedly connected to the bottom plate of the chamber; a plurality of fuel baskets are arranged in each storage container, and each fuel basket is used for placing spent fuel; each shielding plug is adapted to plug the opening at the top end of one storage container, thereby closing the storage container. By fixedly connecting the bottom end of each storage container to the bottom plate of the storage module, the supporting manner of the storage container is changed from being supported by the top plate of the storage module to being supported by the bottom plate of the storage module. It not only meets the requirement of dense storage of spent fuel but also meets the seismic requirement.

This invention provides a target loading device for producing short-half-life isotopes in a heavy water reactor, comprising a channel, an electric cylinder, and a target drum assembly. One end of the channel leads to the reactor core. The target drum assembly sequentially pushes the loaded targets to the target outlet. The main shaft of the electric cylinder is coaxially arranged with the target outlet, driving the main shaft to move towards the target outlet to push the target at the outlet into the channel. The target enters the reactor core through the channel. This invention enables the loading of short-half-life radioactive isotope targets for several days or tens of days during heavy water reactor power operation, facilitating installation and maintenance. Different types and quantities of isotope targets can be loaded into the core production channel during heavy water reactor power operation.

This invention discloses a method for producing radioactive isotopes in the core of an advanced high-temperature reactor using a spherical target. This invention overcomes the limitations of using neutron irradiation channels or fuel-related components of nuclear reactors, such as conditioning rods in heavy water reactors and combustible poison rods in pressurized water reactors, to produce radioactive isotopes. It sets a precedent for the mass production of radioactive isotopes using the core of a nuclear reactor, and has significant strategic implications for exploring high-value applications of commercial advanced high-temperature reactors to reduce operating costs and mass-produce radioactive isotopes, including the strategic isotope tritium.

The invention relates to the technical field of fuel rod bundle underwater size measurement, and provides a heavy water reactor fuel rod bundle underwater size measurement system and method. The heavy water reactor fuel rod bundle underwater dimension measurement system comprises a supporting platform; a rotary positioning mechanism; a three-dimensional moving mechanism; the first underwater measurement camera is used for collecting an overall image of the fuel rod bundle and an image of a single fuel element; the camera measurement calibration platform comprises a rod bundle measurement calibration plate and an element measurement calibration plate; the first reflector assembly is used for reflecting the image of the first end plate to the first underwater measurement camera; the camera movement module comprises two module guide rails and a camera bracket capable of sliding along the module guide rails; the second underwater measurement camera is used for acquiring an image of the second end plate; and the second reflector assembly is used for reflecting the image of the second end plate to the second underwater measurement camera. Fuel rod bundle image acquisition and splicing can be completed through the mobile platform and the underwater camera, the detection speed is high, and the measurement precision is high.

The invention belongs to the technical field of heavy water reactor core flux detectors, and aims to solve the problems that the manual flux detector installation speed is low, the personnel irradiation dose is high, and the detector is easy to damage. The invention discloses a heavy water reactor core flux detector installation device and method, the device comprises a front guide pipe and a rear guide pipe which are coaxially arranged, the rear end of the front guide pipe is installed at a reserved installation position of a driving system shell, and a driving pair wheel set is arranged between the front guide pipe and the rear guide pipe; the driving pair wheel set is in transmission connection with the driving motor; the rear guide pipe is in butt joint with a detector reel bin, a reel is installed in the detector reel bin, the tail of the flux detector is clamped into the reel, and a flux detector cable is wound on the reel. According to the method, components of the device are used for installing the device, the heavy water reactor core flux detector installation device is removed, and flux detector installation is completed. The installation speed of the flux detector can be increased, the radiation dose of personnel is reduced, and the probability that the detector is prone to damage is reduced.

The invention provides a blind measurement method for an inner hole of an underwater vertical deep hole channel of a heavy water reactor. The blind measurement method comprises the following steps: establishing a measurement zero reference surface; a working platform is established by taking a measurement zero reference surface as a reference, and three references are carried on the working platform; establishing an axis system with the measurement zero reference surface on the working platform; measuring the distance between the pre-measuring hole and the measuring zero reference surface; and measuring the diameter of the pre-measuring hole, wherein the diameter value is equal to the outer diameter value of a certain step of the inner diameter gauge. The invention further provides a heavy water reactor underwater vertical deep hole channel inner hole blind measurement device which comprises a working platform, a measuring scale, a positioning gauge and an inner diameter gauge. The device can safely and accurately measure the inner diameter size of the vertical deep hole channel in the heavy water reactor underwater high-irradiation environment, realizes blind measurement of the inner hole of the vertical deep hole channel in nuclear power heavy water reactor high-irradiation, remote and underwater measurement, realizes accurate contact measurement, is high in accuracy, is far away from the high-irradiation environment to the greatest extent, and greatly reduces the irradiation dose of operators.

The application discloses a kind of digital fuel physics supervision method and system suitable for heavy water reactor, it is related to nuclear reactor fuel management technical field, system includes data acquisition module, document scale calculation interface, system management module and central processing module, central processing module configuration basic data, new fuel, FCO, spent fuel, dry storage area management unit.Method is standardized by data acquisition, whole-process digital control, dynamic monitoring and tracing three steps, realize fuel full life cycle closed-loop management.System adopts PDA to cooperate bar code / RFID to realize entity information acquisition and offline synchronization, constructs hierarchical correlation model, with new fuel isolation control, FCO closed-loop approval, spent fuel transfer verification, fine authority control and other functions, can improve management efficiency, reduce human error, strengthen data tracing and operation standardization, guarantee nuclear safety, provide support for management decision, suitable for heavy water reactor fuel physics supervision and whole-process management.

This invention relates to the field of radioactive waste management technology, specifically disclosing a trolley device for removing residual tritium from heavy water reactor waste filter cartridges. The device comprises a trolley body, a molecular sieve absorption chamber, and a tail gas treatment filter, connected in sequence. The trolley body has a built-in heating element, a weight monitor at the bottom, and a gas outlet at the right end. The trolley body transports tritium-containing vapor to the molecular sieve absorption chamber through the outlet. The molecular sieve absorption chamber contains several molecular sieves that adsorb the tritium-containing vapor, forming tail gas. The tail gas is transmitted through a pipeline to the tail gas treatment filter, which contains several filter elements for supplementary absorption. A tail gas monitor is located at the outlet of the tail gas treatment filter. This device allows the waste filter cartridges to be treated to be lifted and loaded into the trolley for tritium removal during transport and storage. After treatment, the tritium water content of the waste filter cartridges can be significantly reduced, making them meet relevant disposal requirements.

The invention belongs to the technical field of laser welding, and particularly relates to a super-long slender rod type zirconium 4 alloy integrated laser welding device which comprises an assembling platform, a device control table, a protective gas hood, a welding robot provided with a laser welding gun, a robot power source, a rotating tool and a main body mechanism support. And the robot power supply is connected with the device console and a welding robot equipped with a laser welding gun. The device console is connected with the welding robot and the rotating tool. The length of the workpieces is larger than or equal to 12,000 mm, the wall thickness of the workpieces is larger than or equal to 3 mm, and the workpieces are zirconium-4 alloy workpieces with different diameters. The device console is arranged to control the rotating tool to rotate to the welding position and control the welding robot provided with the laser welding gun to conduct rotating motion in the welding process so as to achieve multi-posture welding at different positions; the technical problem that a production channel assembly and a protective sleeve assembly in an existing heavy water reactor online irradiation isotope production device are difficult to assemble and weld is solved.

An apparatus and a method for separating a shielding ball are disclosed. An apparatus for separating a shielding ball comprises: an end shield which is installed in a spinning structure of a heavy water reactor facility and includes a fueling tubesheet and a calandria tubesheet facing each other and multiple shielding balls for filling a space between the calandria tubesheet and the fueling tubesheet; as a device for removing the shielding balls from a lattice tube connected to the calandria tube by being connected to the calandria tubesheet through the fueling tubesheet, a magnetic force applying unit inserted into the lattice tube in a direction of the fueling tubesheet to apply a magnetic force for fixing positions of the multiple shielding balls positioned within a first distance from the calandria tubesheet; a mover for moving the magnetic force applying unit in a direction of the lattice tube; and a storage rack for separately storing, inside the end shield, the shielding balls positioned within the first distance from the calandria tubesheet and shielding balls positioned at a distance greater than the first distance.

The invention discloses a method for producing radioactive isotopes in a reactor core of an advanced high-temperature reactor by adopting a spherical target piece. According to the invention, the limitation that the radioactive isotope can only be produced by using a neutron irradiation channel of a nuclear reactor or fuel-related components such as an adjusting rod of a heavy water reactor and a burnable poison rod of a pressurized water reactor is broken through, and a precedent of mass production of the radioactive isotope by using the reactor core of the nuclear reactor is created; the method has important strategic significance for exploring high-value application scenes of commercial advanced high-temperature reactors to reduce operation cost and producing radioactive isotopes including strategic isotope tritium in batches.

The present invention belongs to the technical field of dry storage of spent fuel from heavy water reactors, and specifically relates to a device and method for filling lead into a loading and unloading plug. The device comprises a loading and unloading plug, a lead filling funnel, a boiler, an inlet valve, a water trough, an electronic scale, a drain valve, and a heating tape. The loading and unloading plug is placed on the water trough, with a pouring hole formed on the top of the loading and unloading plug. The side of the loading and unloading plug is wrapped with a heating tape. The entire water trough is placed on the electronic scale. The top of the water trough is provided with an inlet valve connected to a pipeline, and the bottom of the water trough is connected to a drain valve via a pipeline. The lead filling funnel and boiler are located outside the water trough. The beneficial effect is that the top of the loading and unloading plug has four evenly distributed pouring holes along the circumference. The lead filling device, combined with the sequential pouring and water trough filling and draining methods, produces a "rock spray"-like liquid lead during the pouring process, maximally removing all gases from the lead solution, ensuring the final density and compactness, and effectively reducing deformation of the shielding plug caused by metal shrinkage in the loading and unloading plug. The lead filling quality fully meets technical requirements.

The invention belongs to the technical field of heavy water reactor nuclear power plant auxiliary facilities, and particularly relates to a heavy water reactor nuclear power plant auxiliary channel shielding plug dismounting device. The fixed support is hinged to the rotating arm, a rotating bolt bearing is installed in a bolt hole below the fixed support, the rotating arm rotates relative to the fixed support and is provided with a through hole, and the shielding plug connecting rod can penetrate through the through hole of the rotating arm to be connected with the shielding plug. The shielding plug tray is fixedly installed on the rotating arm, a through hole is designed in the middle of the shielding plug tray, and the lifting top block penetrates through the through hole. A threaded hole is designed in the tail end of the shielding plug tray, the centering bolt is connected with the threaded hole, the centering angle of the shielding plug tray is controlled by adjusting the exposed length of the centering bolt, and the shielding plug tray is locked and fixed through the centering locking nut. The shielding plug connecting rod is connected with the shielding plug, and when the shielding plug is horizontally located in the shielding plug tray, the shielding plug connecting rod penetrates through the through hole of the rotating arm. The dismounting and mounting efficiency of the shielding plug is improved, and the safety of personnel and equipment in the dismounting and mounting process is ensured.

The present invention specifically relates to a method for predicting the radial creep of a pressure tube in a heavy water reactor, belonging to the technical field of nuclear reactor physics. The method includes: Step 1, calculating the radial creep rate at each fuel bundle of the pressure tube in each fuel channel of the heavy water reactor; Step 2, calculating the predicted value of the average inner diameter at each fuel bundle of the pressure tube in each fuel channel of the heavy water reactor; Step 3, taking the maximum value of the predicted values of the average inner diameters at all fuel bundles of the pressure tube in each fuel channel of the heavy water reactor as the predicted value of the inner diameter of the pressure tube in each fuel channel of the heavy water reactor. The device implements the steps of this method. The present invention obtains the relationship between the radial creep rate and the fast neutron flux at the fuel bundle of the pressure tube in the fuel channel of the heavy water reactor through linear fitting of the radial expansion measurement data of the pressure tube in the fuel channel of the heavy water reactor, and finally realizes the accurate prediction of the radial creep of the pressure tube after the long-term operation of the heavy water reactor.

The present application particularly relates to a method for producing helium-3 from heavy water in a heavy water reactor, comprising: step 1, subjecting the heavy water in a moderator system of a heavy water reactor unit to negative pressure rectification to obtain concentrated high-tritium heavy water; step 2, subjecting the concentrated high-tritium heavy water in step 1 to beta decay to produce helium-3 raw gas; and step 3, subjecting the helium-3 raw gas in step 2 to active carbon adsorption, helium-argon separation and hydrogen isotope gas adsorption to obtain helium-3. The method for producing helium-3 from heavy water in a heavy water reactor avoids technical difficulties and high costs of helium-3 / helium-4 isotope separation, and has high technical maturity and good economy.

The invention belongs to the technical field of nuclear power, and particularly relates to a heavy water reactor nuclear power station atmosphere relief valve grinding device and method. According to the grinding device for the sealing surface of the atmospheric relief valve of the heavy water reactor nuclear power station, the to-be-ground part is effectively clamped through the positioning module, the grinding mould matched with the to-be-ground part is designed and manufactured according to the shape and the size of the to-be-ground part, it is ensured that grinding is in place, and the maintenance quality is guaranteed; rotation driving and up-down driving of the grinding mould are achieved through the grinding module, so that the grinding force can be adjusted at any time in the part grinding process, the grinding face can be kept to be completely attached to the part to be ground, and effective grinding pressing force can be provided in the grinding process. Therefore, safe and efficient grinding treatment on the sealing surface of the atmospheric relief valve of the heavy water reactor nuclear power station is realized.

This invention specifically relates to a heavy water recovery and purification system for heavy water reactor units, comprising an ultraviolet (UV) oxidation subsystem, an adsorption-exchange subsystem, a circulating pump, a filtration and oil removal subsystem connected to the UV oxidation and adsorption-exchange subsystems respectively, a negative pressure exhaust filtration subsystem connected to the UV oxidation subsystem, a monitoring subsystem for monitoring heavy water leakage during heavy water reactor unit recovery, and a control subsystem for controlling the start-up and shutdown of the circulating pump, the UV oxidation subsystem, and the negative pressure exhaust filtration subsystem. The inlet of the circulating pump is connected to the filtration and oil removal subsystem and the UV oxidation subsystem respectively, and the outlet of the circulating pump is connected to the UV oxidation subsystem and the adsorption-exchange subsystem respectively. This heavy water recovery and purification system for heavy water reactor units effectively removes ionic impurities, organic matter, and solid particulate matter while preventing the spread of radioactive materials and avoiding the additional generation of radioactive waste.

The present invention relates to a device and method for conducting a detritium and dehumidification experiment on a waste filter element of a heavy water reactor. The device comprises: a sealed oven for simulating a tritium precipitation environment of a waste filter element and accommodating a filter element to be tested; a circulating air module for establishing a closed gas circulation system and carrying precipitated moisture, comprising a variable frequency blower and an air inlet and outlet duct system; a tritium water adsorption and purification module for simulating the adsorption and purification process of tritium water, comprising at least one molecular sieve adsorption tank that can be connected in series; an automated control module for regulating experimental process parameters and recording experimental data, the automated control module being capable of controlling at least the operating parameters of the variable frequency blower and the temperature of the sealed oven, and monitoring and recording system operating data in real time; the circulating air module, the tritium water adsorption and purification module and the automated control module being integrated into an integrated equipment body, the integrated equipment body being sealedly connected to the sealed oven via the air inlet and outlet duct system, and being capable of maintaining sealing performance under an operating pressure of 0.5±0.1 kPa.

The present application belongs to the technical field of radioactive waste management, and particularly relates to a cutting and volume reduction method for a water filter cartridge. The method comprises the following steps: Step 1: preparation before cutting; Step 2: cutting; and Step 3: cutting waste packaging and cutting machine decontamination. The cutting and volume reduction method provided by the present application is convenient to use, easy to understand, and highly practical, and can ensure effective volume reduction of the water filter cartridge generated by the heavy water reactor process system, is safe and reliable, has high economic benefits, effectively improves efficiency, fully meets the existing cutting needs of the water filter cartridge, and can reduce the volume of the original water filter cartridge from 2.49 m 3 to 0.5 m 3 , thereby reducing the disposal cost in the later stage and achieving significant economic benefits.

This invention relates to the field of target box transportation technology after heavy water reactor irradiation. Addressing the problem that existing transportation channels cannot effectively reduce radiation exposure to personnel during the transportation of radioactive target boxes, this invention provides a target box transportation device after heavy water reactor irradiation, comprising a transportation pipe, a shielded transfer container, and a conveyor belt. A negative pressure is established within the transportation pipe to draw the target box into a vertical section. A separator is installed on the vertical section of the transportation pipe to allow the target boxes to enter the shielded transfer container one by one within a shielded box. The shielded transfer container is transported via the conveyor belt. Furthermore, this invention also provides a method for transporting target boxes after heavy water reactor irradiation, sequentially loading the target boxes from the vertical section into designated positions within the shielded transfer container in the shielded box, and completing radiation protection treatment of the target boxes within the shielded box. By loading the irradiated target boxes into a shielded working box, compared to containerized transport, the dose rate in the operation and maintenance area can be reduced from 2 mSv / h to 25 μV / h, significantly reducing the dose level for operators.

The present invention belongs to the field of reactor physics technology for heavy water reactor nuclear power plants, and specifically relates to a method for preemptively applying the ROP set value for a heavy water reactor transition core with improved fuel. The method comprises the following steps: Step 1: Calculating the X% increase in the ROP set value for the full core with both the original design fuel and the improved fuel; Step 2: Calculating the D EFPDs required to fully fill the high power region of the core with the improved fuel and the corresponding EFPD at the time point; Step 3: Calculating the conservative increase in the ROP set value by Y% compared to the original design fuel core after the fuel channels in the high power region are fully filled with the improved fuel; Step 4: ROP set value evaluation and calculation of the minimum increase by Y% at full power. The present invention has the beneficial effect of increasing the ROP margin earlier, thereby generating more power generation revenue for the nuclear power plant, and reducing the power reduction during refueling earlier, thereby enhancing the operational safety of the nuclear power plant.

Thorium-based fuel bundles are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300 / 600 / 900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to 13-19.95% 235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt%. Reprocessing of spent fuel bundles is also avoided.

The invention belongs to the technical field of nuclear power, and particularly relates to a storage device for heavy water reactor spent fuel and an installation method of the storage device. The whole storage barrel is divided into the manufacturing process of multiple sections such as the upper structure, the multiple sections of barrel bodies and the lower structure, machining treatment is carried out with the annular plate at the bottom of the support as the datum plane, the manufacturing difficulty of all the sections is reduced, the machining precision of all the sections can be improved, and the machining efficiency is improved. The machining precision of each section is further controlled to be higher than the overall machining precision requirement of the storage barrel, allowance is reserved for micro deformation generated after each section is welded, and the overall machining precision of the storage barrel is effectively guaranteed; in the assembling process, the upper plate serves as a datum plane, so that the levelness of the surface of the bottom plate and the concrete pouring technology do not affect leveling of the storage cylinder any more, and errors gradually accumulated from bottom to top are effectively overcome. The overall installation precision of the storage cylinder is ensured, and spent fuel loading requirements are met.

Disclosed is a method for dismantling a D2O tank of a heavy water reactor installed on an inner wall of a containment building and surrounded by a concrete structure, the method including cutting and opening a front surface of the concrete structure so as to dismantle the D2O tank, descending the D2O Tank by using an dedicated dismantling apparatus, and taking the D2O tank out of the containment building by using a crane, and a dismantling apparatus.

The invention belongs to the technical field of nuclear reactors, and particularly relates to a high-strength corrosion-resistant Zr-Sn alloy plate for a heavy water reactor tube bank and a preparation method of the high-strength corrosion-resistant Zr-Sn alloy plate. The high-strength corrosion-resistant Zr-Sn alloy plate for the heavy water reactor drainage pipe, provided by the invention, comprises the following components in percentage by mass: 1.5 to 1.7 percent of Sn, 0.12 to 0.2 percent of Fe, 0.1 to 0.15 percent of Cr, 0.03 to 0.08 percent of Ni, 0.002 to 0.015 percent of C, 0.09 to 0.16 percent of O, 0.002 to 0.01 percent of Si, less than or equal to 0.0030 percent of N, less than or equal to 0.0020 percent of H and the balance of zirconium and inevitable impurities. By controlling the components and impurity content of the alloy, the prepared alloy has higher strength and excellent corrosion resistance, so that the plate can meet the application requirements of the heavy water reactor tube bank.

The application discloses a kind of heavy water reactor production 99 The target nucleus of Mo includes target tube, target material layer and neutron absorber material, the target material layer is sleeved on the target tube, and the both ends of the target tube are respectively connected with inner end plug, and the neutron absorber material is arranged in the target tube. 99 The target nucleus production element of Mo includes target nucleus, cladding tube, isolation block, support block and outer end plug, the target nucleus is arranged in the circular tube-shaped cladding tube, the both ends of the cladding tube are welded with outer end plug to seal it, and the outer part of the cladding tube is respectively welded with isolation block and / or support block. The production assembly includes production element and end plate, and the end plate is welded and connected with the both ends of the production element. The beneficial effects of the application are that the effect of the application is to produce carrier-free 99 Mo while producing electricity by heavy water reactor, which meets the demand of nuclear medicine field for 99 Mo.

The invention particularly relates to a method for reducing hydrogen and deuterium concentration at a key position of a heavy water reactor pressure pipe, and belongs to the technical field of aging protection of heavy water reactor pressure pipe materials, the pressure pipe is made of zirconium alloy, two ends of the pressure pipe are in expanded connection with end parts, and an expanded connection pressure stress area, an expanded connection tensile stress area and a pipe body area are defined; the method comprises the steps that at least one hydrogen absorption material block is fixedly installed in the area, corresponding to an expanded connection tensile stress area, of the outer surface of the pressure pipe; the hydrogen absorption material block is made of yttrium or yttrium alloy; and the hydrogen absorption material block is used for absorbing hydrogen and deuterium diffused to the expanded connection tensile stress area from the expanded connection pressure stress area of the pressure pipe. According to the method, the hydrogen equivalent concentration of the expanded connection tensile stress area, especially the polishing mark position is obviously reduced, the hydrogen equivalent concentration is prevented from exceeding the safety limit value, and the service life of the pressure pipe is prolonged.