Low pressurized water reactor and control method thereof

Abstract

To provide a nuclear reactor that operates at low pressure with less chance of accidents and less maintenance. The low pressure water reactor 1 of the present invention comprises a reactor vessel 100, an ascending pipe 104, a water tank 200, a steam drum 150, a water injection system, a steam turbine 161 operating at a pressure of 1-10 bar, and an outer annulus surrounding the vessel 100, in which the ascending pipe 104 forms a descending pipe 105 for generating a high velocity flow. The pressure of the core 101 is 6-15 bar. The drum 150 is connected to the ascending pipe 104, and the water tank 200 is connected to the vessel 100 to store boron water. The water injection system transfers the boron water from the tank 200 to the vessel 100. The ascending pipe 104 converts the hot water from the vessel 100 into steam and sends it to the turbine 161. The primary coolant naturally circulates in the ascending pipe 104, the descending pipe 105, the drum 150, and the core 101 due to the density difference.

Description

[Technical field]

[0001] The present invention relates generally to the field of nuclear reactors, and more particularly to low pressurized water reactors (LPWRs) and methods for controlling same. [Background technology]

[0002] Since the advent of electricity as a source of energy, the quality of life of human beings has improved significantly and has enjoyed its benefits. Often, along with food and water as basic needs, the availability of electricity is directly related to economic growth. Global energy demand is estimated to grow at an annual rate of 2.5% between 2011 and 2035 (Non-Patent Document 1). Fossil fuels, which account for about 87% of the world's energy mix, are also predicted to be depleted in the next 80 years. However, if the whole world lives the same way as the United States, such lifestyles can only be sustained until 2035 (Non-Patent Document 1).

[0003] The most common method of generating electricity today is to convert the energy stored in fossil fuels into electrical energy. However, converting fossil fuels into electrical energy has problems, such as limited availability of natural resources, emission of harmful emissions, and national security due to dependency on foreign countries for fossil fuel resources. On the other hand, generating electricity from renewable sources faces problems such as economies of scale, technological maturity / development, and reliability / availability of resources to provide alternatives to fossil fuel production, not to mention the political debate that accompanies the construction of large dams for hydroelectric plants. Large lakes and rivers provide the energy required by hydroelectric plants, but are not widely available. Wind is a strong candidate in the energy mix and could replace a lot of baseload generation in terms of annual energy generating units (kWh). However, the wind can stop, making this technology unreliable and unable to provide large on-demand power (kW). This makes it unusable unless paired with energy storage such as battery arrays or pumped hydroelectric plants. The alternative is the use of nuclear power, which, like renewable electricity, faces political and environmental challenges, as well as concerns about accidents and waste disposal. However, nuclear power has the well-known advantages of availability of high-energy-density nuclear fuel, allowing reliable continuous operation, and easy access to nuclear fuel. Like renewable electricity generation, nuclear power plants also have the advantage of not emitting greenhouse gases and returning harmful industrial waste to the biosphere.

[0004] Nuclear power plants produce electricity by the release of energy from fissile materials (U, Pr), which is only possible through nuclear fission in a controlled chain reaction. This heat is usually either transferred to the reactor coolant, which becomes the working fluid (BWR: Boiling Water Reactor), or transferred to a secondary working fluid (PWR: Pressurized Water Reactor), which drives the blades of a turbine-generator to produce electricity.

[0005] Nuclear power generation relies on complex modern technology and safety is a concern. Equipment malfunctions can lead to accidents that can expose the public to ionizing radiation and cause death or large-scale contamination of land. In fact, nuclear technology can cause severe environmental damage over large areas and even beyond national borders. Safety concerns, related to the risks and complexity of nuclear technology, drive public opinion and negatively influence contemporary public decision-making.

[0006] Reactor designers and relevant regulators envision a number of events that could lead to an accident at a nuclear power plant. In particular, the potential consequences of an accidental loss of coolant from the reactor (known as a LOCA) could result in core meltdown and release of radioactive materials (3,4). The dependency of many nuclear technologies on the availability of electricity (used to drive electric pumps and control devices) is also a concern. As evidence, the 2011 Fukushima nuclear accident was cited. Submersion of the reactor vessels led to a total loss of power, which led to meltdowns in three reactor vessels. Reliance on highly skilled personnel to operate the plant's safety systems is another potential deficiency, resulting in vulnerabilities due to human error. An example of this is the 1979 accident at the Three Mile Island nuclear power plant in the United States.

[0007] To address the above safety issues, reactor designers have proposed improved reactor design concepts, which allow for a high level of safety and high reliability. These novel design features aim to increase safety and reliability, rebuild public confidence in nuclear energy, ensure investment protection for nuclear power plant owners (providing competitive costs over the entire life cycle), and reduce economic risks. Novel reactor technologies can be broadly divided into two categories (Non-Patent Document 5): Generation 3 / 3+ reactor vessels, which are innovative designs, and Generation 4 reactor vessels.

[0008] Generation 3 / 3+ designs are based on evolutionary improvements to current Generation 2 reactors, with an emphasis on maintaining current improved design features, minimizing engineering risk, and incremental improvements to the next generation design. (6) These designs have several advantages over Generation 2 reactor designs. ●Finance / Project Management: Improved ease of obtaining operational permits, shortened construction time, reduced investment costs, and reduced vulnerability to operational malfunctions. ●Extended equipment service life: Reducing shutdown periods and extending the operating life to over 60 years. • Improved nuclear fuel efficiency: This is achieved by increasing the burning of nuclear fuel isotopes, increasing the use of burnable absorbents, and improving thermal efficiency. Reducing the likelihood of a core meltdown: The rate of damage to the core must be reduced by an order of magnitude compared to current Generation II reactors. • Increased resistance to structural damage from aircraft crashes and earthquakes.

[0009] However, 3 / 3+ generation nuclear reactors built within the past 20 years have many problems. The most difficult are the significant increase in capital costs and long delays to projects. Two of the European Pressurized Reactors (EPRs) currently under construction (Olkilto Nuclear Power Plant in Finland and Flamaville Nuclear Power Plant in France) are facing rising construction costs, and the project completion dates have been pushed back by many years. The EPRs under construction in China are also significantly delayed. Non-Patent Document 7 suggests a review of nuclear power plant construction costs around the world until 2014, and concludes that significant cost increases and construction schedule delays will occur in all projects. The reasons are as follows: 1. The need for excessive resource development to accelerate construction plans. 2. Adoption of fundamentally uneconomical designs. 3.Increased rules / regulations to enhance safety. 4. Increasing complexity with successive generations of plant designs that fail to learn from the past. 5. Lack of standard design. Similar trends of rising costs have been noted in other publications8-11, particularly in Western countries.

[0010] Primarily as a result of high capital costs and associated cost uncertainties, the selling price of electricity from nuclear plants in Western countries is higher than current wholesale electricity prices and also compared to electricity prices from gas-fired power plants and renewables (wind, solar) which tend to be less vulnerable to cost uncertainties.

[0011] To overcome the economic drawbacks associated with current power generation from larger reactors (with outputs of around 600-1600 MW(e)), several small reactor designs with outputs of around 300 MW(e) have been proposed. Many of these designs feature modularity and factory assembly, allowing for greater standardization of components and processes compared to traditional larger reactors (with their potentially higher costs). Smaller reactors have the potential for advantages in terms of overall structural simplicity, faster construction speeds, and reduced financial risks.

[0012] Small modular reactors have a number of attractive design features, including: ● Scalability: Many companies are cautious about the financial commitment. Large nuclear power plants take 5-10 years before they can start selling / selling power. The scalability of small modular power plants allows them to sell power and earn revenue as soon as the first modules are up and running. These revenues allow companies to start investing in new modules and other power plants.

[0013] ● Modularization: In contrast to large nuclear power plants, which have traditionally been built on site, modular construction allows the components of a nuclear plant to be assembled in a factory and transported to the site. Construction in a controlled environment improves construction quality and allows for more precise manufacturing and quality inspection. Another advantage of modular construction is that different components can now be assembled simultaneously anywhere in the world and transported to their desired location at the planned time. The development of modular construction in other engineering disciplines has allowed for significantly shortened construction times, without compromising quality.

[0014] ●Simplification: Smaller, simpler designs with fewer components reduce the incidence of accidents that can damage a reactor. (Non-Patent Document 12) Another major benefit of simple designs is the reduction in capital costs associated with designing and constructing complex safety systems.

[0015] ● Site selection: Small modular reactors have more locations to build than traditional large reactors. They are therefore suitable for construction in remote and off-grid locations. Examples of locations include desalination plants, oil drilling sites, etc.

[0016] These factors suggest that there is a case / benefit for developing small reactors with power output of 300 MW(e) or less for commercial electricity generation. However, according to Non-Patent Document 13, currently published cost information provides little reliable information to realistically evaluate costs compared to larger reactors. In the case of water-cooled reactor technology, small reactor designs have the complexities (scaled down to size) associated with many of the safety features required for larger reactors. As none of the small modular reactor designs are licensed in Western countries, the problem of increased design complexity in response to regulatory requirements leads to increased costs, similar to what is currently experienced with larger plant designs.

[0017] Current reactors used for commercial electricity generation are therefore water-cooled reactors operating at high pressure with a power output of around 1000 MW(e).

[0018] Assuming an accident in such a plant, as an example, due to leakage of primary coolant or failure of the core cooling system, could result in a core meltdown, with potentially catastrophic environmental damage. To address such safety concerns, advanced reactor designs have been developed to reduce the risk of core meltdown and to improve resistance to hazards such as earthquakes and aircraft collisions. However, the increased complexity of these designs has led to increased construction costs and delays in current projects, further delaying investment in nuclear power in the Western world. To improve the utility of nuclear power, several small reactor designs have been proposed, but these are merely scaled-down versions of existing large reactors, and the increased design complexity due to safety requirements results in high and uncertain investment costs. Therefore, a new low-pressurized water underground reactor is needed that requires only a minimal number of simple safety mechanisms. Summary of the Invention [Problem to be solved by the invention]

[0019] The objective of this invention is to provide an inherently safe, natural circulation nuclear reactor capable of generating approximately 300 MW(E) of electricity with a simple design that minimizes reliance on complex safety features, while reducing construction and maintenance costs and reducing the risk to the environment from a reactor accident. [Means for solving the problem]

[0020] The object of the invention is achieved by a low pressurized water nuclear reactor according to claim 1 and a method according to claim 9.

[0021] According to a first aspect thereof, the low pressurized water reactor of the present invention comprises a reactor vessel, a steam drum, a water tank, an automatic water injection system and a low pressure steam turbine. The reactor vessel comprises an internal cavity for receiving a primary coolant (usually water), an ascending pipe, a core arranged below the ascending pipe, and an outer annulus surrounding the reactor vessel and the ascending pipe, which forms a descending pipe. The descending pipe has a predetermined flow area and produces a flow rate of more than 10000 kg / s. The core is arranged underground at a depth of at least 55 m and has an internal pressure in the range of 6-15 bar. The steam drum is connected to said ascending pipe via its upper end, is arranged above ground and has an internal pressure in the range of 1-10 bar. The water tank is connected to the reactor vessel and stores boron-containing water (hereinafter referred to as "boron water"). The automatic water injection system injects the boron water from the water tank into the reactor vessel. The automatic water injection system is a safety system that has pressure-activated valves that allow coolant to enter the reactor. The driving force for the control rod drive mechanism is the energy stored in the springs and the gravity action of the coolant (density-driven flow). The low-pressure steam turbine operates at pressures between 1 and 10 bar.

[0022] The reactor vessel heats the water to a desired temperature, but does not reach the thermodynamic saturation point. The ascending pipe converts the heated water into steam, which is then sent to the low pressure steam turbine through the steam drum. This conversion to steam creates a density difference in the primary coolant, which starts a natural circulation of the primary coolant through the ascending pipe, the descending pipe, the steam drum, and the reactor core.

[0023] The simplified automated LPR features eliminate many of the components and safety features that require large capital and maintenance costs in conventional HPR vessels, and can generate electricity at lower costs than HPR vessels.

[0024] In particular, steam is generated by means of flashing in the ascending tubes, i.e., as the primary coolant rises in the ascending tubes, pressure reduction occurs due to the pressure drop associated with its upward displacement and the pressure drop due to friction as it moves up the ascending tubes. Flashing, as is well known in the art, is the production of steam when a high pressure liquid enters a lower pressure area or when the saturation point of water drops due to a pressure drop in a tube section.

[0025] For flow rates above 10000 kg / s (hereafter also referred to as "flow rate"), there are other features in the outer annulus that affect the mass flow rate (e.g. inlet resistance, core geometry (number of fuel assemblies), spacer grids). This is an internal feature of the outer annulus, but its external features (e.g. core control packages, openings for maintenance and parts installation) do not affect the mass flow rate.

[0026] In one embodiment, the interior of the ascending tube is empty, but may contain rails, pipes, structures, and cables for the control rod mechanisms that descend through the core.

[0027] In one embodiment, the nuclear reactor core (hereinafter simply referred to as the "core") is located at a depth of more than 100 meters below the ground.

[0028] In one embodiment, the reactor vessel includes a control rod drive module above the reactor core.

[0029] In one embodiment, the core is constructed with off-the-shelf EPR fuel assemblies. In another embodiment, the low pressurized water reactor of the present invention is placed on top of conventional high pressurized water reactor fuel assemblies. European Pressurized Water Reactor fuel assemblies (EPR) are employed because they are the most current and available, but the present invention is not limited thereto.

[0030] In one embodiment, the core includes 120-180 EPRs.

[0031] In one embodiment, the steam drum includes a dryer, a top cover, and a splash plate disposed within a watertight chamber.

[0032] In one embodiment, the descending pipe comprises a concrete shaft (hereinafter also referred to as the "tube") lined with steel.

[0033] In one embodiment, the water tank is connected to the reactor vessel via an automatic on-off valve (hereinafter referred to as an "automatic valve") or a manual on-off valve.

[0034] The low pressure water reactor vessel of the present invention is shut down or shut down by means of either a shutdown rod or a reservoir containing borated water.

[0035] In one embodiment, the boric acid water stored in the reservoir contains 1000 ppm boric acid and the reservoir is located at a height of 20 m or more above ground level, in another embodiment, the boric acid water contains 1500 ppm boric acid, and in a further embodiment, the boric acid water contains up to 3000 ppm boric acid.

[0036] In one embodiment, the core has an inlet temperature between 100° C.-150° C. and an outlet temperature between 140° C.-170° C. The inlet resistance of the core reduces static instabilities. In another embodiment, the steam drum has a steam outlet temperature between 100° C.-170° C., a diameter of 10 m or more, and a height of 3 m or more above ground level.

[0037] According to a second aspect, the present invention provides a method of controlling a low pressure water reactor vessel of the first aspect. The method of the present invention comprises the steps of (A) heating water in the reactor vessel to a predetermined temperature without reaching a saturation point, and (B) converting the heated water into steam in the ascending pipe. The steam is sent to a low pressure steam turbine via a steam drum. Step (B) thereby creates a density difference in the primary coolant. This density difference initiates natural circulation of the primary coolant through the ascending pipe, the descending pipe, the steam drum and the reactor core.

[0038] In one embodiment, the method further comprises the step of (C) lifting the core out of the reactor vessel during a nuclear refueling operation of the low pressurized water reactor vessel.

[0039] In one embodiment, the method further comprises the step of (D) controlling or closing the automatic valve when the boric acid concentration in the primary coolant reaches a predetermined value.

[0040] In one embodiment, the core operates in a single-phase coolant modulator during steady-state operation of a low-pressurized water reactor, producing steam with a coolant (moderator).

[0041] In one embodiment, the parameters of the low pressurized water reactor of the present invention are: Depth of reactor core: 50m or less underground Thermal power of the core: 1500MW-2500MW Number of fuel assemblies in the core: 120-180 EPR type fuel assemblies Core inlet resistance K:1-3 Core inlet pressure: 6 bar or more Core inlet temperature: 100℃-150℃ Core outlet temperature: 140℃-170℃ Ascending pipe flow area: 5m 2 End Steam drum pressure: 1-10 bar Steam drum outlet temperature: 100℃-170℃ Steam drum diameter: 10m or more Steam drum placement height: 3m or more above ground Turbine inlet pressure: 1-10 bar Flow area of ​​the descending pipe: 5m 2 End Mass flow rate of the descending pipe: 10,000 kg / sec or more Boric acid concentration in the Reactor Water Storage Tank (RWST): 1000ppm or more Height of water in Reactor Storage Tank (RWST) above sea level: 20m or more Steam drum piping outlet height: 10m-20m [Brief description of the drawings]

[0042] [Figure 1] 1 is an exploded view of a low pressurized water reactor during normal operation according to an embodiment of the present invention; [Diagram 2] A diagram showing the structure of the reactor room and nuclear fuel pool during normal operation of the reactor. [Diagram 3] A diagram showing the structure of the reactor room and nuclear fuel pool during nuclear refueling operations. [Figure 4] FIG. 13 shows a DHRS system according to another embodiment. [Diagram 5] FIG. 13 is a diagram showing an FPCS system according to another embodiment. [Figure 6] 1 is a diagram showing a core of a low pressurized water reactor according to an embodiment of the present invention. [Figure 7] FIG. 2 is a diagram showing a system of a low pressurized water reactor according to another embodiment of the present invention. [Figure 8] A diagram showing the relationship between system pressure and thermal efficiency. The horizontal axis is system pressure, and the vertical axis is thermal efficiency. [Figure 9] A diagram showing natural circulation flow (loop) during steady operation. [Figure 10] Diagram showing the subchannel morphology used in the low pressure water reactor core model. [Figure 10] Diagram showing subchannels at peak and average core temperatures. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0043] The present invention proposes a low pressurized water reactor (hereinafter also referred to as "LPWR"). The low pressurized water reactor of the present invention achieves high power output of about 300 MW(e) but is a simple design with inherently high safety. The low pressurized water reactor 1 of the present invention is a light water cooled and moderated reactor, operating at pressure close to atmospheric pressure. The reactor of the present invention is of pool design and is located at the bottom of a shaft located deep underground. Primary coolant flow is achieved by density-driven natural circulation and does not rely on electrically driven circulation pumps. The reactor of the present invention has minimal reliance on active / electrical systems for reactivity control or post-shutdown cooling.

[0044] FIG. 1 shows an embodiment of a low-pressurized water reactor 1 according to the invention. In this low-pressurized water reactor 1, the core 101 is placed at the bottom of a concrete-lined cylinder buried deep in the ground. This cylinder is connected to a high-capacity water tank 200 (containing boric acid water) by a reliable automatic flood valve 130. This design allows the connection to the water tank 200 by the automatically operating flood valve 130 to shut down the reactor and provide effective cooling for several weeks in all foreseeable failures and dangerous situations, even in the absence of operator intervention. In addition, the risk of nuclear fuel damage in an accident involving loss of primary coolant or loss of the normal heat removal system is effectively avoided.

[0045] In particular, according to this embodiment, the reactor vessel 100 consists of a pipe (5 m diameter, 100 m long) with a closed hemispherical bottom. The pipe is placed in a steel-lined concrete shaft that extends to a depth of more than 100 m. A narrow annulus between the reactor vessel 100 outer wall and the shaft is used for leakage monitoring. The upper end of the reactor vessel 100 is connected to a 15 m diameter upright cylindrical steam drum (hereinafter simply referred to as the drum) 150 that acts as a water separator. The drum 150 includes a dryer, a top cover 151, and a splash plate 152, which are removed during the refueling operation. The drum 150 is placed in a watertight room (reactor room 155) that is flooded (filled with water) during the refueling operation.

[0046] At normal power operation, the steam pressure in the drum 150 is 2.6 bar and the pressure in the inlet plenum (space) below the core 101 is 11.5 bar. The reactor vessel 100 includes a concentrically arranged ascending tube 104 (diameter 3.5 m) which is connected at its outlet end to the drum 150. A wide annulus (0.75 m wide) between the reactor vessel 100 and the ascending tube 104 constitutes the descending tube 105, which provides the driving head for the natural circulation flow through the core 101. The core 101 is located at the bottom of the ascending tube 104 and consists of 120-180 (150 in this example) EPR type nuclear fuel assemblies, which is the same design as used in conventional high pressure water reactors. During refueling operations, the core 101, with all of the fuel assemblies loaded on it, is lifted completely off the ascending tube 104 and placed in a placement area on the floor of the reactor room (also called reactor pool) 155. The reactor vessel control rod drive mechanisms are contained within the fuel assembly module 103 located above the core 101. The fuel assemblies are relocated into the reactor room 155 during refueling operations. The use of the fuel assembly module 103 inside the reactor vessel pressure circuit is similar to that found in current SMR designs (e.g., NUSCALE design, mPower design) (Non-Patent Documents 14, 15).

[0047] In the low pressurized water reactor 1 of the present invention, the circulation of the primary coolant during reactor operation is achieved by natural convection and does not use a circulation pump, as is used in conventional high pressurized water reactor designs with comparable power output. The water entering the upper part of the descending pipe 105 from the drum 150 is close to saturation temperature at drum pressure, while the water in the core 101 is subcooled by hydrostatic pressure in the descending pipe 105. Thus, as in conventional high pressurized water reactors, the primary coolant in the core 101 (away from the small boiling region at the top of the central fuel assemblies) is mainly subcooled, avoiding potential instabilities due to void-induced reactivity feedback. As the primary coolant rises along the ascending pipe 104, the associated pressure drop leads to the occurrence of a two-phase condition (flashing) above a certain height. The reduction in density in the two-phase region significantly improves the driving head pressure for natural circulation, and as a result, it becomes possible to achieve a core mass flow rate equivalent to that of a conventional high pressurized water reactor using a circulation pump. Therefore, the low pressurized water reactor 1 of the present invention can achieve a thermal power level equivalent to that of a conventional high pressurized water reactor.

[0048] After phase separation in the drum 150, the steam travels through multiple main steam lines 173 to a low pressure steam turbine 161 in a turbine building 160. Condensate from a condenser 162 in the turbine building 160 is returned to the drum 150 by a feedwater pump 164 through multiple main inlet lines 172. Due to the low operating pressures, the mechanical consequences of a failure in the aboveground primary coolant circuit piping are tolerable and will not result in an uncovering or dryout of the core 101 causing a nuclear fuel failure. A failure in the underground pressure circuit piping is also tolerable and will not result in a nuclear fuel failure. Since the low pressurized water reactor 1 of the present invention does not release a large amount of radioactivity into the environment even in the event of a loss of coolant accident (LOCA), the primary coolant circuit in the low pressurized water reactor 1 of the present invention can be constructed of low pressure piping and does not use complete forged parts required in normal high pressurized water reactor designs, and therefore does not require expensive manufacturing and inspection techniques that were previously required. In addition, since the primary coolant released in a LOCA does not contain many radioactive materials, the reactor building of the present invention can be an industrial building with only filter ventilation equipment. This industrial building is not an airtight vessel that can withstand the high pressures required in conventional light water reactor designs.

[0049] The water reservoir 200 is a multi-function storage tank that facilitates nuclear fuel evacuation and transfer while also serving as an automated means of providing the primary safety functions of reactor shutdown and reactor cooling. The high volume water reservoir 200 holds approximately 12,000 m3 of water containing 1500 ppm boron. 3 It can store water and performs the following safety functions 1 to 3. 1. To convert the steam from the reactor back into water after a turbine trip when the condenser 162 is unavailable. 2. Provide a source of water to flood the reactor room 155 during nuclear refueling operations. 3. Provide a source of boron water to ensure that the reactor can be flooded (filled with water) in an emergency, if necessary, to shut down and cool the reactor and to ensure a source of make-up water for the spent fuel storage tank.

[0050] During normal plant shutdowns or unscheduled trips with the condenser live, the turbine bypass valve automatically opens to divert steam generated in the reactor to the condenser and activates the Decay Heat Removal System (DHRS) to bring the reactor into cold shutdown. During accident conditions (when the turbine bypass fails or the condenser is not operational), the flood valve 130 and the steam dump valve 131, which inject water into the reactor, automatically open and connect the water tank 200 to the reactor water network. To perform reliable safety operations, the flood valve 130 and the steam dump valve 131 are held closed by auxiliary electromagnets during reactor operation and automatically open (under the action of a spring or stored compressed air) in the event of an emergency (when the electrical current driving the electromagnets is lost).

[0051] Following operation of the flood valve 130 and the steam dump valve 131, the reactor steam is discharged through the sparge header 201 into the water tank 200. The reactor steam is condensed until the reactor is flooded through with boron water. After flooding, short-term decay heat removal is achieved by heating the water in the water tank 200, and long-term decay heat removal is achieved by the Decay Heat Removal System (DHRS). In the event of a DHRS failure, e.g., failure due to a long-term power outage, it may take 24 hours or more for the water in the water tank 200 to reach saturation temperature. Reactor cooling can then be performed indefinitely by boil-off of water from the water tank 200 and make-up of water to the water tank 200 if the DHRS cannot be restored.

[0052] Nuclear fuel storage and handling In Fig. 2, the spent nuclear fuel removed from the low pressurized water reactor of the present invention is stored in a storage shelf 191. The storage shelf 191 is located in a nuclear fuel storage pool 190 adjacent to the reactor room (reactor pool) 155. After the storage period has elapsed, the spent nuclear fuel assemblies are transferred to a storage facility 192 via a transfer room 193 (located next to the nuclear fuel storage pool 190). Then, they are transferred to a long-term storage facility or a reprocessing facility. FIG. 3 shows the building configuration during a refueling operation. To perform the refueling operation, the top lid 151 and splash plate 152 are removed from the drum 150. The reactor room is flooded (inundated) with borated water from the water tank 200. The slotted gate 195, which connects the reactor room 155 to the fuel storage pool 190, is opened to allow the transfer of the fuel assemblies to the storage shelf 191. Before the fuel assemblies are transferred (position 157 in FIG. 3), the fuel assembly module 103 and the core 101, fully loaded with fuel assemblies, are lifted and placed in a loading area in the reactor room 155 for easy removal / stowing. Following the loading of new fuel assemblies and the relocation of the reused fuel assemblies, the core reloaded with fuel assemblies is lowered into the reactor vessel (tube), the fuel assembly module 103 is replaced, and the control rod latches are reengaged.

[0053] Work safety features Nuclear power plants have ESF (Engineered Safety Features), i.e., work safety features. of They control the nuclear reaction state of the core during the operating conditions and accidents that the plant will experience during its entire life, and they control the cooling of the core and the contamination by radioactive materials. In the current design of water-cooled reactor vessels, the main ESFs include the following three: 1. Watertight vessel structure: This vessel structure houses the reactor water network and contains radioactive materials that may be released in the event of a loss of coolant accident (LOCA). 2. Emergency core cooling and water injection system: This system removes decay heat and refills the reactor vessel in the event of an accident. 3. A system for cooling spent nuclear fuel in the nuclear fuel pool during nuclear refueling operations. These ESPs require support mechanisms in the event of an emergency (flood, fire, explosion). The support mechanisms provide the system structures and components that provide power and control functions, cooling / ventilation functions, and protect the essential components. ESFs and support mechanisms are classified from a safety perspective according to the importance of their safety maintenance functions and specifications. These structures / systems / components classified from a nuclear safety perspective must meet design / construction / maintenance standards that exceed those of conventional equipment. As a result, providing the necessary ESFs and support mechanisms have a significant impact on the construction and maintenance costs of a modern nuclear plant.

[0054] The ESF of the low pressurized water reactor 1 of the present invention has a much smaller number of inspection items than the ESF required by the third generation high pressurized water reactor, so the construction and maintenance costs of the low pressurized water reactor 1 of the present invention may be significantly lower. The main ESFs required for the low pressurized water reactor 1 of the present invention are listed and compared with the equivalent requirements of a conventional high pressurized water reactor. In addition to the ESFs, other major systems are required for the operation of a low pressurized water reactor plant. For example, a refrigerant chemistry / volume control system, a power conversion system to drive the turbine / condenser, are required for operation. However, these do not need to be considered in the low pressurized water reactor 1 of the present invention, because their contribution to the construction and operation costs of current nuclear power plants is relatively small.

[0055] Reactor containment structure Current designs of conventional high pressure water reactors require the reactor primary circuit to be enclosed in a large, overpressure-resistant containment structure. The stringent requirements for the reactor containment building are due to the fact that in the event of a major loss of coolant accident (LOCA), the fuel can overdry and overheat, leading to fuel pin failure and the release of radioactive fission products into the reactor pressure circuit. Enclosing the pressure circuit in a containment vessel allows radioactivity that would otherwise leak into the environment to be confined to these devices and kept at a safe level. The reactor containment provides a barrier against the release of radioactive material in the event of severe core damage (due to failure to restore core cooling) during a LOCA or other accident.

[0056] The design concept of the low pressurized water reactor of the present invention is to prevent the occurrence of nuclear fuel failure under any accident conditions. This design concept maintains the integrity of the concrete walls of the water tank 200, reactor room 155 and nuclear fuel storage pool 190. Therefore, there is no need for a leak-tight sealed containment building to hold the fission products released from the damaged nuclear fuel in the event of an accident. Thus, the reactor of the present invention can be housed in a conventional building with a ventilation system.

[0057] Emergency Core Cooling System The emergency core cooling system used in water-cooled nuclear reactors consists of multiple subsystems or "trains". The subsystems consist of high capacity pumps, heat exchangers, and storage tanks. These devices inject coolant into the reactor and re-establish long-term core cooling in the event of an emergency / emergency. To meet the safety design requirements that apply to nuclear safety systems, the emergency core cooling system of a conventional high pressurized water reactor consists of four identical "trains". These "trains" have various backup systems to cover the case of a common cause failure that would cause the failure of all four "trains".

[0058] In the low pressurized water reactor 1 of the present invention, short-term emergency core cooling is performed by opening a number of redundant valves, namely, a flood valve 130 and a steam dump valve 131. The flood valve 130 and the steam dump valve 131 are provided in a water injection pipe 175 and a steam dump pipe 176. The water injection pipe 175 and the steam dump pipe 176 connect the reactor vessel 100 and the water tank 200. Thus, density-driven flooding of the reactor is performed with boron water. To ensure high reliability of the valves that open when necessary, the flood valve 130 and the steam dump valve 131 are kept closed by electromagnets and are designed to open automatically (by spring actuation or release of compressed air) when the external current is interrupted. In particular, to achieve redundancy, two independent sets of valves are manufactured in various designs.

[0059] After the reactor vessel 100 is flooded (injected with water), short-term decay heat is removed by natural circulation (thermosyphon). This natural circulation passes through the descending pipe 105, the core 101, the ascending pipe 104, the steam dump pipe 174, and the water injection pipe 175. Long-term decay heat can be removed by manually starting the DHRS. The DHRS has a pump and a heat exchanger, and rejects heat to the component cooling system. Even in the worst case accident situation, boiling of the water in the water tank 200 can be avoided by connecting to the DHRS within 24 hours.

[0060] In FIG. 4, in one embodiment, to achieve redundancy and diversity, the DHRS is configured with three "trains," two of which are for normal operation (Main) and the remaining one is for backup (Back-up) and is normally on standby. The two normal operation trains reject heat to a freshwater component coolant system (CCWS), which is then cooled by a seawater or river water component coolant system (SWS) and ultimately becomes a heat sink. The third backup train is cooled by a diverse cooling system (DCS) and rejects heat to a diverse heat sink (DHS). The DHRS design allows any one of the three heat rejection trains to maintain the reactor cooling below acceptable temperatures for 24 hours after reactor shutdown.

[0061] Nuclear fuel pool cooling system The main objectives of the Fuel Pool Cooling System (FPCS) are to: 1. To remove decay heat from spent nuclear fuel in the nuclear fuel storage pool 190 or reactor room (reactor pool) 155 during nuclear refueling operations. 2. Purify the water in the pool to control the radiation levels. 3.Transferring water between various compartments for nuclear fuel handling operations. 4. Supplying make-up water into the nuclear fuel storage pool 190 or reactor room 155 in the event of failure of the alternative cooling system. The design of the nuclear fuel storage system and the nuclear fuel handling system in the low pressurized water reactor of the present invention is the same as that of the conventional high pressurized water reactor.

[0062] As shown in Figure 5, to achieve the redundancy and diversity discussed in the DHRS, in one embodiment, the FPCS is similar to the DHRS, but with two main operational banks (hereafter also referred to as the "M bank") rejecting heat to the CCWS / SWS heat sinks and one backup bank (hereafter also referred to as the "B bank") rejecting heat to the various DCS / DHS heat sinks. Each bank of the FPCS is designed to keep the reactor coolant well below 100°C at the maximum heat load due to the decay heat generation of the spent nuclear fuel.

[0063] The wall penetrations of the fuel pool 190 and reactor pool 155 (which are connected to the FPCS intake and exhaust piping networks) are located at high elevations. As a result, a break in the FPCS piping network will not directly result in exposure of the fuel in the storage shelf 191. The penetrations connected to the FPCS M-row are located above the penetrations of the B-row. As a result, the B-row is available even after a partial pool drain due to a piping break in the M-row. In the event of a total failure of the FPCS row, pool cooling is still possible because makeup water from the water tank 200 is gravity fed to replace losses due to evaporation in the pool.

[0064] Control and Protection Systems A safety-classified control and protection system (C&PS) drives the safety features, including the reactor trip system, flood valve 130, steam dump valve 131, and decay heat removal system. These mechanisms are activated when an abnormality in the plant operation is detected or by the operation of an operator in the control room. Since the low pressurized water reactor 1 design of the present invention has a relatively small number of active safety features, the C&PS of the present invention is involved in a much smaller number of safety functions than that of conventional high pressurized water reactor designs, whose safety features include hundreds of safety features.

[0065] The C&PS of the low pressurized water reactor 1 of the present invention consists of a main (M) system and an independent backup (B) system. The M system drives the first-line safety equipment. The B systems are manufactured with various designs and operate various backup safety mechanisms (e.g., B group trains and B generators of DHRS). Both the main and backup C&PS are classified as nuclear safety protection equipment.

[0066] Emergency Power System Under normal circumstances, power to the plant equipment is supplied from the external power grid. In an emergency, the external power grid is considered not to be classified as a safety protection equipment (cannot be used). Therefore, an emergency generator is required to supply power to operate the safety devices.

[0067] The emergency power system for a low pressure water reactor 1 of the present invention, in one embodiment, consists of two main diesel (MD) generators and one backup diesel (BD) generator. The former operate the first line safety equipment, the latter are manufactured in various designs and operate various backup features. A fourth D generator is used during maintenance operations. An uninterruptible battery backup power supply provides temporary power and operation to the plant control system in the event of loss of all emergency generators.

[0068] Upon loss of off-site power (from the off-site grid), the reactor automatically shuts down and the flood valve 130 and steam dump valve 131 automatically open. As a result, the reactor vessel 100 is under control for at least 24 hours before boiling begins in the water reservoir 200 or the fuel storage pool 190. Long-term cooling is established by operating the DHRS & FPCS after restoration of the power supply. If this is not possible, cooling can be maintained indefinitely by providing make-up water to the water reservoir 200 to replace water lost as a result of evaporation.

[0069] Only the ESF requires a large amount of power in the water tank 200, and therefore belongs to the cooling train of the DHRS and FPCS (operating at low pressure). The power requirement of the emergency D generator in the water tank 200 of the low pressurized water reactor 1 of the present invention is much smaller than that of the conventional high pressurized water reactor. The conventional high pressurized water reactor has many ESFs built-in and requires multiple high pressure, high current pumps for emergency core cooling. Considering the energy requirement to operate the cooling train of the DHRS and FPCS, the power required for the D generator of the low pressurized water reactor 1 of the present invention is only about 100 kW per unit.

[0070] Core Design The low pressurized water reactor 1 of the present invention uses the same nuclear fuel as conventional high pressurized water reactor designs, so the extensive operating experience can be applied to this type of nuclear fuel. 235 It uses UO2 fuel with a maximum enrichment of 5 w%. The UO2 fuel is in the form of cylindrical pellets housed in a 9.5 mm diameter tube clad in M5 zirconium alloy. The tube is pressurized with helium gas. The fuel assembly is approximately 4.8 m long with a fuel section length of 4.2 m. The fuel assembly has 265 fuel rods and 24 guide tubes. The guide tubes are used to house absorber rods or in-core instrumentation. The EPR core contains 241 fuel assemblies.

[0071] 6 shows an embodiment of a core 101 of the invention containing 145 fuel assemblies of the EPR type, specifically selected to achieve a target power of 2000 MW(th).

[0072] According to one embodiment, the main dimensions and material properties of the nuclear fuel rods, fuel assemblies and core 101 envisaged for the design of the low pressurized water nuclear reactor 1 of the present invention are summarized in Table 1. TIFF2024542300000002.tif149165Table 1: Dimensions and physical properties of fuel assemblies and cores of low pressure water reactors The following describes an analysis of the steady-state performance characteristics of a power plant in which the low pressurized water reactor 1 of the present invention provides the steam source.

[0073] thermal hydraulic design As mentioned above, the low pressurized water reactor 1 of the present invention is a light water cooled moderated reactor, which operates at near atmospheric pressure and effectively eliminates the risk of accidents that could lead to a large release of radioactivity into the environment. The increased safety is achieved by locating the core 101 at the bottom of a long cylindrical body. This body is located deep underground and is connected to a large water tank 200 above ground via a flood valve 130 and a steam dump valve 131. This allows the core 101 to be cooled in all foreseeable faults and dangerous conditions. In particular, due to the combination of low operating pressure and the storage of a large amount of primary coolant, the low pressurized water reactor 1 of the present invention effectively eliminates the risk of nuclear fuel overheating in the event of a loss of coolant accident or the subsequent loss of the heat removal system during normal operation. The design objective of the low pressurized water reactor 1 of the present invention is to keep the capital cost per megawatt (MW) of electricity generated low. This design objective is achieved by eliminating the need for ESF or operational safety features (eg, the cost of a sealed containment building, which dominates the capital costs of conventional high pressure water reactor designs).

[0074] 7 illustrates another design concept of the low pressurized water reactor 1 of the present invention. As mentioned above, during reactor operation, the upper portion of the ascending tube 104 above the core 101 contains two-phase steam-water flow produced by flash boiling, while the core 101 itself is cooled with sub-cooled pressurized water.

[0075] Preliminary steam circulation analysis Conventional water-cooled nuclear reactors are designed to operate at high pressures in order to achieve high thermal efficiency. In contrast, in the low pressurized water reactor 1 of the present invention, low thermal efficiency is theoretically not a big problem. This is because the price of uranium fuel is relatively low. However, the economic feasibility of the low pressurized water reactor 1 of the present invention depends on achieving sufficient thermal efficiency. A preliminary evaluation of thermal efficiency is described below.

[0076] The low pressurized water reactor 1 of the present invention, in one embodiment, uses a modified Rankine Cycle. The basic Rankine cycle includes a boiler, a turbine (hereafter simply referred to as the "turbine") 161 powered by low pressure steam, a condenser 162, and pump components. Heat is fed to the boiler, which transfers heat to a working fluid (usually water) to produce steam. The pressurized steam is used to drive the turbine 161, which produces useful work. The steam leaving the turbine 161 is cooled and returned to a liquid state in the condenser 162, where heat is rejected to an external heat sink (river or seawater). The condensed liquid (coolant) leaving the condenser 162 is returned to the original operating pressure by a feed pump.

[0077] Evaluate the thermal efficiency of the system operating at various turbine inlet pressures. The starting point is a simple Rankine cycle. Consider a condensation temperature of 25°C, i.e. T3 = T4 = 298K. At the outlet of the condenser, the condensing pressure is increased by the feed pump to the system pressure P sys Compression causes a slight increase in the fluid temperature at the boiler inlet, assuming isentropic flow.

[0078] The enthalpy and entropy of the fluid at the inlet of the feed pump are the saturated liquid enthalpy at the condenser outlet temperature (298 K) and are obtained from equations (1.1) and (1.2). Compression is assumed to be isentropic so that the enthalpy at the outlet of the feed pump is given by (1.3), which is taken from the Standard Steam Tables (www.thermopedia.com). TIFF2024542300000003.tif41121

[0079] The fluid at the outlet of the feed pump is heated and vaporized by thermal energy from the core 101, eventually reaching a saturated steam state in the turbine 161. The steam enthalpy, temperature, and entropy at the inlet of the turbine 161 are obtained from standard steam pressure tables. TIFF2024542300000004.tif57161

[0080] The fluid expands at the condensing temperature in turbine 161. In one embodiment, the expansion is assumed to be isentropic and the inefficiencies of turbine 161 are ignored. The steam quality at location 3 in Figure 7 is represented by x3 in equation (1.7). TIFF2024542300000005.tif25170 From the steam quality at the inlet of the condenser 162, the enthalpy at position 3 is expressed as X3 in equation (1.8). The amount of work done by a system per kg of working fluid, evaluated in terms of enthalpy, is given by equations (1.9), (1.10), and (1.11). Therefore, the thermal efficiency of this cycle is expressed by equation (1.13). TIFF2024542300000006.tif49121 The heat input per kg to the system is the total enthalpy change between the outlet of the feed pump 164 and the inlet of the turbine 161. It is expressed by equation (1.12). The thermal efficiency of the cycle is given by (1.13). TIFF2024542300000007.tif29117 The values ​​of h1, h2, h3, and h4 are obtained from equations (1.3), (1.4), (1.8), and (1.1), respectively. The efficiency of the cycle is then obtained for various values ​​of system pressure.

[0081] The expansion from the inlet of the turbine 161 to the inlet of the condenser 162 is assumed to be isentropic. In reality, however, this expansion results in an increase in entropy. Therefore, the enthalpy at position 3, the inlet of the condenser 162, is increased by the isentropic efficiency factor η s By including and evaluating, we obtain (1.14). s This is a typo. Then the enthalpy is given by (1.15). TIFF2024542300000008.tif55169 Therefore, the thermal efficiency in the non-isentropic case is obtained from (1.13) as (1.16). Here, from (1.9) and (1.10), we get (1.17). TIFF2024542300000009.tif43170

[0082] Various system pressures P sys From the Rankine cycle calculation, the thermal efficiency of the power plant decreases as the system pressure decreases. According to one embodiment, the steam cycle calculation is calculated from (1.13)(1.16). This can be calculated in Microsoft Excel, and the analysis of how the thermal efficiency relates to the system pressure is shown in Figure 8.

[0083] From the above relationship, the thermal efficiency of the power plant decreases significantly as the system pressure decreases. However, surprisingly, the thermal efficiency at 1 bar is half that at 50 bar (nominal steam pressure of conventional high pressurized water reactor (PWR)). This means that the low pressurized water reactor (LPWR) of the present invention can obtain a larger economical power output than the conventional high pressurized water reactor.

[0084] The steam cycle of the low pressurized water reactor 1 of the present invention is slightly different from the Rankine cycle described above. In the inlet plenum, the fluid is sub-cooled water, which has an enthalpy approximately equal to that of the saturated water in the drum 150. This saturated water comes from the top of the descending pipe 105. This sub-coolant enters the core 101 and is heated, but remains in a sub-coolant state at the reactor outlet. As the water enters and rises in the ascending pipe 104, the pressure decreases due to a decrease in hydrostatic pressure, eventually reaching the saturation pressure at the onset of flashing. At the onset of flashing, nucleate boiling begins and two-phase flow is generated and flows from the ascending pipe 104 into the drum 150. Phase separation occurs in drum 150 and the saturated steam is sent to turbine 161 via main steam line 173 and expanded to sub-atmospheric pressure in condenser 162. The saturated condensate leaves condenser 162 and is pre-pressurized to drum 150 pressure by feedwater pump 164. The water is then mixed with the saturated water in drum 150. It is then returned to the inlet plenum via downcomer 105.

[0085] In the low pressurized water reactor 1 of the present invention, steam is generated by flashing in the ascending pipe 104 and fed to the turbine 161. The generation of a two-phase condition in the ascending pipe 104 increases the driving head of the net pressure, resulting in a natural circulation flow in the descending pipe 105 and the ascending pipe 104. This results in a mass velocity in the core 101 that is much greater than the natural circulation. The mass velocity in the core 101 is comparable to that achieved in conventional high pressurized water reactors that utilize pumped flow. Preliminary calculations have been performed for the flow velocities achievable in the low pressurized water reactor 1 of the present invention for various power levels and core depths. This was done using the model of FIG. 9 and the results are shown in FIG. 8.

[0086] In the above system, an iterative calculation was performed using the one-dimensional energy and momentum conservation equations in the loop to calculate the steady state natural circulation flow rate W. The drum pressure p2 and the core power Q were taken as boundary conditions.

[0087] The loop flow velocity W is first estimated, and the core exit enthalpy is calculated from the energy balance using (1.18). TIFF2024542300000010.tif13114 h fSAT is the saturated liquid enthalpy at drum pressure. This is given by the law of conservation of energy and is equal to the enthalpy of the liquid entering the reactor core 101. h'2 is the enthalpy at the core outlet. h″2 is the liquid enthalpy at the flashing point. where we neglect heat losses from the loop and small kinetic and potential energy terms.

[0088] Using the estimated loop flow, pressure drop measurements were made for three regions: 1. Descending canal 105. 2. A single phase region of the ascending pipe 104 between the ascending pipe 104 inlet and the flushing location. 3. The two-phase region of the ascending tube 104 above the flashing point. For each of these regions, the pressure loss coefficient K is calculated, which is defined in (1.19). TIFF2024542300000011.tif23149 ρ is the average density in the two-phase region. The pressure loss coefficients in these three regions are given by (1.20), (1.21), and (1.22). TIFF2024542300000012.tif50116 Here, C IN and C. OUT refers to the pressure loss coefficient due to expansion or contraction at the inlet and outlet of the component. The subscript DC refers to the descending pipe 105, the subscript C refers to the core 101, the subscript R refers to the ascending pipe 104, H core is the core height, K in indicates the coefficient of flow resistance 102 at the inlet of the core (introduced to achieve stability). Pressure loss due to wall friction is ignored in the ascending pipe 104 and the descending pipe 105 (except for the core 101). This is because the pipe diameter is large and the pressure loss is negligible.

[0089] Pressure at the flashing point p " 2 is the core exit enthalpy h expressed by (1.18) " is the saturation pressure corresponding to 2. The pressure at the core outlet is calculated by applying the momentum equation between the downflow pipe 105 inlet and the core 101 outlet (location 2'). TIFF2024542300000013.tif25132 The resistance coefficient is given by (1.20)(1.21). The liquid density near the loop, ρ f Changes in are ignored.

[0090] Using (1.23), the height of the flashing point above the core inlet H tp is obtained from the momentum balance between locations 2' and 2". Pressure losses due to wall friction are neglected. TIFF2024542300000014.tif17168 Pressure at the flashing point p " 2 is the core exit enthalpy h expressed by (1.18) "is the saturation pressure corresponding to 2. This can be obtained from a vapour pressure table.

[0091] Flow quality of ascending pipe 104 outlet out,R can be determined from the energy balance of the two-phase region. TIFF2024542300000015.tif11123 h gSAT and fSAT are the saturation enthalpies of the vapor and water phases, respectively. These can be estimated by the saturation value at the drum pressure.

[0092] X out,R Given, the void fraction and two-phase density at the exit of the ascending pipe 104 can be obtained assuming homogeneous two-phase flow. TIFF2024542300000016.tif31141 ρ gSAT and ρ fSAT are the saturation densities of the vapor and water phases. These can be estimated by their saturation values ​​at the drum pressure.

[0093] From (1.27), the two-phase region in the ascending tube 104 can be approximated as (1.29). TIFF2024542300000017.tif37156 Finally, the loop momentum balance is found to deduce a second equation for the height of the flushing point, which corresponds to an estimate of the loop flow velocity W. The loop flow rate W corresponds to the core power and drum pressure, but the location of the flashing point H is given by (1.24)(1.29). TP This is obtained by adjusting the estimate of W until

[0094] Description of the steady-state core model In particular, for the core model of the low pressurized water reactor 1 of the present invention, the heated core 101 is represented by an open array of fuel pins. Between the fuel pins are subchannels formed by the flow paths between them. A representative subchannel is shown in FIG. 10. Two representative subchannels are modeled in the core 101. These are shown in FIG. 11 for the subchannels at peak and average power. The peak subchannel corresponds to the subchannel close to the highest rated fuel pin, while the average subchannel represents the condition of the other fuel pins in the core. The core model calculates the fluid enthalpy, pressure and density, and temperature of the fuel rods in each axial cell in the subchannels at both peak and average power, given the boundary conditions of rod power, mass flow rate, pressure and enthalpy of the coolant at the subchannel inlet. The governing equations of the model are described below.

[0095] Calculation of fluid conditions in the reactor core The inlet mass flux is assumed to be uniform across all of the core subchannels. However, in reality, in the low pressure water reactor 1 of the present invention, the fluid density is low in the subchannels at high power (peak), so the driving head pressure is somewhat high due to natural circulation, and the inlet flow W of the subchannel at peak power is c,PK is the mean time subchannel inlet flow W cAV Therefore, the uniform flow assumption is conservative for calculating the margin to the thermal limit in the peak subchannel.

[0096] Assuming uniform flow, we get: TIFF2024542300000018.tif53152

[0097] The axial power distribution is assumed to be the same at the radial position of the nuclear fuel rod. The axial power distribution in both the peak and average subchannels is assumed to be described by the symmetric cosine distribution (1.31) that applies to an empty cylindrical reactor, ignoring the effect of temperature on the local core power. TIFF2024542300000019.tif47148

[0098] For the mean subchannel, the peak heat flux is given by the axial peaking factor of the core: TIFF2024542300000020.tif28130 The average fuel pin heat flux through the core is the total core power divided by the total heat transfer area. TIFF2024542300000021.tif38123 For the peak subchannel, the maximum heat flux is given by (1.34). TIFF2024542300000022.tif15125 Here, F ΔHis the so-called hot channel coefficient. It is defined as the ratio of the enthalpy rise in the hot subchannel to the average enthalpy in the core. It is assumed that the boundary conditions are known.

[0099] Using (1.31), the average heat flux in the mean-time subchannel is given by (1.35). TIFF2024542300000023.tif31150 Comparing with (1.32), ε and P f is related as shown in (1.36). TIFF2024542300000024.tif27121 The enthalpy of the coolant at the centre of the nth cell of any subchannel at peak or average time is given by (1.37). This can be expressed as (1.38) using (1.31). Using the enthalpy obtained from TIFF2024542300000025.tif42166(1.38), the local flow quality X g is obtained using (1.40) for the condition when two-phase flow exists in the subchannel. TIFF2024542300000026.tif13123 Here, h g,sat and f,sat are the saturation enthalpies of the vapor and water phases, respectively, at the local pressure.

[0100] Using the flow quality, the coolant density at the center of the cell, ρ c,n is determined by the following equation (1.41). Flow quality is expressed in terms of local void (void, defect) fraction (Non-Patent Document 16) TIFF2024542300000027.tif61135

[0101] Void fraction is flow quality x gn is obtained using the drift flux correlation from: For the analysis of the core 101, the widely used drift flux correlation

[17] , which is for churn-tubulant two-phase gas liquid flow in a channel, is applied, resulting in (1.42). TIFF2024542300000028.tif86169Allowances are provided in the core model for the pressure drop in the subchannels due to wall friction and gravity. The pressure in the nth cell is calculated from (1.43). TIFF2024542300000029.tif65170P n is given, the saturated phase density and enthalpy of (1.40)(1.41) can be obtained from the standard vapor pressure table. In particular, (1.38)(1.43) can be solved by the marching method (in both peak and average subchannels), starting from the inlet of the core.

[0102] Nuclear fuel temperature calculation In this invention, to calculate the temperature of UO2 and the nuclear fuel rod cladding in each core cell, the nuclear fuel rod was treated as a cylinder containing UO2 nuclear fuel pellets concentrically. A one-dimensional steady-state heat conduction equation in circular coordinates was applied to determine the radial temperature distribution in the pellets and the cladding, with the following assumptions: The outer surface of the cladding is assumed to be at the local coolant temperature in each cell. This assumption is reasonable due to the high thermal transfer coefficients achieved by forced convection boiling in water-cooled reactors. The calculations of temperature drop across the cladding and in the gas gap (the gap between the cladding and the nuclear fuel pellets) were performed assuming that the thermal conductance values ​​of the cladding and the gas gap (given as input data) are constant. The values ​​used for the cladding are given in Table 2 of paragraph 0114. The values ​​used for the conductance of the gas gap are described below. A uniform volumetric heat source due to fission and fission product decay is assumed to be within the fuel pellet. Heat generation within the cladding is neglected.

[0103] With the above assumptions, from the steady-state solution of the heat conduction equation in a cylindrical geometry (Non-Patent Document 18), the radial temperature gradient in the nuclear fuel pellet is given by (1.44). TIFF2024542300000030.tif59167

[0104] The local volumetic heat source within the pellet at height z of the core 101 is related to the local pin surface heat flux given by (1.31) and (1.45): TIFF2024542300000031.tif15138 The thickness of the cladding and the thickness of the gas void are small compared to the diameter of the nuclear fuel rod, so that the curvature effect is negligible in determining the temperature drop between the outer pellet surface and the outer cladding surface. Assuming one-dimensional steady-state heat conduction through the cladding and the gas gap, the outer surface temperature of the cladding is the local fluid temperature T f Assuming that the temperature of the outer surface of the pellet is close to T po is given by the following equation (1.46), which is obtained by equating the heat flowing into the coolant per unit length with the heat generated within the pellet per unit length. TIFF2024542300000032.tif73169

[0105] The relationship between the center and the outer surface of the pellet (1.44) is T po Using the above equation (1.46), the pellet temperature at radial location r becomes (1.47). From (1.47), the average temperature inside the pellet is (1.48). The average temperature in the gas cavity is (1.49). TIFF2024542300000033.tif73157 The average pellet temperature, i.e., the left hand side of (1.48), is calculated for each cell in the average subchannel, and the average temperature of UO2 in the core 101 is obtained from (1.50). TIFF2024542300000034.tif40168

[0106] Similarly, the average density of the coolant in the core 101 (ie, the average moderator temperature) is obtained by taking the average over all the cells in the mean time subchannel. Here, the coolant density in the cell is ρ cj is obtained from (1.41) for the average time subchannel. TIFF2024542300000035.tif23122

[0107] Modelling the conductance of gas gaps. The temperature of UO2 is affected by the thermal resistance (mainly dependent on the gas and thermal conductivity (conductance) and gap width) of the gap between the fuel pellets and the cladding (both of which constantly fluctuate during reactor operation). A method for determining the gap thermal conductivity in the fuel pin of a low-pressurized water reactor (Non-Patent Document 19) states that the fuel pin of a low-pressurized water reactor (1) is usually pre-compressed with helium gas and mixed with other gases (especially xenon gas) that are produced during the fission process during the life of the core (101). According to Non-Patent Document 19, the thermal conductivity of the filling gas varies greatly depending on the temperature and gas composition. Furthermore, the gap width decreases at the operating pressure and temperature of the reactor due to the external pressure load on the fuel cladding and the thermal expansion of the fuel pellets. Initially, when the fuel pin is cold, the gas gap width is 1-2% of the diameter of the fuel pin, but under reactor operation conditions, the cladding comes into contact with the fuel pellets and the gap width is determined by the surface roughness of the fuel pellets.

[0108] The low pressurized water reactor 1 of the present invention operates at near atmospheric pressure. The internal pin pressure within the nuclear fuel of such a reactor may be adjusted downward (lowered) to reduce the pressure load on the nuclear fuel. Therefore, a "closed gap" condition continues to exist when the reactor is in an operational state. This is due to the thermal expansion of the nuclear fuel relative to the cladding. In one embodiment, the gap conductance constant C gap =1.5x10 4 W / m 2K is adopted in the present analysis. This value is typical for conventional high pressurized water reactor nuclear fuel at normal operating temperature and pressure when the gap width is small (Non-Patent Document 19). The average nuclear fuel temperature of the present low pressurized water reactor is 200°C lower than that of the conventional high pressurized water reactor because the coolant temperature is lower. According to the equations in Non-Patent Document 19, this reduces the electrical conductivity of the fill gas by 20% and the gap conductance (thermal conductivity) by the same amount. However, this correction is small compared to other uncertainties in the model and can be neglected.

[0109] Calculating Segregation from Nucleate Boiling Ratio Critical Heat Flux is an important safety limit in the operation of water-cooled nuclear reactors. If the local heat flux at the fuel surface exceeds the CHF, the heat transfer regime passes from the nuclear boiling regime to the film boiling regime, resulting in a significant decrease in the heat transfer coefficient and an increase in the temperature of the cladding. This can result in structural failure of the cladding and / or the release of radioactive fission products into the reactor water network. The margin for CHF at a given reactor state is expressed as the minimum separation from the nuclear boiling ratio (DNBR) reached at any point in the core 101. It is defined by (1.52). TIFF2024542300000036.tif39165CHF depends on the local thermal-hydraulic conditions (flow quality and mass velocity, etc.) in the subchannel. Therefore, to find the minimum value of DNBR associated with a particular core state, the DNBR in (1.52) is evaluated at each core cell in the subchannel at peak power.

[0110] The value of CHF in a fuel pin assembly is specific to the specific fuel geometry, power configuration, and the presence of flow obstacles (e.g., spacer grids). Therefore, accurate prediction of CHF can only be obtained by testing prototype models of real fuel assemblies using electrically heated fuel pin simulators. However, a number of generic engineering correlations have been developed to estimate CHF values ​​in channels and rod bundles. In the absence of a specific fuel model, a widely used correlation in the Croeneveld method for CHF in tubes and rod bundles has been used in the present low pressurized water reactor core to calculate DNBR (Non-Patent Document 20). The Groeneveld correlation is in the form of a three-dimensional table in which the CHF values ​​are listed in the form of parameters. P = pressure (N / m 2 ) G=mass velocity (kg / m 2 / sec) X=thermodynamic quality=(hh f,sat ) / (h g,sat -h f,sat ) The scope of the amendment in Non-Patent Document 20 is as follows: P = 1 to 200 bar, G=0~7500kg / m 2 / sec, X=-0.5~10. These cover the low pressure operating conditions of the low pressurized water reactor of the present invention.

[0111] The table in Non-Patent Document 20 was copied into Microsoft Excel, an Excel function was created, and CHF was calculated for given values ​​of P, G, and X by linear interpolation from the table.

[0112] A CHF correction factor proposed by Croeneveld et al. can be applied to the tabulated CHF values ​​to correct for the effects of spacer grids, rod bundle geometry, etc. However, this correction was not applied in the present analysis.

[0113] Data analysis of the CHF table proposed by Croeneveld et al. predicted higher CHF margins at lower pressures (other parameters being the same). Therefore, the DNBR values ​​for the low pressure pressurized water reactors of the present invention are predicted to be significantly higher than for European pressurized water reactors.

[0114] Calculation of core reactivity The core reactivity is defined in (1,29), and in one embodiment is calculated by the low pressure water reactor core model of the present invention using predefined reactivity coefficients. The reactivity is given by (1.53). TIFF2024542300000037.tif13128 Here, ΔR is the change in reactivity relative to the initial reference state. The initial reference state is when the reactor is in a critical state (IR=0, k eff =11). The four left-hand terms in (1.54) are, from top to bottom, the contributions of the following items to the change in reactivity from the reference state: *Nuclear fuel temperature, *Moderator (coolant), *Boron concentration, *Control rod position. The reactivity contribution in (1.53) can be obtained from (1.54). TIFF2024542300000038.tif31137 Where: TIFF2024542300000039.tif58126 The average UO2 temperature and average moderator density in the core are calculated in the core model using (1.48) and (1.51), respectively. The boric acid concentration in the coolant in the core is given as a boundary condition of the model. The reaction coefficients used in the low pressurized water reactor analysis of the present invention are selected from Table 2 below. TIFF2024542300000040.tif39149

[0115] In one embodiment, to calculate the thermodynamic and nuclear reaction states of the core, the fluid pressure, enthalpy, and nuclear fuel temperature are obtained for each cell, for both peak and average subchannels, in a marching manner, starting from the inlet of the core. The boundary conditions for the calculation are the core power, the core inlet mass flow rate (obtained specifically from (1.55)), the core inlet enthalpy, and the core inlet pressure. TIFF2024542300000041.tif19159The input data for the nuclear fuel geometric properties, thermal properties, and axial and radial power distributions are given in Tables 1 and 3. TIFF2024542300000042.tif74133 Many calculations have been performed in Microsoft Excel, and in all calculations the core 101 is divided axially into 20 sections. cell =20. [Prior art documents] [Non-patent literature]

[0116] TIFF2024542300000043.tif102134TIFF2024542300000044.tif112134

[0117] The bar used in this specification is absolute pressure. The above description relates to one embodiment of the present invention, and those skilled in the art may conceive of various modifications of the present invention, all of which are within the technical scope of the present invention. The numbers in parentheses following the components of the claims correspond to the part numbers in the drawings, and are provided for easy understanding of the invention, and should not be used to interpret the invention in a restrictive manner (Article 24-4 of the Enforcement Regulations of the Patent Law and Item 14-B of "Notes" in Form 29-2). In addition, even if the part numbers are the same, the part names in the specification and the claims are not necessarily the same. This is for the reasons described above. "At least one or more" and "and / or" are not limited to one of them. For example, "at least one of A, B, and C" may include not only "A", "B", and "C" alone, but also multiple ones such as "A, B, or B, C, and also A, B, and C". "At least one of A, B, and C" may be not only A, B, and C alone, but also a combination of A and B, A, B, and C. "A, B and / or C" may include not only A, B, or C alone, but also two of A and B, or all of A, B, and C. In this specification, "including A" or "having A" may include things other than A. Unless otherwise specified, the number of devices or means may be singular or plural. [Explanation of symbols]

[0118] 1: Low pressurized water reactor vessel 100:Reactor vessel 101: Core 104: Ascending canal 105: Descending pipe 150: Steam drum 151: Upper lid 152: Splash board 161: Low pressure steam turbine 200: Water tank

Claims

1. In a low-pressure water reactor vessel, it has a reactor vessel (100), a steam drum (150), a water storage tank (200), an automatic water injection system, and a low-pressure steam turbine (161), * The reactor vessel (100) has an upward pipe (104), a reactor core (101) disposed below the upward pipe (104), and an outer annular body surrounding the reactor vessel (100), The upward pipe (104) forms a downward pipe (105), and the downward pipe (105) has an annular portion with a predetermined flow area and generates a flow velocity of 10,000 kg / sec or more, The reactor core (101) is disposed at a depth of at least 55 m underground and has an internal pressure between 6 - 15 bar, * The steam drum (150) is connected to the upper end of the upward pipe (104), is disposed on the ground, and has an internal pressure between 1 - 10 bar, * The water storage tank (200) is connected to the reactor vessel (100) and stores boric acid water, * The automatic water injection system injects the boric acid water from the water storage tank (200) into the reactor vessel (100), * The low-pressure steam turbine (161) operates at a pressure of 1 - 10 bar, The reactor vessel (100) heats water to a predetermined temperature without reaching the saturation point, The upward pipe (104) converts the heated water into steam, and this steam is further sent to the low-pressure steam turbine (161) via the steam drum (150), whereby this conversion into steam generates a density difference of the primary coolant, and this density difference initiates natural circulation driven by the density of the primary coolant in the upward pipe (104), downward pipe (105), steam drum (150), and reactor core (101). A low-pressure water reactor vessel characterized by the above.

2. The steam is generated by flashing means in the upward pipe (104). The low-pressure water reactor vessel according to Claim 1, characterized by the above.

3. The reactor core (101) contains 120 - 180 nuclear fuel assemblies. The low-pressure water reactor vessel according to Claim 1, characterized by the above.

4. The steam drum (150) has a dryer, an upper lid (151), and a splash plate (152), and is disposed in a watertight room. The low-pressure water reactor vessel according to Claim 1, characterized by the above.

5. The downward pipe (105) has a concrete cylindrical body lined with steel. The low-pressure water reactor vessel according to Claim 1, characterized by the above.

6. The water storage tank (200) is connected to the reactor vessel (100) via an automatic valve. The low-pressure water reactor vessel according to claim 1, characterized in that.

7. The low-pressure water reactor vessel (1) is stopped by either a shutdown rod or the water storage tank (200). The low-pressure water reactor vessel according to claim 1, characterized in that.

8. The boric acid water stored in the water storage tank (200) contains 1000 ppm of boric acid. The water storage tank (200) is arranged at a height of 20 m above the ground. The low-pressure water reactor vessel according to any one of claims 1-7, characterized in that.

9. In a method for controlling a low-pressure water reactor vessel. The low-pressure water reactor vessel has a reactor vessel (100), a steam drum (150), a water storage tank (200), an automatic water injection system, and a low-pressure steam turbine (161). * The reactor vessel (100) has an upward pipe (104), a reactor core (101) disposed below the upward pipe (104), and an outer annular body surrounding the reactor vessel (100). The upward pipe (104) forms a downward pipe (105) having a predetermined flow area and generating a flow velocity of 10,000 kg / sec or more. The reactor core (101) is disposed at a depth of at least 55 m underground and has an internal pressure between 6 and 15 bar. * The steam drum (150) is connected to the upper end of the upward pipe (104), is disposed on the ground, and has an internal pressure between 1 and 10 bar. * The water storage tank (200) is connected to the reactor vessel (100) and stores boric acid water. * The automatic water injection system injects the boric acid water from the water storage tank (200) into the reactor vessel (100). * The low-pressure steam turbine (161) operates at a pressure of 1-10 bar. (A) A step of heating water in the reactor vessel (100) to a predetermined temperature without reaching the saturation point. (B) A step of converting the water heated in step (A) into steam in the upward pipe (104). The steam generated in step (B) is further sent to the low-pressure steam turbine (161) via the steam drum (150). In step (B), a density difference of the primary coolant is generated, and this density difference starts natural circulation due to the density drive of the primary coolant in the upward pipe (104), downward pipe (105), steam drum (150), and reactor core (101). A method for controlling a low-pressure water reactor vessel, characterized by the following.

10. The inlet temperature of the reactor core (101) is between 100°C and 150°C, The outlet temperature of the reactor core (101) is between 140°C and 170°C, The inlet resistance of the reactor core (101) attenuates static instability, The steam outlet temperature of the steam drum (150) is between 100°C and 170°C The method according to claim 9, characterized by the above.

11. (C) A step of generating steam by flashing in the riser pipe (104) further comprising The method according to claim 9, characterized by the above.

12. The reactor core (101) contains 120 to 180 nuclear fuel assemblies, (D) A step of lifting the reactor core (101) out of the reactor vessel (100) during the nuclear fuel replacement operation of the low-pressure water reactor vessel (1) comprising The method according to claim 9, characterized by the above.

13. The boric acid water stored in the water storage tank (200) contains 1000 ppm of boric acid, The water storage tank (200) is arranged at a height of 20 m above the ground The method according to claim 9, characterized by the above.

14. The water storage tank (200) is connected to the reactor vessel (100) via an automatic valve or a manual valve, (E) A step of controlling the automatic valve when the concentration of boric acid in the primary coolant reaches a predetermined value further comprising The method according to claim 9, characterized by the above.

15. During the stable state operation of the low-pressure water reactor vessel (1), while generating steam with the primary coolant, the reactor core (101) operates with a single-phase primary coolant The method according to any one of claims 9 to 14, characterized by the above.

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