KR20100078200A - Nuclear fuel rod including nuclear fuel cladding increased corrosion resistivity - Google Patents

Abstract

PURPOSE: A nuclear fuel rod including a nuclear fuel cladding increased corrosion resistivity is provided to reinforce erosion resistance of a cladding material. CONSTITUTION: A nuclear fuel rod includes a fissile material, a zirconium based fuel cladding, and film evenly covering on the surface of the fuel cladding. The film prevent hydrogen overpotential by lowering a hydrogen overvoltage of the fuel cladding. The film is comprised of a conductive material having the hydrogen overpotential of 0.5V. The film is formed with one of platinum(Pt), gold(Au), palladium(Pd), and rhodium(Rh) or a compound of them.

Description

Nuclear fuel rod including nuclear fuel cladding increased corrosion resistivity}

The present invention relates to a nuclear fuel rod comprising a nuclear fuel cladding having enhanced corrosion resistance, wherein the cladding surface is modified to have a function of suppressing corrosion of the cladding by hydrogen embrittlement, thereby preventing damage to the cladding and enhancing surface stability of the cladding. This can suppress corrosion of the coating material and deposition of corrosion products.

The fuel assemblies used in PWR (Pressurized Water Reactor) are composed of multiple fuel rods. 1 and 2 show schematic diagrams of the fuel assembly and the fuel rods, respectively. As shown in Fig. 2, the nuclear fuel rod has a tubular sheath and a plug for accommodating the fuel pellets. The cladding of the nuclear fuel rod serves to prevent the release of fission products into the coolant, and to prevent the chemical reaction and the contact between the fuel and the coolant. The cladding should have a small neutron absorption cross section to economically use fission fuel material. Moreover, the coating | covering material is essential in corrosion resistance in cooling water reaching 345 degreeC.

Zirconium alloys were developed in 1960 and have been used as cladding materials to date. However, as the operating conditions in the light water reactor are changed to harsh conditions such as high pH operation and high combustion operation, the fuel cladding is ZIRLO (Zr-1Nb-1Sn-0.1Fe) containing Nb in the Zircaloy-4 alloy. Switched to alloy

Starting with the zirconium-based cladding, many techniques have been developed for the cladding alloy composition. Korean Unexamined Patent Publication No. 2001-0047592 (composition of zirconium alloy for niobium-added nuclear fuel cladding), 2000-0056306 (composition of zirconium alloy composition for nuclear fuel cladding and manufacturing method), 2000-0026542 (excellent corrosion resistance and mechanical properties) Zirconium alloy compositions), 1999-0069103 (zirconium alloy compositions having excellent corrosion resistance and high strength), and 1999-0069104 (zirconium alloy compositions having low corrosiveness and high strength) disclose zirconium alloy compositions of nuclear fuel rod cladding.

Since then, Republic of Korea Patent No. 10-0441562-0000 (Jul. 14, 2004), Republic of Korea Patent No. 10-0835830-0000 (May 30, 2008), Republic of Korea Patent No. 10-0461017-0000 (Dec. 1, 2004), Republic of Korea Patent No. 10 Many technologies such as -0461017-0000 (2004.12.01), Korean Patent No. 10-0334252-0000 (2002.04.12), US Patent No. 6,125,161, US Patent No. 5,838,753, US Patent No. 5,230,758, etc. New alloy design techniques have been developed to add microelements mainly to zirconium metals to provide excellent corrosion resistance and mechanical properties.

In addition, Korean Patent No. 10-0709389-0000 (2007.04.12) anticipated to improve corrosion resistance by forming an oxide film on the outside through heat treatment, and Korean Patent No. 10-0060310-0000 (1993.03.12) and Korean Patent No. 10-0076710-0000 (1994.08.27) and Korean Patent No. 10-0076709-0000 (1994.08.27) design a coating material having a separate inner layer to suppress the reaction between the fission material and the coating material inside the coating material. It was. Korean Patent No. 10-2004-0042927 (filed on Jun. 11, 2004) and US Patent No. 10 / 459,777 (2003.06.12) cover the surface of the coating material which suppresses the unsaturation boiling which induces the generation of the crack by polishing the surface defect of the coating material to 0.1 μm or less. The treatment technique has been described.

Meanwhile, 特 開平 11-326573 (published date: November 26, 1999) disclosed in Japan has designed a coating material that improves corrosion resistance by coating precious metals such as platinum and gold on the surface of the coating material. However, this technique simply considered a noble metal film as a physical barrier to hydrogen diffusion, so that an appropriate coating thickness of 10 μm was estimated. In Japan 特 재 平 8-170992 (published: July 2, 1996), a gold coating material was also designed. This technique also considered gold as a physical barrier to hydrogen diffusion.

As mentioned above, the performance improvement of the nuclear fuel rod cladding is limited to the design of zirconium alloys or to the application of precious metal coating as a physical diaphragm for improving the corrosion resistance of the cladding. Thus, a fundamental solution to corrosion, especially hydrogen embrittlement, has not been disclosed. there is a problem.

Originally, zirconium-based alloys form a very stable metal oxide film called zirconium oxide, so there is no need to consider uniform corrosion in which metal surfaces are uniformly corroded in the presence of oxidants such as dissolved oxygen. Titanium-based alloys also form a very stable metal oxide film called titanium oxide and exhibit a very similar corrosion behavior to zirconium.

A condition in which a zirconium-based metal is not stable to corrosion is when exposed to an aqueous solution having a high pH. Zirconium is a metal with a high hydrogen overvoltage, similar to titanium. Therefore, the potential region in which hydrogen adsorbed on the metal surface is stably exists. With this characteristic, when the pH of the exposed aqueous solution increases, the adsorption region of the hydrogen atom, which is the first step of equilibrium reaction between water and hydrogen occurring on the metal surface, 2H ⁺ + 2e = H ₂ (gas), moves toward the cathodic direction. (See Scheme 1 below).

_{H 2 O + e = H ad} + OH - (1)

_{H ad + H 2 O = H} 2 + OH - (2a)

H _ad + 1/2 Zr = 1/2 ZrH ₂ (permeation into zirconium matrix) (2b)

In general, according to the Volmer-Heyrovsky mechanism, the reaction (2b) in which the adsorbed hydrogen atoms penetrate into the metal, the higher the corrosion potential in the cathodic direction, the more difficult the hydrogen bonding reaction (2a), the more hydrogen atoms into the metal. It penetrates well. The above reaction mechanism is shown in detail in FIG. 4. In addition, when dissolved hydrogen is present in excess (˜50 cc / kg), the hydrogen molecule formation reaction (2a) is further suppressed, while the adsorbed hydrogen is further stabilized, and the penetration of hydrogen atoms into the metal rapidly increases. have.

When the reaction (2b) proceeds as a path through which hydrogen atoms penetrate into the coating material, the zirconium metal is changed to an unstable zirconium hydride, which may weaken the physical properties of the coating material. In particular, there is a problem in that a high stress corrosion cracking (SCC) phenomenon is observed under a condition in which stress is applied. Accordingly, in order to suppress corrosion of stainless steel, a primary system structural material, and to reduce stress corrosion cracking (SCC) of heat pipe material in a pressurized water reactor, it is difficult to convert fuel into a long cycle. Increasing concentrations in and around cc H ₂ / kg are being injected. The decrease in the safety against corrosion of the zirconium alloy, which is a coating material, has increased as the dissolved hydrogen concentration and pH increase due to such changes in the reactor cooling water environment.

Accordingly, the present inventors have completed the present invention by considering the environmental change, which is the operating condition of the nuclear power plant, and developing a fundamental coating material improvement method based on the hydrogen embrittlement mechanism.

SUMMARY OF THE INVENTION An object of the present invention is to provide a nuclear fuel rod comprising a nuclear fuel cladding having enhanced corrosion resistance based on a hydrogen embrittlement mechanism.

In order to achieve the above object, the present invention provides a nuclear fuel rod comprising a nuclear fuel coating material enhanced corrosion resistance.

The nuclear fuel rods according to the present invention can fundamentally prevent corrosion of the coating material, thereby enhancing the hydrogen embrittlement corrosion resistance of the coating material even if the thickness and area of the coating material are reduced. Accordingly, the fuel assembly including the nuclear fuel rods may be useful for maintaining and maintaining the economic efficiency and corrosion stability of the nuclear fuel reactor and the coolant system of the nuclear power plant.

Hereinafter, the present invention will be described in detail.

The present invention provides a nuclear fuel rod with enhanced corrosion resistance including a fissile material, a zirconium-based fuel cladding encapsulating the fissile material, and a coating formed uniformly or non-uniformly on the surface of the fuel cladding.

Nuclear fuel rods according to the present invention by performing the coating treatment based on the Volmer-Heyrovsky mechanism shown in the reaction scheme 1 and 4 to provide a fuel rod reinforced corrosion resistance by essentially inhibiting the corrosion of the zirconium-based fuel cladding.

In the nuclear fuel rod according to the present invention, the coating can prevent hydrogen embrittlement by lowering the hydrogen overvoltage of the fuel coating material.

In FIG. 4 and Scheme 1, the hydrogen atoms adsorbed on the surface of the coating material are quickly sent to the reaction (2a) path where hydrogen molecules are generated. The rate determining step of the reaction in which water of FIG. 4 is reduced to hydrogen molecules is a reaction (2a of FIG. 4). In order to accelerate the reaction (2a) and generate hydrogen molecules before the reaction (2b), it is important to lower the hydrogen overvoltage corresponding to the reducing energy. Accordingly, the present invention can accelerate the reaction (2a) by coating a material having a low hydrogen overvoltage on a zirconium fuel coating material having a relatively high hydrogen overvoltage.

In this case, the film used may be a conductive material having a hydrogen overvoltage of 0.5 V or less, and may be any one or an alloy thereof selected from the group consisting of platinum (Pt), gold (Au), palladium (Pd), and rhodium (Rh). Preference is given to using, more preferably platinum (Pt), gold (Au) or alloys thereof.

In the nuclear fuel rod according to the present invention, the coating can prevent hydrogen penetration by using a material having an atomic gap shorter than the atomic gap of zirconium which is a fuel coating material.

The atomic-atomic spacing of the zirconium metal is wide enough for hydrogen atoms to penetrate, so that the reaction is likely to occur before the reaction (2a) occurs. Zirconium metal has HCP (hexagonal close-packed) structure, the average spacing between atoms is 317.9 pm, loose structure compared to other materials. For this reason, the density of zirconium has a low 6.5 g / cm ³ relative to the atomic weight.

Thus, a material having an interatomic spacing of less than 317 pm that is less than 317.9 pm of zirconium atoms is coated on the fuel cladding to reduce the interatomic spacing on the cladding surface, thereby reducing the possibility of hydrogen permeation in the cladding fuel Hydrogen embrittlement of a coating material can be suppressed. Further, the coating reduces surface defects due to corrosion induced by hydrogen embrittlement, and as shown in FIG. 5, unsaturation boiling easily generated in the defective portion and thus suppression of deposition of corrosion products can be expected.

At this time, the film is carbon (C), nitrogen (N), DLC (Diamond like carbon), zirconium carbide (ZrC), zirconium nitrile (ZrN), titanium carbide (TiC), titanium nitrile (TiN) group Preference is given to using one or a mixture of materials selected from them.

The DLC has an interatomic spacing of 159 pm, and when the coating material is coated using the DLC, the penetration rate of the hydrogen atoms is lowered, thereby making it resistant to hydrogen embrittlement. The surface of the zirconium can be similarly treated with zirconium carbide (ZrC density: 6.7 g / cm ³ ) or zirconium nitride (ZrN density: 7.1 g / cm ³ ) with a narrower interatomic gap than zirconium metal (density: 6.5 g / cm ³ ). The effect can be obtained.

In this case, the material to be used for the coating should be selected in consideration of the decrease in heat transfer that may occur after the coating treatment and the neutron absorption of the coating material. Materials with low heat transfer coefficients or large neutron cross-sections should be partially limited in thickness, as they degrade the inherent function of the fuel cladding. Heat transfer coefficients of metals such as zirconium have a value in the range of 100-300 W / m K, whereas DLC is more than 900 W / m K, and has about three times more conductivity than ordinary metals. It is not restricted. Therefore, in the case of DLC, the thickness of the coating can be suppressed hydrogen embrittlement by uniformly coating or dispersing the surface of the coating material up to 100 μm in which the thermal expansion coefficient is not a problem. In addition, zirconium carbide (ZrC) is also easier to transfer heat than zirconium oxide, which is a passivating film component of zirconium, can be easily used in terms of heat transfer. The neutron absorption cross section is about 0.0035 barn (10 ^-24 cm ² ) for carbon, 1.9 barn for nitrogen, and 10 barn for platinum.

In the nuclear fuel rod according to the present invention, the average thickness of the coating can be formed to 1 nm ~ 10 ㎛. The coating can change the electrochemical and adsorption characteristics of the coating material even if a thinner thin film than the conventional physical protective film can increase the resistance to hydrogen embrittlement. At this time, the average thickness of the coating is preferably 1 nm ~ 1 ㎛, more preferably may be 1 nm ~ 100 nm.

Furthermore, the film according to the present invention may be formed on the entire fuel rod from the half point from the top of the nuclear fuel rod.

Recently, in order to improve the economics of nuclear power plants, the fuel has been converted into a high-burning fuel, which increases local unsaturation boiling swelling on the surface of nuclear fuel and increases the amount of corrosion product deposition. In addition, in order to prevent corrosion of the coating material, the concentration of dissolved hydrogen as a reducing agent in the reactor cooling water is increasing to 50 cc / kg. The Yeon group is more likely to boil when the subcooled phenomenon occurs in a high temperature aqueous solution in which dissolved hydrogen is present, since the dissolved hydrogen in the molecular form is more easily converted to gas than water, and the pore generated by boiling ( In the interior of the pore, the dissolved hydrogen gas is concentrated and the concentration of the address gas at the periphery of the pore has been reported [Journal of Nuclear Materials, 354, 163-170, 2006]. Therefore, at the site where unsaturated response occurs due to dissolved hydrogen, there is a region where the hydrogen concentration is higher and lower than the average so that a concentration gradient occurs. Such a gradient of hydrogen concentration can promote the hydrogen embrittlement of the coating by forming an electrochemical local circuit as shown in FIG. 7. As the coolant passes through the reactor fuel, the temperature rises as it moves from the bottom of the fuel assembly to the top, and as a result, unsaturated boiling from the light reactor fuel surface is concentrated in the upper third of the fuel rod.

Thus, the film according to the present invention can be formed on the fuel cell rod 1/2 to the entire point from the top of the nuclear fuel rods to improve the hardness.

In addition, the nuclear fuel rod of the present invention may further include a buffer layer (buffer layer) between the fuel coating and the coating.

The buffer layer may enhance the bonding strength between the zirconium coating material and the coating film and minimize the residual stress generated when the film is formed. In this case, the thickness of the buffer layer is preferably 1 nm ~ 10 ㎛.

Furthermore, the present invention provides a nuclear fuel assembly including a plurality of nuclear fuel rods formed with a coating as described above.

As described above, the coated nuclear fuel rod is chemically prevented from hydrogen embrittlement, thereby improving corrosion resistance due to hydrogen embrittlement, and the fuel assembly including a plurality thereof has excellent corrosion resistance, thereby increasing the replacement period of the nuclear fuel rod, thereby resulting in economical efficiency. Can be increased.

1 is a schematic representation of a typical light reactor nuclear fuel assembly;

2 is a schematic view of the nuclear fuel rods constituting the light reactor reactor fuel assembly;

FIG. 3 is a view showing an encapsulated fuel rod end face (a) and an encapsulation region (b) of the fuel rod; FIG.

FIG. 4 is a view illustrating two reaction pathways for reducing water by oxidation of dissolved hydrogen at the surface of a nuclear fuel cladding composed of a zirconium alloy;

FIG. 5 is a view for explaining a process in which a hydrogen embrittlement coating material expands to generate defects, thereby accelerating boiling; FIG.

6 is a schematic diagram (a) in which microparticles are precipitated by the boiling of the cooling water and a schematic diagram (b) in which the cooling water is caused by dissolved hydrogen at the surface of the nuclear fuel cladding; And

7 is a view for explaining the process of the hydrogen embrittlement reaction is induced by the difference in boiling concentration hydrogen concentration of dissolved hydrogen.

Claims (14)

And a zirconium-based fuel cladding for encapsulating the fissile material and the fissile material, and a nuclear fuel rod with enhanced corrosion resistance including a film formed uniformly or non-uniformly on the surface of the fuel cladding. The nuclear fuel rod of claim 1, wherein the coating prevents hydrogen overpotential by lowering hydrogen overvoltage of the fuel cladding. The corrosion resistant nuclear fuel rod according to claim 2, wherein the coating is a conductive material having a hydrogen overvoltage of 0.5 V or less. The corrosion resistance of claim 2, wherein the coating is any one or an alloy thereof selected from the group consisting of platinum (Pt), gold (Au), palladium (Pd), and rhodium (Rh). Nuclear fuel rods. The corrosion resistant nuclear fuel rod according to claim 2, wherein the coating is platinum (Pt), gold (Au) or an alloy thereof. The corrosion resistant nuclear fuel rod according to claim 1, wherein the coating prevents hydrogen penetration by using a material having an interatomic spacing shorter than the interatomic spacing of zirconium which is a fuel coating material. 7. The corrosion resistant nuclear fuel rod of claim 6, wherein the molecular spacing is less than 317 pm. The method of claim 6, wherein the coating is carbon (C), nitrogen (N), DLC (Diamond like carbon), zirconium carbide (ZrC), zirconium nitrile (ZrN), titanium carbide (TiC), titanium nitrile (TiN) Corrosion resistance-enhanced nuclear fuel rod, characterized in that any one selected from the group consisting of or a mixture thereof. The corrosion resistant nuclear fuel rod according to any one of claims 1 to 8, wherein the coating has an average thickness of 1 nm to 10 µm. 10. The nuclear fuel rod with enhanced corrosion resistance according to claim 9, wherein the average thickness of the coating is 1 nm to 1 m. 10. The nuclear fuel rod with enhanced corrosion resistance according to claim 9, wherein the average thickness of the coating is 1 nm to 100 nm. The corrosion resistant nuclear fuel rod according to any one of claims 1 to 8, wherein the coating is formed on the entire nuclear fuel rod from a half point from the top of the nuclear fuel rod. 9. A corrosion resistant nuclear fuel rod according to any one of claims 1 to 8, further comprising a buffer layer between the fuel cladding and the coating. The nuclear fuel rod of claim 13, wherein the buffer layer has a thickness of 1 nm to 10 μm.

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