阅读说明：本技术 用于核反应堆系统的复合慢化剂 (Composite moderator for nuclear reactor system ) 是由 弗朗切斯科·文内里 保罗·弗朗切斯科·文内里 兰斯·刘易斯·斯尼德 于 2019-01-22 设计创作，主要内容包括：用于核反应堆系统的复合慢化剂介质和制造由所述复合慢化剂介质形成的复合慢化剂块的方法。所述复合慢化剂介质包含两种或更多种慢化剂,例如低慢化材料和高慢化材料。高慢化材料与低慢化材料相比,具有更高的中子减速能力。低慢化材料包含碳化硅或氧化镁的慢化基体。高慢化材料分散在慢化基体内并且包含铍、硼、或其化合物。高慢化材料被包封在低慢化材料内使得高慢化材料不暴露在低慢化材料的外部。所述方法可以包括选择烧结助剂和烧结助剂在复合慢化剂混合物中基于低慢化材料的重量百分比以及火花等离子体烧结。(Composite moderator media for a nuclear reactor system and methods of making composite moderator blocks formed from the composite moderator media. The composite moderator medium includes two or more moderators, such as a low moderating material and a high moderating material. The high moderating material has a higher neutron moderating capability than the low moderating material. The low moderating material includes a moderating matrix of silicon carbide or magnesium oxide. The high moderator material is dispersed within the moderator matrix and comprises beryllium, boron, or a compound thereof. The high moderator material is encapsulated within the low moderator material such that the high moderator material is not exposed outside of the low moderator material. The method may include selecting a sintering aid and a weight percentage of the sintering aid in the composite moderator mixture based on the low moderator material and spark plasma sintering.)

1. A nuclear reactor system, comprising:

a nuclear reactor core comprising:

an array of fuel elements; and

a composite moderator medium formed from two or more moderators;

wherein:

the two or more moderators include a low moderating material and a high moderating material; and

the high moderating material has a higher neutron moderating capability than the low moderating material.

2. The nuclear reactor system of claim 1 wherein:

the low-moderating material includes a moderating matrix of silicon carbide (SiC) or magnesium oxide (MgO); and

the high moderator material is dispersed within the moderator matrix and comprises beryllium (Be), boron (B), or a compound thereof.

3. The nuclear reactor system of claim 2 wherein the high and slow materials comprise beryllium boride (Be)₂B、Be₄B、BeB₂Or BeB₆) Beryllium carbide (Be)₂C) Zirconium beryllium (ZrBe)₁₃) Titanium beryllium (TiBe)₁₂) Beryllium oxide (BeO) or boron carbide (B)₄C) At least one of (a).

4. The nuclear reactor system of claim 2, wherein the high slowness material is encapsulated within the low slowness material such that the high slowness material is not exposed outside of the low slowness material.

5. The nuclear reactor system of claim 4 wherein:

the fuel elements each include a fuel composite moderator mass formed from the composite moderator medium and nuclear fuel;

the fuel compound moderator block includes a fuel opening; and

the nuclear fuel is disposed inside the fuel opening such that the nuclear fuel is surrounded by the composite moderator medium.

6. The nuclear reactor system of claim 5 wherein:

the fuel composite moderator block also includes coolant channels for flowing a coolant gas or liquid.

7. The nuclear reactor system of claim 5 wherein:

the nuclear fuel comprises a fuel compact comprising:

tri-structure isotropic (TRISO) fuel particles embedded within a silicon carbide matrix; or

Tri-structural isotropic (TRISO) fuel particles embedded within a graphite matrix.

8. The nuclear reactor system of claim 5, wherein the nuclear reactor core further comprises at least one reflector region comprising reflector composite moderator blocks formed from the composite moderator medium.

9. The nuclear reactor system of claim 5 wherein:

the nuclear reactor core including an inner reflector region and an outer reflector region;

the inner reflector region comprises inner reflector composite moderator blocks;

the outer reflector region comprises outer reflector composite moderator blocks; and

the inner reflector composite moderator block and the outer reflector composite moderator block are formed from the composite moderator medium.

10. The nuclear reactor system of claim 9, wherein:

the array of fuel elements surrounding the inner reflector region; and

the outer reflector region surrounds the array of fuel elements.

11. A method, comprising:

selecting two or more moderators comprising a low-moderating material and a high-moderating material to form a composite moderator medium;

selecting a sintering aid and a weight percentage (wt/wt%) of the sintering aid in the composite moderator mixture based on the low moderator material;

mixing the selected sintering aid with the two or more moderators in a selected weight percentage (wt/wt%) to form the composite moderator mixture; and

spark plasma sintering the composite moderator mixture to produce a composite moderator mass formed from the composite moderator medium.

12. The method of claim 11, wherein spark plasma sintering the composite moderator mixture comprises:

pouring the composite moderator mixture into a mandrel; and

pressing a mold into the mandrel to apply a processing temperature and pressure to the composite moderator mixture to produce the composite moderator mass formed of the composite moderator medium.

13. The method of claim 12, wherein:

the low-moderating material includes silicon carbide (SiC) or magnesium oxide (MgO); and

the high-slowdown material contains beryllium (Be), boron (B), or a compound thereof.

14. The method of claim 13, wherein the high and slow materials comprise beryllium boride (Be)₂B、Be₄B、BeB₂Or BeB₆) Beryllium carbide (Be)₂C) Zirconium beryllium (ZrBe)₁₃) Titanium beryllium (TiBe)₁₂) Beryllium oxide (BeO) or boron carbide (B)₄C) At least one of (a).

15. The method of claim 13, wherein:

the low moderator material includes silicon carbide (SiC); and

the sintering aid comprises yttrium oxide (Y)₂O₃) Or aluminum oxide (Al)₂O₃)。

16. The method of claim 15, wherein the selected weight percent (wt/wt%) of the sintering aid in the composite moderator mixture is from 4 wt% (wt/wt%) to 10 wt% (wt/wt%) yttria or alumina.

17. The method of claim 15, wherein the processing temperature is in a range of 1400 degrees celsius (° c) to 1800 degrees celsius (° c).

18. The method of claim 13, wherein:

the low moderating material includes magnesium oxide (MgO); and

the sintering aid comprises lithium.

19. The method of claim 18, wherein the selected weight percent (wt/wt%) of the sintering aid in the composite moderator mixture is from 3 wt% (wt/wt%) to 10 wt% (wt/wt%) lithium.

20. The method of claim 19, wherein the processing temperature is in a range of 1300 degrees celsius (° c) to 1600 degrees celsius (° c).

Technical Field

The subject matter of the present application relates to examples of nuclear systems and nuclear reactor systems that include a composite moderator. The subject matter also encompasses methods for making composite moderators.

Background

Nuclear fission reactors include thermal or fast reactors. Currently, almost all operating reactors are of the thermal type and therefore require moderators to decelerate the fast neutrons so that nuclear fission can continue. The materials used for moderation need to have a very specific set of properties. First, the moderator itself cannot absorb neutrons. Generally, this means that the moderator should have a low neutron absorption cross-section. However, the moderator should be able to decelerate the neutrons to an acceptable speed. Therefore, in an ideal moderator, the neutron scattering cross-section is high. This neutron scattering is a measure of the likelihood that a neutron will interact with the atoms of the moderator. If the collision between a neutron and a nucleus is an elastic collision, this means that the closer the size of the nucleus is to the neutron, the more the neutron slows down. Thus, lighter elements tend to be more effective moderators.

Commonly used moderators such as light water (H)₂O), heavy water (D)₂O) and graphite (C) have a low neutron absorption cross section but a comparatively large neutron scattering cross section. Neutron scattering cross-section (σ) for light water, heavy water and graphite_s) Respectively 49, 10.6 and 4.7 target-ens. Neutron absorption cross section (σ) for light water, heavy water and graphite_s) 0.66, 0.0013 and 0.0035 target respectively. The moderators differ in their moderating ability and their cost.

Currently, operating thermal nuclear reactors use a single monomeric moderator material as the neutron moderator. The unitary moderator material is typically excavated from the ground. To determine the most appropriate monomeric moderator material for a nuclear reactor core, an engineer analyzes whether the neutron properties of the monomeric moderator material are appropriate for the nuclear reactor, for a relatively long period of time, and then performs optimization. Graphite is a neutron moderator that is commonly used in nuclear reactors. Graphite is a crystalline form of the naturally occurring elemental carbon in which the atoms are arranged in a hexagonal structure. Graphite is the most stable form of carbon under standard conditions.

However, a single monomeric moderator material has limitations, one of which is moderator lifetime, which is the physical limitation of the crystals of the monomeric moderator material when the single monomeric moderator material is subjected to nuclear radiation inside the nuclear reactor core. Furthermore, highly moderated materials such as graphite are unstable under nuclear radiation, which causes the highly moderated material to eventually structurally degrade before the nuclear fuel reaches the fuel life limit.

Nuclear graphite was originally developed as a moderator for Chicago Pile nuclear reactors (i.e., the first nuclear reactor in the world) and was the first and perhaps most studied nuclear material. Today, many air-cooled systems (e.g. prismatic or pebble beds) and salt-cooled systems carry a very large load on the core of a graphite nuclear reactor. While processes are continually improved to process graphite into a single moderator material, thereby providing graphite of higher purity and better (more isotropic form), there are still severe moderator lifetime limitations for graphite. In essence, the physical phenomenon of irradiation induced anisotropic crystal expansion results in gross dimensional changes, microcracking, and loss of integrity of the graphite moderator material.

A typical high-temperature gas-cooled reactor (HTGR) with about 200 megawatts of electrical power (MWe) has an associated graphite load of about 600 tons. The newly contemplated salt cooling system will have a similarly large graphite waste stream. Unfortunately, contaminated graphite poses a serious waste problem for these nuclear reactor systems, as evidenced by the approximately 250000 tons of graphite waste disposed of to date. Carbon 14 (although the degree of contamination depends on the nuclear reactor system, the nuclear fuel, and the nuclear fuel quality: (¹⁴C) And³t contamination is unavoidable. This nuclear waste problem is exacerbated by the fact that: graphite moderator life for high power (high neutron impact) systems requires replacement of a large number of nuclear reactor cores in operation.

Disclosure of Invention

Various examples disclosed herein relate to composite moderator technology for a nuclear reactor system, including a nuclear reactor core comprising a composite moderator and a method for manufacturing the composite moderator. Several benefits are realized with composite moderators over a single moderator material such as graphite, water, and molten salt (e.g., FLiBE in combination with lithium fluoride and beryllium fluoride). First, the composite moderator reduces nuclear waste by serving the fuel life of the nuclear fuel without requiring replacement from the nuclear reactor core as compared to a single moderator material. Second, the composite moderator is dimensionally radiation stable (i.e., suffers less structural degradation). Third, the composite moderator improves safety characteristics by eliminating the current graphite oxidation problem.

In a first example, a nuclear reactor system includes a nuclear reactor core. A nuclear reactor core includes an array of fuel elements and a composite moderator medium formed of two or more moderators. The two or more moderators include a low moderating material and a high moderating material. A high moderating material has a higher neutron moderating capability than a low moderating material.

In a second example, a method includes selecting two or more moderators including a low-moderating material and a high-moderating material to form a composite moderator medium. The method also includes selecting a sintering aid and a weight percentage (wt/wt%) of the sintering aid in the composite moderator mixture based on the low moderator material. The method also includes mixing the selected sintering aid with two or more moderators in a selected weight percentage (wt/wt%) to form a composite moderator mixture. The method also includes spark plasma sintering the composite moderator mixture to produce a composite moderator mass formed from the composite moderator medium.

Additional objects, advantages and novel features of the examples will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following and the accompanying drawings or may be learned by production or operation of the examples. The objects and advantages of the subject matter of the present application may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims.

Drawings

The drawings depict one or more embodiments in accordance with the present concepts by way of example only and not by way of limitation. In the drawings, like reference numerals refer to the same or similar elements.

FIG. 1 is a diagram of a nuclear reactor system depicting a nuclear reactor core, control rods, and other components of the assembly.

Fig. 2A is a diagram of fuel particles and fuel compacts of nuclear fuel used in the nuclear reactor core of fig. 1.

FIG. 2B is a diagram of a fuel composite moderator block of the nuclear reactor core of FIG. 1 containing the nuclear fuel of FIG. 2A and formed of a composite moderator medium.

FIG. 2C is a cross-sectional view of a nuclear reactor core and components including an array of fuel elements and reflector composite moderator blocks formed of a composite moderator medium.

FIG. 3 is an enlarged plan view of a portion of the fuel composite moderator block of FIG. 2B depicting nuclear fuel surrounded by a composite moderator medium.

Fig. 4 is a graph showing the dimensional change of a graphite moderator material with time in a nuclear reactor core.

FIG. 5 is a table depicting the properties of graphite moderator material, including neutron moderation capability, compared to two types of low moderator materials of a composite moderator medium and eight types of high moderator materials of a composite moderator medium.

FIG. 6 is a graph illustrating the reactivity over time of a nuclear reactor core containing the graphite moderator material of FIG. 5 compared to a nuclear reactor core containing seven different types of composite moderator media.

FIG. 7 is a flow chart of a method that may be implemented to fabricate a composite moderator block of a composite moderator medium.

FIG. 8A is a photograph of a processing of the method of FIG. 7 in which a composite moderator block is fabricated using Spark Plasma Sintering (SPS).

Fig. 8B is a graph showing process temperature and pressure (mold displacement) over time during spark plasma sintering of the method of fig. 7.

FIG. 8C is a micron level photograph of a composite moderator medium showing the low moderator material encapsulating the high moderator material.

Parts list

100 nuclear reactor system

101 nuclear reactor core

102A to 102N fuel elements

103 composite moderator medium

104 low-moderating material

105 high-moderating material

110 casing structure

115A to 115N control rods

120 steam generator

125 steam line

130 steam turbine

135 electric generator

140 electricity

145 condenser

150 spray

155 steam

160 cooling tower

200 nuclear fuel

201A to 201N nuclear fuel rod

205 fuel compact

206A to 206N fuel pellets

207 silicon carbide substrate

208 graphite matrix

220 composite moderator block

225 fuel composite moderator block

226A-226N fuel openings

227A-227B coolant channels

228 refrigerant

230 reflector region

235A to 235N reflector composite moderator block

240 internal reflector region

245A-245N internal reflector composite moderator blocks

250 outer reflector region

255A to 255N external reflector composite moderator block

260 barrel

265 permanent external reflector

270A to 270N operation control rod

275A to 275N start control rod

280A to 280N backup closed channels

Detailed Description

In the following detailed description, numerous specific details are set forth by way of examples in order to provide a thorough understanding of the relevant teachings. It should be apparent, however, to one skilled in the art that the present teachings may be practiced without such details. In other instances, well known methods, procedures, components, and/or circuits have been described at a relatively high-level, without detail, in order to avoid unnecessarily obscuring aspects of the present teachings.

To treat nuclear radiation, composite moderator media (e.g., solid materials) are designed that contain several moderator materials that change in direction generally for a longer duration than a single moderator material. For example, the composite moderator medium enables a nuclear reactor core to have an extended life without replacing the moderator material and to be more compact than a graphite moderator material. Further, the composite moderator medium can be deployed in various nuclear reactor system embodiments, such as a land reactor for power generation or a high temperature Nuclear Thermal Propulsion (NTP) system (e.g., a compact space nuclear reactor).

Reference will now be made in detail to the examples illustrated in the accompanying drawings and discussed below.

Fig. 1 is a diagram of a nuclear reactor system 100 depicting a nuclear reactor core 101, control rods 115A-115N, and other components of the assembly. In this example, the nuclear reactor system 100 includes a nuclear reactor 101 in which a controlled nuclear chain reaction occurs and energy is released. In this example, the nuclear reactor system 100 is a nuclear power plant in a land-based application. However, nuclear reactors and composite moderator technology may be used in a space environment, such as in a Nuclear Thermal Propulsion (NTP) system. In such NTP systems, the thrust generated propels a vehicle, such as a rocket, a drone, an Unmanned Air Vehicle (UAV), an aircraft, a spacecraft, a missile, or the like, that houses the nuclear reactor core 101, is integral with the nuclear reactor core 101, is connected to the nuclear reactor core 101, or is attached to the nuclear reactor core 101. In addition, NTP systems may be used to propel submarines or vessels.

The nuclear reactor core 101 includes an array of fuel elements 102A-102N and a composite moderator medium 103. The nuclear reactor core 101 is a nuclear fission reactor core that includes nuclear fuel to produce a thermal power (MWt) of megawatts or greater. A plurality of circumferential control rods 115A-115N may surround the array of fuel elements 102A-102N to change the reactivity of the nuclear reactor core 101 by rotating the control rods 115A-115N. The containment structure 110 houses the nuclear reactor core 101, control rods 115A-115N, and a steam generator 120. The control rods 115A-115N may be positioned in the region of the reflector regions 240, 250 (see fig. 2C) of the nuclear reactor core 101 to adjust the number of neutrons and the reactor power level during operation by changing the reactivity of the nuclear reactor core 101.

The control rods 115A to 115N are composed of chemical elements (e.g., boron, silver, indium, and cadmium) capable of absorbing many neutrons without fissioning themselves. The nuclear reactor core 101 generates thermal energy that is released as heat. Other components of the nuclear reactor system 100 convert thermal energy into a usable form of energy, such as electricity 140. In this example, the nuclear reactor core 101 provides thermal energy to the steam generator 120, the steam generator 120 extracts the thermal energy into the steam line 125, and the steam line 125 rotates the steam turbine 130. The steam turbine 130 drives the generator 135, and the generator 135 then converts the thermal energy to electricity 140. Subsequently, the thermal expansion cycle is repeated.

In the exemplary nuclear reactor system 100, the condenser 145 generates a coolant, such as a high pressure liquid or gas, for feeding the nuclear reactor core 101 and cooling the components of the nuclear reactor system 100. For example, during an expansion cycle, propellant stored in the cooling tower 160 may be drawn through the nuclear reactor core 101 to cool the nuclear reactor core 101. Heat from the coolant may be extracted into the cooling tower 160 in the form of the spray 150 and released from the cooling tower 160 as water vapor 155. It is noted that some coolant may be returned, for example, exhausted from the nuclear reactor core 101 via a bypass to rotate the steam turbine 130. In some examples, the nuclear reactor system 100 may be used in a molten salt circulation application.

The neutron chain reaction in the nuclear reactor core 101 is critical-a single neutron from each fissile nucleus causes fission in another nucleus-the chain reaction must be controlled. The composite moderator medium 103 is formed of two or more moderators that effectively adjust the criticality and provide an extended moderator lifetime that can match the lifetime of the nuclear fuel. The two or more moderators include a low moderating material 104 and a high moderating material 105. The high moderating material 105 has a higher neutron moderating capacity than the low moderating material 104, which can be associated with neutron absorption and scattering cross sections. The composite moderator medium 103 in the nuclear reactor core 101 moderates fast neutrons (generated by splitting atoms in fissile compounds such as uranium-235) to make them more efficient in nuclear fission chain reactions. This slowing or moderation of the neutrons allows the neutrons to be more easily absorbed by the fissile nuclei, thereby producing more fission events. Two or more moderators may be tailored to a very specific set of characteristics depending on the implementation environment of the nuclear reactor core 101 (e.g., power generation or NTP).

As will be further explained in fig. 5 to 6, the low moderator material 104 includes a moderator matrix of silicon carbide (SiC)104A or magnesium oxide (MgO) 104B. The high moderator material 105 is dispersed within the moderator matrix and comprises beryllium (Be)105H, boron (B), or a compound thereof. More specifically, the high moderator material 105 includes beryllium boride (Be)₂B 105A、Be₄B 105B、BeB₂Or BeB₆) Beryllium carbide (Be)₂C105C), zirconium beryllium (ZrBe)₁₃105D) Titanium beryllium (TiBe)₁₂105E) Beryllium oxide (BeO105F) or boron carbide (B¹¹B₄C105G). The high moderator material 105 is encapsulated within the low moderator material 104 such that the high moderator material 105 is not exposed outside of the low moderator material 104.

Fig. 2A is a diagram of fuel particles 206A-206N and a fuel compact 205 (e.g., a fuel pellet) of a nuclear fuel 200 used in the nuclear reactor core 101 of fig. 1. In one example, the nuclear fuel 200 includes a fuel compact 205 composed of tri-structure isotropic (trisopic) fuel particles 206A to 206N embedded inside a silicon carbide matrix 207. In another example, the nuclear fuel 200 includes tri-structural isotropic (TRISO) fuel particles 206A to 206N embedded inside a graphite matrix 208 to form fuel pellets. The TRISO fuel particles 206A to 206N centrally comprise a fuel core consisting of UC or uranium oxycarbide (UCO) coated with one or more layers surrounding one or more isotropic materials. As shown in fig. 2A, the TRISO fuel particles 206A to 206N include four layers of three isotropic materials. In this example, the four layers are: (1) a porous buffer layer made of carbon, followed by (2) a dense inner layer of pyrolytic carbon (PyC), followed by (3) a ceramic layer of SiC for retaining fission products at elevated temperatures and imparting strong structural integrity to the TRISO particles, followed by (4) a dense outer layer of PyC.

The TRISO fuel particles 206A-206N are designed not to crack due to stress or fission gas pressure at temperatures in excess of 1600 deg.C, and therefore may contain fuel in the worst case of accident. The TRISO fuel particles 206A-206N are designed for use in an exemplary cross-section of a nuclear reactor core 101 of a high temperature gas cooled reactor (HTGR) such as that shown in FIG. 2C, operating at a temperature much higher than the temperature of LWR. The fuel compact 205 may be loaded into a fuel pin or rod, clad, and stacked inside a plurality of columns of fuel elements 102A-102N. Among the possible matrix 207, 208 materials for the TRISO fuel particles 206A to 206N, silicon carbide (SiC) provides good irradiation behavior and production. SiC rapidly forms dense, coherent silicon dioxide (SiO) due to exposure to air at elevated temperatures₂) The surface scale is excellent in oxidation resistance, which prevents further oxidation.

The use of coated fuel particles 206A-206N makes it more difficult to achieve high heavy metal densities in the nuclear fuel 200 because the net heavy metal density within the fuel particles 206 decreases rapidly as the coating thickness increases. This fact requires that the ratio of coating thickness to core diameter be kept as small as possible while maintaining utility as a fission product barrier. However, it is clear that the use of dispersion fuels in LWR would require higher enrichment and lower power density. The most likely fission particle types for composite nuclear fuels are uranium/plutonium carbide (UC or PuC) and uranium/plutonium nitride (UN or PuN) due to the combination of high melting temperature and high actinide density. Uranium silicide may provide even higher densities of fissile uranium, but may be unstable under expected manufacturing and operating conditions. Other types of fuel particles 206A-206N may be used, including quadurio fuels, which contain one or more burnable neutron poisons (e.g., erbium oxides) surrounding the fuel core or TRISO particles to better manage excess reactivity; and CerMet fuels (e.g., ceramic fuel particles 206A to 206N embedded in a metal matrix, such as uranium oxide); and so on.

In some examples of nuclear fuel 200, uranium dioxide (UO)₂) The powder is compacted to form a cylindrical fuel compact 205 and sintered at high temperature to produce ceramic nuclear fuel pellets having high density and well defined physical properties and chemical composition. A grinding process is used to achieve a uniform cylindrical geometry with narrow tolerances.

FIG. 2B is a diagram of a fuel composite moderator block 225 of the nuclear reactor core 101 of FIG. 1, the fuel composite moderator block 225 comprising the nuclear fuel 200 of FIG. 2A and being formed from the composite moderator medium 103. Fuel elements 102A through 102N (shown in FIG. 2C) each include a composite moderator block 220 formed from composite moderator medium 103 and nuclear fuel 200. Fuel compound moderator block 225 includes fuel openings 226A through 226N. The nuclear fuel 200 is disposed inside the fuel openings 226A to 226N such that the nuclear fuel 200 is surrounded by the composite moderator medium 103. Fuel composite moderator block 225 also includes coolant channels 227A through 227B to flow coolant 228 (e.g., gas or liquid).

A number of such fuel compacts 205 (as shown in fig. 2A) are stacked and filled into the depicted nuclear fuel rods 201A-201N (e.g., sealed tubes). The cladding is the outer layer of the nuclear fuel rods 201A to 201N, preventing radioactive fission fragments from escaping from the nuclear fuel 200 into the coolant 228 and contaminating the coolant 228. The metal used for the cladding of the nuclear fuel rods 201A to 201N depends on the design of the nuclear reactor core 101, but may include stainless steel, magnesium containing aluminum, or zirconium alloy having low neutron absorption in addition to being highly corrosion resistant. The completed nuclear fuel rods 201A-201N are combined into a fuel assembly for constructing the nuclear reactor core 101, as described in fig. 2C.

FIG. 2C is a cross-sectional view of a nuclear reactor core 101 and components, including an array of fuel elements 102A-102N and a plurality of reflector composite moderator blocks 245A-245N, 255A-255N formed from a composite moderator medium 103. Generally, the nuclear reactor core 101 includes at least one reflector region 230 (shown as an inner reflector region 240 and an outer reflector region 250), the reflector region 230 including reflector composite moderator blocks 235A to 235N formed from the composite moderator medium 103. In the example arrangement of fig. 2C, the nuclear reactor core 101 includes an inner reflector region 240 and an outer reflector region 250. Internal reflector region 240 includes internal reflector composite moderator blocks 245A through 245N. The outer reflector region 250 includes outer reflector composite moderator blocks 255A through 255N. The inner-reflector composite moderator blocks 245A through 245N and the outer-reflector composite moderator blocks 255A through 255N are formed from a composite moderator medium 103. The array of fuel elements 102A to 102N forming the hexagonal fuel block region surrounds the inner reflector region 240. The outer reflector region 250 surrounds the array of fuel elements 102A-102N. Thus, the fuel elements 102A-102N are between the inner reflector region 240 and the outer reflector region 250.

In fig. 2C, the nuclear reactor core 101 is a high-temperature gas nuclear reactor core 101 in a prism shape. Of course, the composite moderator technology may be used within any nuclear reactor core 101 that is not gas-based. In this example gas nuclear reactor core 101, the composite moderator block 220 is a block of composite moderator medium 103, which is a solid material formed of both low moderator material 104 and high moderator material 105. The composite moderator block 220 may be prism-shaped (e.g., hexagonal-shaped) and include a plurality of openings (holes) drilled therein, such as coolant channels 227A to 227B for flowing coolant 228 and fuel openings 226A to 226N for placing nuclear fuel 200 inside. A number of nuclear fuel rods 201A-201N of nuclear fuel 200 are dropped into each composite moderator block 220 to form each fuel composite moderator block 225 (e.g., a fuel bundle). The fuel composite moderator blocks 225A-225N (e.g., fuel bundles) are then loaded into the nuclear reactor core 101.

Many of the composite moderator blocks 220 are fueled, which are shown as fuel composite moderator blocks 225A through 225N. Most of the compound moderator blocks 220 are not fueled (i.e., no nuclear fuel 200 is present) and therefore do not include the fuel openings 226A through 226N. These reflector composite moderator blocks 235A-235N (shown as inner reflector composite moderator blocks 245A-245N and outer reflector composite moderator blocks 255A-255N) include coolant channels 227A-227B for flowing coolant 228.

The array of fuel elements 102A-102N includes hundreds of fuel composite moderator blocks 225A-225N in the shape of hexagons, with 102 fuel columns in this example, where each fuel column is ten (10) fuel composite moderator blocks 225A-225J high. The cross-section of the nuclear reactor core 101 is about six meters wide and each composite moderator block 220 is about 30cm wide. In the middle of the cross section of the nuclear reactor core 101 are internal reflector composite moderator blocks 245A to 245N, which comprise a number of columns of internal reflector composite moderator blocks 245A, stacked 10 per column. Outside of the cross section of the nuclear reactor core 101 are external reflector composite moderator blocks 255A through 255N that include a number of columns of external reflector composite moderator blocks 255A through 255N stacked 10 on each column. Typically the columns of control rods 115A-115N, fuel compound moderator blocks 225A-225N, the columns of inner reflector compound moderator blocks 245A-245N, and the columns of outer reflector compound moderator blocks 255A-255N are the same length, however, it is understood that the lengths may vary depending on the implementation.

The control rods 115A-115N as shown in FIG. 1 include both operating control rods 270A-270N and actuating control rods 275A-275N that are inserted through the top of the nuclear reactor core 101 and through a subset of the composite moderator blocks 220. Thirty-six (36) operating control rods 270A through 270N pass through a subset of the outer reflector composite moderator blocks 255A through 255N. Twelve (12) activation control rods 275A-275N pass through a subset of the fuel composite moderator blocks 225A-225N. The control rods 270A to 270N, 275A to 275N absorb neutrons. The cylinder 260 is a metal core support formed of, for example, steel, that surrounds a bundled set of the array of fuel elements 102A to 102N, the inner reflector region 240, and the outer reflector region 250 of the nuclear reactor core 101 on the outer periphery of the nuclear reactor core 101. A permanent external reflector 265 (which may be formed from the composite moderator medium 103) is disposed between the external reflector region 250 and the barrel 260. The permanent outer reflector 265 includes a partially hexagonal shaped filler element surrounding the perimeter of the outer reflector region 250 that makes up the nuclear reactor core 101. Eighteen (18) backup shut-off passages 280A-280N are positioned in regions within the array of fuel elements 102A-102N.

In general, the composite moderator blocks 220A through 220N used in the fuel composite moderator blocks 225A through 225N, the inner reflector composite moderator blocks 245A through 245N, and the outer reflector composite moderator blocks 255A through 255N are each formed of the same composite moderator medium 103, have the same outer shape (e.g., helical), and have coolant channels 227A through 227B for flowing the coolant 228. However, the composite moderator blocks 220 for the inner reflector composite moderator blocks 245A to 245N and the outer reflector composite moderator blocks 255A to 255N do not include the drilled-in fuel openings 226A to 226N. Thus, from one hundred feet out, the nuclear reactor core 101 appears to have many large hexagonal composite moderator blocks 225A-225N that appear to be nearly identical, but the fuel block area (e.g., the central hexagonal-shaped portion of the nuclear reactor core 101) that includes the array of fuel elements 102A-102N contains exactly the nuclear fuel 200 arranged inside the composite moderator blocks 220A-220N.

FIG. 3 is an enlarged plan view of a portion of the fuel composite moderator block 225 of FIG. 2B, depicting the nuclear fuel 200 surrounded by the composite moderator medium 103. Between the fissile material blocks of the two nuclear fuel rods 201A to 201B, the single control rod 115A strongly absorbs 303 neutrons. The deeper the control rod 115A is inserted between the fissile material of the nuclear fuel rods 201A to 201B, the more difficult it is for neutrons generated by the fission 301 to collide, resulting in a more limited chain reaction, andthe heat energy generation is reduced. As shown, composite moderator medium 103 fills the voids between fuel rods 201A-201B, thereby reducing neutron energy by slowing down the neutrons (moderating 302). Without the composite moderator medium 103, the neutrons would move too fast and therefore have very low incidences²³⁵The possibility of fission 301 of U, so these neutrons pass through many nuclei before being absorbed and causing fission.

Fig. 4 is a graph 400 illustrating the dimensional change of a graphite moderator material 505 over time in the nuclear reactor core 101. In the graph 400, displacement per atom (dpa) 401 is shown on the x-axis with respect to life in the nuclear reactor core and delta Δ V/V (%) 402 is the dimensional change of the graphite moderator material 505. Large dimensional changes occur in the nuclear graphite moderator material 505 due to neutron irradiation, meaning lifetimes in the range of 10dpa (displacement per atom) to 20dpa (displacement per atom). Typically, high power, high temperature gas cooled reactors (HTGR) require that nuclear reactor core graphite be replaced once or twice during the life of the plant.

Four different types of graphite moderator materials 505 are plotted in the graph 400, at both 750 degrees Celsius (C.) and 900 degrees Celsius (C.). The temperature of the graphite moderator material 505 within the core of the nuclear reactor is a function of location, which has large variability. The graphite moderator material 505 may be 750 degrees celsius (c) at or near the bottom of the nuclear reactor core and 900 degrees celsius (c) at or near the top of the nuclear reactor core. Thus, the graphite moderator material 505 has different lifetimes depending on the particular location or arrangement in the nuclear reactor core.

As shown in the graph 400, the graphite moderator material undergoes large dimensional changes and then decomposes while in the nuclear reactor core, and this is referred to as "moderator life". While this expansion mechanism is compensated to some extent by the inherent porosity of the nuclear graphite (substantially all of the nuclear graphite material is about 18% porous), the moderator lifetime is related to the point at which the graphite returns to its original zero expansion value or the zero point of the two traces in the plot of fig. 4 that are interpolated for the range of four types of nuclear graphite moderator material 505 that are plotted. Thus, a problem with the graphite moderator material 505 is the very limited and defined moderator lifetime due to radiation damage during the runtime of the nuclear reactor core.

FIG. 5 is a table 500 depicting properties (including neutron moderation capacity) of graphite moderator material 505 compared to two types of low moderator materials 104A through 104B of composite moderator medium 103 and eight types of high moderator materials 105A through 105H of composite moderator medium 103. As can be seen, the goal is to place (e.g., encapsulate) high moderator materials 105A through 105H with poor radiation performance inside low moderator materials 104A through 104B with good radiation performance, thus significantly extending the moderator lifetime of the composite moderator medium 103 compared to the graphite moderator material 505. As can be seen in Table 500, the deceleration capacity 510 of the low moderator materials 104A through 104B is relatively lower than the high moderator materials 105A through 105H. Another comparative characteristic included in table 500 is melting temperature (T)_Melting)515, density 520, chemical reactivity 525, crystal 530, irradiation properties 535, and thermal conductivity 540.

The composite moderator medium 103 has significantly less dimensional changes without requiring replacement. The composite moderator medium 103 is a material that can extend the fuel life of the nuclear reactor core 101. This is achieved through two-phase structure-matched neutron moderation while increasing irradiation stability through the use of an excellent moderating matrix material, such as SiC 104A or MgO 104B, for the low moderating material 104.

The selection of the second high moderator material 105 in the encapsulated phase or entrained phase in the moderator matrix phase of the low moderator material 104 is driven by the need for enhanced moderation, as understood by examining table 500 of fig. 5. A simple measure of slowness is the average logarithmic decay

And the possibility of this interaction (macroscopic section ∑)_D) The product of (c) is referred to as the deceleration capacity 510. Thus, the composite moderator medium 103 includes moderator material trapped in the low moderator materials 104A through 104B (e.g., moderator matrix for entraining the high moderator materials 105A through 105H)) The inner highly moderating materials 105A to 105H. The low moderator materials 104A through 104B are, in fact, relatively stable under radiation. The first exemplary low slowness material in the table is silicon carbide (SiC)104A, which may be, for example, Chemical Vapor Deposition (CVD) SiC.

The high-moderating materials 105A to 105H are put inside the moderating matrix of the low-moderating materials 104A to 104B. The moderator matrix of the low moderator materials 104A to 104B is weakly moderated (silicon carbide or magnesium oxide), but the high moderator materials 105A to 105H have a large moderating ability. All of the high moderator materials 105A to 105H are better than graphite in terms of deceleration capacity 510, and all of the low moderator materials 104A to 104B are not as good as graphite moderator material 505 in terms of deceleration capacity 510. The goal is to average the slowing capabilities (slowing capabilities 510) of the low slowing materials 104A-104B and the high slowing materials 105A-105H together to produce a composite moderator medium 103 that is more radiatively stable and lasts longer within the nuclear reactor core 101 than the graphite moderator material 505. In some examples of the composite moderator medium 103, the moderator matrix of the low moderator materials 104A to 104B is matched with the high moderator materials 105A to 105H, which are beryllium (Be) or boron (B) compounds.

To successfully replace the nuclear graphite moderator material 505, the composite moderator medium 103 has similar moderating capabilities and achieves a longer moderator life than the nuclear graphite moderator material 505. The composite moderator medium 103 can be considered a two or more phase structure (e.g., components), a fiber structure, or an alloy. Unlike the graphite moderator material 505 shown in FIG. 4, silicon carbide (SiC), as the low moderator material 104A, was demonstrated to survive to greater than 100 per atom displacement (dpa) under a nominal "saturated" volume change. An example of such a structure presented herein is a 45% volume fraction of the high moderator material 105A to 105H into the SiC body moderator matrix by Spark Plasma Sintering (SPS) processing as described in fig. 7 and fig. 8A to 8C. The second body moderating matrix is similarly processed magnesium oxide (MgO). Fabrication of the designed composite moderator medium 103 by this rapid advanced fabrication SPS technique shows economical fabrication. Similar manufacturing techniques may be hot pressing and sintering.

FIG. 6 is a graph 600 illustrating the reactivity over time of a nuclear reactor core containing the graphite moderator material 505 of FIG. 5 compared to a nuclear reactor core containing seven different types of composite moderator media 103A through 103G. The graph 600 shows K as shown by the reactivity factor, i.e., on the Y-axis_eff602, some of the compound moderator media 103A through 103G are better, some are worse, on the first day as a function of time, i.e., year 601 shown on the X-axis. The composite moderator media 103A to 103G may perform better or worse than the graphite 505 depending on the selection of the low moderator materials 104A to 104B and the high moderator materials 105A to 105H. There are two forms of lifetime: (1) fuel life-nuclear fuel 200 decays and burns out and the nuclear reactor is shut down; and (2) moderator life-moderator decomposition and nuclear regulatory mechanisms determine that the nuclear reactor is too hazardous and must be shut down.

When K is_eff602 falls below 1, the initial load of the nuclear fuel 200 reaches the fuel life. The graph 600 does not show the moderator lifetime of the composite moderator media 103A to 103G, but rather the graph 600 shows the fuel lifetime and the composite moderator media 103A to 103G are as effective as the graphite moderator material 505. The graph 600 also shows that the slowing down (i.e., increasing or decreasing) may be adjusted based on the selected composite slowing down media 103A through 103G.

Although not shown in FIG. 6, the moderator life of all of the composite moderator media 103A to 103G is significantly extended relative to the graphite moderator material 505, for example, to match fuel life without replacement. A problem with the graphite moderator material 505 is the significant expense of opening up the nuclear reactor core 101 to replace the graphite moderator material, which can be about 1 billion dollars. The composite moderator media 103A to 103G may be a large upfront investment, but will save the cost required to replace the graphite moderator material 505 in the nuclear reactor core 101 after a period of time. Although more expensive on the first day, the composite moderator media 103A through 103G will be less expensive in the future because there is no need to replace the moderator elements (needed for the graphite moderator material 505). Typically, when the graphite moderator material 505 becomes damaged, all of the nuclear reactor core with the graphite moderator material 505 is replaced except for the low volume permanent external reflector 265 (see fig. 2C). Such replacement is inevitable unless the graphite moderator material 505 is replaced with a higher performance composite moderator medium 103.

The graph 600 shows representative calculations for a representative reactor type: 35% by volume of SiC matrix fuel in small modular prismatic high temperature gas cooled reactor (HTGR) ((r))Nuclear fuel). In the diagram 600, a graphite moderator material 505 is added with 9% concentrated UO₂TRISOGraphite moderation of nuclear fuel is implemented in nuclear reactors. As shown, some BeSiC type composite moderator media 103A, 103D implemented in prismatic HTGR nuclear reactor cores may have an unacceptably large effect on nuclear reactor core life due to the greater absorption of silicon carbide (SiC). Meanwhile, in prismatic HTGR, decreasing the moderator fraction of SiC or increasing the porosity of the composite moderator media 103B to 103C or increasing the concentration of the TRISO fuel particles provides a comparable lifetime to the graphite moderator material 505. In addition, it is seen that MgO-based composite moderator media 103E to 103G have very good neutron performance.

The graph 600 illustrates the neutron effect of using a composite moderator in a typical small modular high temperature gas cooled reactor (HTGR). In particular, K_eff602 is a measure of reactor core reactivity, predicted as a function of year 601 and compared to standard nuclear graphite moderator material 505. Several examples are set on the variables of the type of moderator matrix of the low moderator material 104 (100% dense SiC, 100% dense MgO, and 20% porous SiC) and the variable uranium enrichment of the nuclear fuel 200. In the example of fig. 6, the nuclear fuel 200 is composed of a fuel compact 205 of tri-structure isotropic (TRISO) fuel particles 206A to 206N embedded inside a silicon carbide matrix 207. An additional variable is the volume fraction of the moderated matrix and beryllium phase. Shown is excessive parasitic moderator, with too much moderation, and proliferative combustionA recipe of materials (too reactive), and a set of selected curves comparable to the nuclear graphite moderator material 505.

FIG. 7 is a flow diagram of a method that may be implemented to fabricate a composite moderator block 220 of the composite moderator medium 103. Beginning at step 700, the method includes selecting two or more moderators including a low moderator material 104 and a high moderator material 105 to form a composite moderator medium 103. This includes selecting powders for the low moderator material 104 and the high moderator material 105. Generally described, the low moderator material 104 comprises silicon carbide (SiC)104A or magnesium oxide (MgO) 104B. The high moderator material 105 contains beryllium (Be 105H), boron (B), or a compound thereof. More specifically, the high moderator material 105 includes beryllium boride (Be)₂B105A、Be₄B 105B、BeB₂Or BeB₆) Beryllium carbide (Be)₂C105C), zirconium beryllium (ZrBe)₁₃105D) Titanium beryllium (TiBe)₁₂105E) Beryllium oxide (BeO105F) or boron carbide (B¹¹B₄C105G).

Continuing to step 710, the method further includes selecting a weight percentage (wt/wt or wt%) or a weight fraction of the sintering aid and the sintering aid in the composite moderator mixture based on the low moderator material 104. This includes selecting one or more suitable sintering aids and weight percentages or fractions based on the combination of powders for the low and high moderator materials 104 and 105. The sintering aid is a eutectic powder, such as an oxide (e.g., yttria and alumina) for silicon carbide and lithium for magnesia. The mixing ratio of the composite moderator mixture may be expressed using mass percentage or mass fraction instead of weight percentage or weight fraction.

The sintering aid varies depending on, for example, the low moderator material 104. Sintering aids include various oxides, such as yttrium oxide (Y), known as yttrium oxide₂O₃) Or alumina (Al) known as aluminum oxide₂O₃) (ii) a And lithium. In a first example, where the low moderator material 104 comprises silicon carbide (SiC)104A, then the sintering aid comprises yttria (Y)₂O₃) Or aluminum oxide (Al)₂O₃). In this first example, the selected weight percent (wt/wt%) of the sintering aid in the composite moderator mixture is 3 to 10 weight percent (wt/wt%), more preferably 4 to 10 wt/wt% yttria or alumina. In a second example, where the low moderator material 104 comprises magnesium oxide (MgO)104B, then the sintering aid comprises lithium. In this second example, the selected weight percent (wt/wt%) of the sintering aid in the composite moderator mixture is 3 to 10 weight percent (wt/wt%) lithium.

Continuing to step 720, the method further includes mixing the selected sintering aid with two or more moderators in a selected weight percentage (wt/wt%) to form a composite moderator mixture. Ending now at step 730, the method further includes spark plasma Sintering (SP) the composite moderator mixture to produce a composite moderator block 220 formed from the composite moderator medium 103. SPS uses additives (e.g., sintering aids) to suppress the sintering temperature, which reduces the processing temperature and pressure required to perform a processing run. The sintering aid reduces the temperature and time for processing, which advantageously minimizes evaporative loss of the highly moderating material 105 (e.g., beryllium and boron compounds) of the composite moderator medium 103.

The step of spark plasma sintering the composite moderator mixture comprises: pouring the composite moderator mixture into the core rod; and pressing a mold into the core rod to apply a processing temperature and pressure to the composite moderator mixture to produce a composite moderator block 220 formed of composite moderator medium 103. The mold is like a piston that applies processing temperature and pressure to the composite moderator mixture. The processing temperature varies depending on, for example, the low moderator material 104.

Returning to the first example, where the low moderator material 104 comprises silicon carbide (SiC)104A and/or the sintering aid comprises yttria (Y)₂O₃) Or aluminum oxide (Al)₂O₃) And then the processing temperature is in the range of 1400 degrees Celsius (. degree. C.) to 1800 degrees Celsius (. degree. C.). SPS addition to a silicon carbide low-moderated matrix 104AAt the end of the process, the yttria or alumina is partially evaporated. Thus, yttria or alumina may be partially present in the composite moderator mass 220 and may be detected in trace amounts after SPS processing of the composite moderator medium 103.

Returning to the second example, where the low-slowness material 104 includes magnesium oxide (MgO)104B and/or the sintering aid includes lithium, then the processing temperature is in the range of 1300 degrees Celsius (C.) to 1600 degrees Celsius (C.). The lithium sintering aid completely evaporates at the end of SPS processing of the magnesium oxide slow slowing material 104B. Thus, lithium is not present in the composite moderator block 220 and is generally not detectable (i.e., lithium leaves as a fugitive additive).

Spark Plasma Sintering (SPS) is a sintering technique, also known as Field Assisted Sintering (FAST) or Pulsed Electric Current Sintering (PECS). The SPS is primarily characterized by a pulsed or pulse-free DC or AC current directly through the graphite mold and the powder compact in the case of conductive samples. It was found that joule heating plays a major role in the densification of powder compacts, which allows achieving close to theoretical densities at lower sintering temperatures compared to conventional sintering techniques. Heat generation is internal as opposed to conventional hot pressing where heat is provided by an external heating element. This facilitates very high heating or cooling rates (up to 1000 kelvin/min) and therefore the sintering process is very fast (within a few minutes). The general speed of SPS processing ensures that it has the potential to densify powders with nanometer dimensions or nanostructures while avoiding coarsening that accompanies standard densification routes. SPS is a good method for making ceramics based on nanoparticles with enhanced magnetic, electromagnetic, piezoelectric, pyroelectric, optical or biomedical properties.

FIG. 8A is a tooling photograph 800 of the method of FIG. 7 in which Spark Plasma Sintering (SPS) is used to fabricate the composite moderator block 220. The SPS processing photograph 800 shows the SPS step viewed through a Direct Current Sintering (DCS) window when the material is Spark Plasma Sintered (SPS) to produce the composite moderator block 220. In this case, the graphite punch (luminescent) inside the graphite die containing the powder is shown through the DCS window of fig. 8A. SPS is an advanced manufacturing technique that enables manufacturing at much lower processing temperatures to quickly produce the composite moderator block 220. The low moderator material 104 powder and the high moderator material 105 powder are mixed and a sintering aid is added. SPS allows the powder to be heated rapidly. To encapsulate the high moderator material 105, the low moderator material 104 is cured during SPS before the high moderator material 105 (e.g., beryllium compound) evaporates to form a composite moderator block 220 of the composite moderator medium 103.

Composite moderator media 103 processing can be performed using high vacuum direct current sintering (Sinterland LABOX 3010KF) of relatively pure SiC powder. Powders currently compacted to high density include nano SiC powders in the range of 35 nanometers (nm) to 100 nanometers (nm) and SiC powders from Acheson in the range of 0.2 micrometers (μm) to 2 micrometers (μm). All materials were kinetically stable, ensuring a sufficiently impurity-free dispersion, cold-pressed sintered in a spark plasma sintering apparatus.

Fig. 8B is a graph 810 illustrating process temperature 815 and pressure (mold displacement) 820 over time 825 during Spark Plasma Sintering (SPS) of the method of fig. 7. Graph 810 shows temperature 815, pressure 820 (e.g., mold displacement), and time 825 for a processing run with SPS to produce composite moderator block 220. Graph 810 provides a relative time-temperature trace of the SPS step, showing the process temperature in the 1500 ℃ range for SiC as the low moderator material 104A. For MgO as the low moderating material 104B, the processing temperature may be in the range of 1300 ℃ to 1600 ℃. Operation of the rolling process of the low moderator material 104, such as silicon carbide (SiC), can take hours for the furnace to reach the required temperature well above 2000 ℃ to produce the composite moderator block 220. By SPS processing, ten minutes at processing temperatures of 1600 ℃ to 1800 ℃ enables the manufacture of composite moderator block 220 of composite moderator medium 103.

To minimize any tendency of the sintering additive to absorb moisture, zirconium or zirconium oxide may be added. The processing temperature for SiC as the low moderator material 104A may be in the range of 1400 ℃ to 1800 ℃ with the addition of the sintering additive alumina or yttria for enhanced densification.

FIG. 8C is a micron level photograph 830 of a polished cross-section of composite moderator medium 103 showing low moderator material 104A (SiC) encapsulating high moderator material 105C (Be)₂C) In that respect The micron level photograph 830 is a micrograph of the composite moderator medium 103 of the produced composite moderator block 220 and shows the crystal microstructure of the low moderator material 104A near the interface of the high moderator material 105C. The silicon carbide moderator matrix of the low moderator material 104A completely encapsulates (covers) the high moderator material 105C (Be)₂C) Porous carbon coating of beryllium material. Since beryllium is toxic, it is advantageous to encapsulate the beryllium with a non-toxic silicon carbide low-moderator material 104A because exposure to the toxic high moderator material 105C is eliminated. A porous carbon intermediate layer is coated on the beryllium of the high moderator material 105C and thus between the silicon carbide low moderator material 104A. The silicon carbide moderator matrix (SiC) of the low moderator material 104A is fully densified around the high moderator material 105C.

Micron level photograph 830 of fig. 8C shows an image of a two-phase SiC matrix composite (moderator second phase volume fraction of about 35%). As depicted in fig. 8B, sintering above 1500 ℃ was applied with a hold time of about 10 minutes, achieving SiC moderated matrix density near full density for relatively small (8mm diameter) parts.

Various composite moderators have been disclosed for use in nuclear reactors, including advanced nuclear fission reactor applications. The composite moderator is, for example, a high moderator material 105 (e.g., a beryllium-containing phase) contained in a low moderator material 104 (e.g., a continuous body of SiC or MgO or a radiation-stable matrix phase). Neutron moderation similar to that of the nuclear graphite moderator material 505 can be provided by using the high moderating material 105, while providing many of the safety, economic, and waste reduction benefits exhibited by using the low moderating material 104. Thus, the composite moderator may replace the nuclear graphite moderator material 505 and has excellent moderator lifetime as well as increased safety and waste disposal characteristics. An exemplary manufacturing process includes using eutectic powders during Spark Plasma Sintering (SPS) of a low moderator material 104 (e.g., a radiation stabilized matrix of silicon carbide and magnesium oxide) and a high moderator material 105.

As described above, a method for manufacturing a composite moderator (e.g., a composite moderator block 220) formed from a composite moderator medium 103 for a nuclear reactor core 101 is disclosed. The method includes producing a composite moderator medium 103 (two-phase composite moderator) comprising a high moderator material 105 (e.g., a second capture phase) in a continuous body of low moderator material 104 (e.g., a first matrix phase). The low moderator material 104 is SiC or MgO. The highly moderating material 105 is a beryllium containing compound (e.g., Be)₂C. BeO) or with Be₂Beryllium metal of the shell of C or BeO. The interphase between the low moderator material 104 and the high moderator material 105 is a porous compliant structure capable of absorbing helium produced by the n-alpha reaction. The composite moderator medium 103 is a life component of the nuclear reactor core 101.

It will be understood that the terms and expressions used herein have the ordinary meaning as is accorded to such terms and expressions with respect to their corresponding respective areas of inquiry and study except where specific meanings have otherwise been set forth herein. Relational terms such as first and second, and the like may be used solely to distinguish one entity or action from another entity or action without necessarily requiring or implying any actual such relationship or order between such entities or actions. The terms "comprises," "comprising," or any other variation thereof, are intended to cover a non-exclusive inclusion, such that a process, method, article, or apparatus that comprises or comprises a list of elements or steps does not include only those elements or steps but may include other elements or steps not expressly listed or inherent to such process, method, article, or apparatus. Without further limitation, the absence of a quantitative term modifying an element at the outset does not preclude the presence of additional identical elements in the process, method, article, or apparatus that comprises the element.

Unless otherwise indicated, any and all measurements, values, ratings, positions, sizes, dimensions, and other measurements set forth in this specification (including the appended claims) are approximate and not exact. Such amounts are intended to have reasonable ranges consistent with the function to which they pertain and the customary usage in the art to which they pertain. For example, unless otherwise specifically noted, parameter values and the like may vary by as much as ± 10% from the recited amounts.

Furthermore, in the foregoing detailed description, it can be seen that various features are grouped together in various examples for the purpose of streamlining the disclosure. This method of disclosure is not to be interpreted as reflecting an intention that: the claimed embodiments require more features than are expressly recited in each claim. Rather, as the following claims reflect, subject matter is claimed in less than all features of any single disclosed example. Thus the following claims are hereby incorporated into the detailed description, with each claim standing on its own as a separately claimed subject matter.

While the foregoing has described what are considered to be the best mode and/or other examples, it is understood that various modifications may be made therein and that the subject matter disclosed herein may be implemented in various forms and examples, and that they may be applied in numerous applications, only some of which have been described herein. It is intended that the appended claims claim cover any and all such modifications and variations as fall within the true scope of this present concept.