INTEGRATED IN-VESSEL NEUTRON SHIELD
To reduce size and mass of a nuclear reactor system, an integrated in-vessel shield separates the role of a neutron reflector and a neutron shield. Nuclear reactor system includes a pressure vessel including an interior wall and a nuclear reactor core located within the interior wall of the pressure vessel. Nuclear reactor core includes a plurality of fuel elements and at least one moderator element. Nuclear reactor system includes a reflector located inside the pressure vessel that includes a plurality of reflector blocks laterally surrounding the plurality of fuel elements and the at least one moderator element. Nuclear reactor system includes the in-vessel shield located on the interior wall of the pressure vessel to surround the reflector blocks. In-vessel shield is formed of two or more neutron absorbing materials. The two more neutron absorbing materials include a near black neutron absorbing material and a gray neutron absorbing material.
1 . A nuclear reactor system comprising:
a pressure vessel including an interior wall;
a nuclear reactor core located within the interior wall of the pressure vessel, wherein the nuclear reactor core includes a fuel element array of a plurality of fuel elements and at least one moderator element;
a reflector located inside the pressure vessel that includes a plurality of reflector blocks laterally surrounding the plurality of fuel elements and the at least one moderator element; and
an in-vessel shield located on the interior wall of the pressure vessel to surround the plurality of reflector blocks, wherein:
the in-vessel shield is formed of two or more neutron absorbing materials, and
the two more neutron absorbing materials include a near black neutron absorbing material and a gray neutron absorbing material.
2 . The nuclear reactor system of claim 1 , wherein:
the near black neutron absorbing material includes a composited ceramic material;
the gray neutron absorbing material includes a heavy metal material; and
the composited ceramic material and the heavy metal material form the in-vessel shield.
3 . The nuclear reactor system of claim 2 , wherein:
the composited ceramic material includes boron carbide (B 4 C), hafnium carbide (HfC), or gadolinium oxide (Gd 2 O 3 ); and
the heavy metal material includes tungsten (W), iron (Fe), Nickel (Ni), or copper (Cu).
4 . The nuclear reactor system of claim 3 , wherein:
the composited ceramic material further includes aluminum oxide (Al 2 O 3 ) or silicon carbide (SiC).
5 . The nuclear reactor system of claim 2 , wherein:
the composited ceramic material includes:
boron-10 carbide ( 10 B 4 C),
a boron-10 carbide and aluminum oxide composite ( 10 B 4 C—Al 2 O 3 ) of 50% boron-10 carbide by weight percent,
a boron-10 carbide and silicon carbide composite ( 10 B 4 C—SiC) of 50% boron-10 carbide by weight percent, or
a borated stainless steel alloy of 5% boron-10 by weight percent; and
the heavy metal material includes atomized tungsten heavy metal with a tungsten content greater than or equal to 90% by weight percent.
6 . The nuclear reactor system of claim 2 , wherein:
the composited ceramic material includes a plurality of composited ceramic particles with an average particle diameter greater than or equal to approximately 80 nanometers and less than or equal to approximately 100 microns.
7 . The nuclear reactor system of claim 6 , wherein:
the plurality of composited ceramic particles are embedded inside a heavy metal matrix of the heavy metal material to form the in-vessel shield.
8 . The nuclear reactor system of claim 7 , wherein:
the in-vessel shield is formed as a plurality of in-vessel shield tiles; and
the plurality of in-vessel shield tiles are disposed on the interior wall.
9 . The nuclear reactor system of claim 8 , wherein all or a subset of the plurality of in-vessel shield tiles are a curved polyhedron shape or a truncated portion thereof.
10 . The nuclear reactor system of claim 8 , wherein:
the plurality of in-vessel shield tiles include a base shape with an interlocking geometry pattern.
11 . The nuclear reactor system of claim 10 , wherein:
the interior wall is formed of a continuous surface or a plurality of discontinuous surfaces; and
the interlocking geometry pattern of the plurality of in-vessel shield tiles are joined to cover and collectively form an in-vessel shield liner on the continuous or discontinuous surfaces of the interior wall.
12 . The nuclear reactor system of claim 11 , further comprising:
an inner density of activation product within the nuclear reactor core; and
an outer density of activation product outside the nuclear reactor core;
wherein the outer density of activation product is lower than the inner density of activation product during operation of the nuclear reactor core.
13 . The nuclear reactor system of claim 1 , wherein:
a reflector block of the plurality of reflector blocks has a reflector thickness, a reflector diffusion coefficient, and a reflector macroscopic absorption cross section;
the in-vessel shield has an in-vessel shield thickness, an in-vessel shield diffusion coefficient, and an in-vessel shield macroscopic absorption cross section;
the reflector block and the in-vessel shield have a combined diffusion length, based at least on: the reflector thickness, the reflector diffusion coefficient, the reflector macroscopic absorption cross section, the in-vessel shield thickness, the in-vessel shield diffusion coefficient, and the in-vessel shield macroscopic absorption cross section; and
the reflector thickness and the in-vessel shield thickness added together are less than double the combined diffusion length.
14 . The nuclear reactor system of claim 1 , wherein:
each of the fuel elements includes a nuclear fuel;
the nuclear fuel includes a fuel compact comprised of coated fuel particles embedded inside a high-temperature matrix; and
the high-temperature matrix includes silicon carbide, zirconium carbide, titanium carbide, niobium carbide, tungsten, molybdenum, or a combination thereof.
15 . The nuclear reactor system of claim 14 , wherein:
the coated fuel particles includes tristructural-isotropic (TRISO) fuel particles or bistructural-isotropic (BISO) fuel particles.
16 . The nuclear reactor system of claim 15 , further comprising a plurality of control drums, wherein:
the control drums are interspersed or disposed within the reflector.
17 . The nuclear reactor system of claim 1 , wherein the nuclear reactor system includes a gas-cooled high-temperature nuclear reactor, a molten salt cooled nuclear reactor, fuel-in-salt nuclear reactor, or a sodium-cooled fast nuclear reactor.
18 . The nuclear reactor system of claim 1 , further comprising a coolant that flows through the plurality of fuel elements, wherein the coolant includes helium, FLiBe molten salt formed of lithium fluoride (LiF) and beryllium fluoride (BeF 2 ), sodium, He, HeXe, CO 2 , neon, or HeN.
19 . A method comprising:
selecting two or more neutron absorbing materials to form an in-vessel shield, wherein the two or more neutron absorbing materials include a near black neutron absorbing material and a gray neutron absorbing material;
eutectic sintering the near black neutron absorbing material to fabricate a ceramic absorbing powder;
kinetically mixing the gray neutron absorbing material and the ceramic absorbing powder to create an in-vessel shield mixture; and
cold press sintering the in-vessel shield mixture into an in-vessel shield.
20 . The method of claim 19 , wherein:
the near black neutron absorbing material includes a composited ceramic material; and
the gray neutron absorbing material includes a heavy metal material.
21 . The method of claim 20 , wherein:
the composited ceramic material includes boron carbide (B 4 C), hafnium carbide (HfC), or gadolinium oxide (Gd 2 O 3 ); and
the heavy metal material includes tungsten (W), iron (Fe), Nickel (Ni), or copper (Cu).
22 . The method of claim 21 , wherein:
the composited ceramic material further includes aluminum oxide (Al 2 O 3 ) or silicon carbide (SiC).
23 . The method of claim 20 , wherein:
the composited ceramic material includes at least one of:
boron-10 carbide ( 10 B 4 C),
a boron-10 carbide and aluminum oxide composite ( 10 B 4 C—Al 2 O 3 ) of 50% boron-10 carbide by weight percent,
a boron-10 carbide and silicon carbide composite ( 10 B 4 C—SiC) of 50% boron-10 carbide by weight percent, or
a borated stainless steel alloy of 5% boron-10 by weight percent; and
the heavy metal material includes atomized tungsten heavy metal with a tungsten content greater than or equal to 90% by weight percent.
24 . The method of claim 20 , wherein:
the composited ceramic material includes a plurality of composited ceramic particles with an average particle diameter greater than or equal to approximately 80 nanometers and less than or equal to approximately 100 microns.
25 . The method of claim 19 , wherein:
the ceramic absorbing powder includes a plurality of composited ceramic particles with an average particle diameter greater than or equal to approximately 80 nanometers and less than or equal to approximately 100 microns.
26 . The method of claim 24 , wherein:
the cold press sintering the in-vessel shield mixture into the in-vessel shield includes embedding the plurality of composed ceramic particles inside a heavy metal matrix of the heavy metal material to form the in-vessel shield.
27 . The method of claim 26 , wherein the cold press sintering the in-vessel shield mixture into the in-vessel shield further includes forming the in-vessel shield as a plurality of in-vessel shield tiles.
28 . The method of claim 27 , further comprising:
disposing the plurality of in-vessel shield tiles on an interior wall of a pressure vessel.
29 . The method of claim 28 , wherein all or a subset of the plurality of in-vessel shield tiles are a curved polyhedron shape or a truncated portion thereof.
30 . The method of claim 28 , wherein:
the plurality of in-vessel shield tiles include a base shape with an interlocking geometry pattern.
31 . The method of claim 30 , wherein:
the interior wall is formed of a continuous surface or a plurality of discontinuous surfaces; and
the method further comprises covering the continuous or discontinuous surfaces of the interior wall with the plurality of in-vessel shield tiles by joining the interlocking geometry pattern.
32 . The method of claim 30 , further comprising selecting the base shape of the in-vessel shield tiles, based in part on:
(i) tolerance towards swelling;
(ii) reduction of neutron streaming paths; or
(iii) a combination thereof.
33 . The method of claim 32 , wherein the base shape is a curved polyhedron shape or a truncated portion thereof.
34 . The method of claim 19 , wherein the selecting the two or more neutron absorbing materials is based on:
(i) an in-vessel shield thickness of the in-vessel shield;
(ii) a geometric configuration of the in-vessel shield;
(iii) an estimated reduction in vessel fast fluence of a pressure vessel, in which the pressure vessel encapsulates the in-vessel shield, and the in-vessel shield encapsulates a nuclear reactor core;
(iv) an anticipated lifetime of the in-vessel shield; or
(v) a combination thereof.
35 . The method of claim 19 , wherein the selecting the two or more neutron absorbing materials is based on:
(i) an interior diameter of a pressure vessel;
(ii) an exterior diameter of a nuclear reactor core;
(iii) a reflector thickness of a reflector; or
(iv) a combination thereof.
36 . The method of claim 19 , wherein eutectic sintering the near black neutron absorbing material to fabricate the ceramic absorbing powder is performed with spark plasma sintering.
37 . The method of claim 19 , further comprising:
selecting a reflector block to pair with the in-vessel shield; and
selecting an in-vessel shield thickness of the in-vessel shield;
wherein:
the reflector block has a reflector thickness, a reflector diffusion coefficient, and a reflector macroscopic absorption cross section;
the in-vessel shield has an in-vessel shield diffusion coefficient and an in-vessel shield macroscopic absorption cross section;
the reflector block and the in-vessel shield have a combined diffusion length, based at least on: the reflector thickness, the reflector diffusion coefficient, the reflector macroscopic absorption cross section, the selected in-vessel shield thickness, the in-vessel shield diffusion coefficient, and the in-vessel shield macroscopic absorption cross section; and
the reflector thickness and the selected in-vessel shield thickness added together are less than double the combined diffusion length.
38 . The method of claim 19 , wherein selecting the near black neutron absorbing material includes selecting an isotopically-tailored near black neutron absorbing material.
39 . The method of claim 19 , further comprising:
selecting a nuclear reactor system; and
wherein:
selecting the two or more neutron absorbing materials to form the in-vessel shield is based on:
(i) a pressure vessel diameter of a pressure vessel of the nuclear reactor system;
(ii) a fast neutron fluence to the pressure vessel; and
(iii) neutron fluence outside of the nuclear reactor system.
40 . The method of claim 19 , further comprising:
mounting the in-vessel shield on an interior wall of a pressure vessel.