GRAIN BOUNDARY ENHANCED UN AND U3Si2 PELLETS WITH IMPROVED OXIDATION RESISTANCE

Patent metadata

Field	Value
Publication number	US-2019237206-A1
Application number	US-201916260889-A
Country	US
Kind code	A1
Filing date	Jan 29, 2019
Priority date	Jan 30, 2018
Publication date	Aug 1, 2019
Grant date	—

Abstract

A method of forming a water resistant boundary on a fissile material for use in a water cooled nuclear reactor is described. The method comprises mixing a powdered fissile material selected from the group consisting of UN and U3Si2 with an additive selected from oxidation resistant materials having a melting or softening point lower than the sintering temperature of the fissile material, pressing the mixed fissile and additive materials into a pellet, sintering the pellet to a temperature greater than the melting point of the additive. Alternatively, if the melting point of the oxidation resistant particles is greater than the sintering temperature of UN or U3Si2, then the oxidation resistant particles can have a particle size distribution less than that of the UN or U3Si2.

First claim

What is claimed is:

1 . A method comprising:
mixing a powdered fissile material selected from the group consisting of UN and U 3 Si 2 with an additive selected from oxidation resistant materials, the powdered fissile material comprising grains having grain boundaries; pressing the mixed fissile and additive materials into a pellet; and, sintering the pellet to a temperature greater than the melting point of the additive sufficient for melting of the additive into the grain boundaries of the fissile material and densifying the pellet.

2 . The method recited in claim 1 wherein the additive is selected from the group consisting of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, and magnesium, alloys containing at least 50 atomic % thereof, magnesium nitride, ZrSi 2 , ZrSiO 4 , CrSi 2 , BeO, and UO 2 and glassy materials.

3 . The method recited in claim 2 wherein the glassy material is a borosilicate glass.

4 . The method recited in claim 1 wherein the oxidation resistant additive materials have a melting point lower than the sintering temperature of the fissile material.

5 . The method recited in claim 1 wherein the oxidation resistant additive materials have a melting point greater than the sintering temperature of the fissile material and a particle size distribution less than that of the UN or U 3 Si 2 .

6 . The method recited in claim 1 wherein the additive is present in amounts less than 20 weight percent of the fissile material.

7 . The method recited in claim 1 wherein sintering is selected from the group consisting of pressureless sintering, hot pressing, hot isostatic pressing, spark plasma sintering, field assisted sintering, and flash sintering.

8 . The method recited in claim 1 wherein the melting point of the additive is at least 200° C. below the sintering temperature.

9 . The method recited in claim 1 wherein the melting point of the additive is from 200° C. to 300° C. below the sintering temperature.

10 . The method recited in claim 1 wherein the particle size distribution of the additive is less than 10% of the particle size of the UN or U 3 Si 2 .

11 . The method recited in claim 1 wherein the particle size distribution of the additive is less than 1% of the particle size of the UN or U 3 Si 2 .

12 . The method recited in claim 1 wherein the additives may be coated to the U 3 Si 2 or UN powders to form protective layers through vapor deposition before pressing into pellets and sintering.

13 . The method recited in claim 12 wherein vapor deposition is selected from the group consisting of physical vapor deposition, chemical vapor deposition, and atomic layer deposition.

14 . The method recited in claim 1 wherein vapor deposition is used to coat the outside of the unsintered pellet with the additive and to penetrate the additive into the pellet body of the UN or U 3 Si 2 .

15 . A nuclear fuel comprising:
a pellet comprised of compressed and densified grains of a fissile material selected from the group consisting of UN and U 3 Si 2 , the grains having grain boundaries; and an oxidation resistant additive coating at least a portion of the grain boundaries of the fissile material.

16 . The nuclear fuel recited in claim 15 wherein the additive is selected from the group consisting of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, and magnesium, alloys containing at least 50 atomic % thereof, magnesium nitride, ZrSi 2 , ZrSiO 4 , CrSi 2 , BeO, and UO 2 and glassy materials.

17 . The nuclear fuel recited in claim 16 wherein the glassy material is a borosilicate glass.

18 . The nuclear fuel recited in claim 15 wherein the additive is present in amounts less than 20 weight percent of the fissile material.

Assignees

• Westinghouse Electric Co Llc

Inventors

• Lahoda Edward J
• Xu Peng
• Oelrich Jr Robert L
• Shah Hemant
• Wright Jonathan
• Cai Lu

Classifications

• G21C3/626 Primary

  Coated fuel particles · CPC title

• G21C3/045 Primary

  Pellets · CPC title

• G21C3/22

  with fissile or breeder material in contact with coolant · CPC title

• G21C3/62

  Ceramic fuel · CPC title

• G21C21/12

  by hydrostatic or thermo-pneumatic canning {in general by pressing without lengthening, e.g. explosive coating} · CPC title