CA1081673A - Tritium removal and retention device - Google Patents

Abstract

TRITIUM REMOVAL AND RETENTION DEVICE

ABSTRACT OF THE DISCLOSURE
Apparatus comprising a two layered composite with an internal core of zirconium or zirconium alloy which retains tritium, and an adherent nickel outer layer which acts as a protective and selective window for passage of the tritium.

Description

~ AC~GROUND OF THE INVENTION
Field of the Invention:
This invention provides a device to remove and store tritium from a gaseous medium, and a method for manu-facturing the device. It specifically provides a device which may be incorporated in a fuel rod of a nuclear reactor to minimize release of tritium to the reactor coolant.
Description of the Prior Art:
The operation of a nuclear reactor necessarily forms tritium. As a product of ternary fission3 which ~s typically the largest source of tritium~, tritium is formed within the solid matrix of uranium containing pellets and other fuels, typically encased in metal tubes or cladding.
Most water reactors utilize fuel cladding of zirconium alloy, known more commonly as Zircaloy, and a typical ccm-mercial reactor includes thousands of such rods. The pro-perties of typical zirconium alloys are defined in ASTM
Standard B 353-71, "Wrought Zirconium and Zirconium Alloy Seamless and Welded Tubes for Nuclear Service." After formation of tritium in the solid fuel pellet matrix, the gaseous tritium may diffuse through the pellet matrix and into the vold volume between the fuel pellets and fuel A.
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45,970 cladding, as do a variety of other fission product gases.
These fisslon product gases are then free to migrate through-out the fuel rod, and contrlbute to a pressure buildup withln the cladding. The tritium, and other fission product gases, circulate inside the fuel rod due to convection. A
typical fuel rod includes a plenum area at top of the rod, where, due to the free volume, these gases tend to collect.
Although the radioactivity emitted by tritium is a weak beta emission, and although it has a relatively short biological half-life (ten days), tritium has a relatively long radioactive half-life (twelve years). Also, tritium will readily diffuse through most materials, including materials such as zirconium, alloys of zirconium, and stain-less steel, whlch are typically used as fuel rod cladding.
Because pressurized water reactors in operation today uti-lize boric acid in the coolant for power level control, ; tritlum ls also formed wlthln the reactor coolant ltself.
Once trltlum reacts wlth water to form HTO, it is technl-cally difficult and very costly to separate.
Regulatory authorities have therefore placed strlngent restrlctions on allowable releases of tritium to the environment. One way to lower the tritium inventory in the reactor coolant, and hence the amount of tritium which may be discharged to the environment, is to provide a means within each fuel rod to speclfically collect and store the tritlum produced wlthln the fuel pellets which diffuses into the void volume. This invention provides such means, which further are easily removable from the fuel rod during re-, processing, As tritium is widely used as a tracer element and in the medical and other fields, being able to simply .

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45,970 lO~t73 and less expensively recover the tritium, as compared to recovery from an aqueous solution, is a further benefit provlded by thls invention. The device disclosed herein also may be utilized in other functions wlthin a nuclear plant, as well as in other applications where it is desir-able to remove tritlum from a gaseous medlum.
Although many systems and modes of operation have been used and proposed to control trltlum subsequent to its enterlng the reactor coolant, in accordance with this inven-tion tritium is specifically collected and controlled within the fuel rod itself. Thls lnvention, in the preferred embodiment, does so by means of a device consisting of an inner core of zirconium or alloys of zlrconlum, covered on all surfaces wlth an adherent layer of nlckel, whlch nickel layer acts as a selectlve and protectlve wlndow for the passage of trltlum. At reactor operating temperatures, the layer of nlckel is generally unreactlve to species in the fuel rod environment, including any high temperature mols-ture present. The nlckel layer, however, ls selectively permeable to tritium, also allowing passage of such atomi-cally small and avallable lsotopes as hydrogen and deuterium.
Once through the adherent nlckel layer, the trltlum reacts with the inner core of zirconium alloy to form a solid solution or hydrlde, and is flxed wlthin the zlrconium alloy matrix until such tlme as lt is desirable to remove the ; trltlum.
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Other devlces have been disclosed which may per-form a somewhat slmllar functlon, although of different deslgn and wlthout the trltium selectlvity provided by the device of the instant invention. A United States patent s'' ~"

45,970 108~73 issued to L. N. Grossman, No. 3,742,367, June 1973, disclosesa non-destructive detection process for nuclear fuel rods.
The Grossman patent provides, in part, a device consisting of a homogeneous alloy of titanium, zirconlum, and nickel, as differentiated from the layered window of nickel over a zirconium alloy core of this invention. An amount of the alloy of the Grossman patent is placed in the fuel rod during assembly. The assembled rod is then heated prior to installation ln the reactor, to vaporize moisture, and free from the fuel pellet matrix gases such as hydrogen, oxygen, nitrogen, carbon monoxide, and carbon dioxide which react with the alloy. The alloy within the rod is then examined by neutron radiography to detect metallic hydrides prior to puttlng the fuel rod into operation in a reactor. Detection of moisture provides an indication that sufficient heating of the fuel has occurred to remove moisture from the fuel pellets. This reaction of the named elements and compounds with the homogeneous alloy is designed to occur to eliminate subsequent embrittlement and induced stresses in the clad-ding durlng reactor operation. Since the homogeneous alloyof the arossman patent is reactive with hydrogen, it should also be reactive with tritium released during reactor operation.
However, it is seen that significant differences exist between this invention and the teachings of the Gross-- man patent. Most notably, these distinctions include dif-ferences in elemental composition and in the methods of ~oining the elements. The prior art device consists of titanium, zirconium, and nickel, compared to a nickel coated 3 zirconium alloy of this invention. More important, the ~;
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45, 970 ~08~6~73 prior art rorms these elements into a homogeneous alloy, with the reactions taking place on the surface and within the alloy. This lnvention, on the other hand, provides a two-layered composite device, containing zlrconium or zir~
conium alloy as an internal core and a layer of nickel on the exterior. The device disclosed herein is much more selective as to what will pass through the nickel layer or window and react inside the device with zirconium alloy.
Addltlonally, the prior art alloy is used to remove moisture and other impurity gases from a fuel rod prior to reactor operation, whereas the device disclosed herein performs its functlon subsequent to reactor startup and durlng the life of the fuel rod. The devlce disclosed herein further has significant benefits in terms of tritium recovery and separa-tion subsequent to reactor operation.
SUMMARY OF THE INVENTION
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This invention provides a device for removing and retaining triti.um from a gaseous medium, and also a method of manufacturing the device. The device, in the preferred embodiment, consists of an inner core of zirconium alloy, B deslrably an alloy known commonly as Zircaloy~ , and an outer adherent layer of nickel which acts as a selective and protective window for passage of tritium. The tritium then reacts with or is absorbed by the zirconium alloy, and is retained until such time as it is desirable to remove it during reprocesslng. In the main embodiment, a small elon-gated annular shaped device is incorporated within a re-tention spring in the upper plenum of a nuclear fuel rod, such that it will remove tritium formed within the rod during the fissioning process which migrates outside the . . .

45 ,970 ~OB1673 fuel pellet matrix.
BRIEF DESCRIPTION OF THE DRAWINGS
The functions and advantages of this invention will become more apparent from the following description and accompanying drawings, in which:
Flgure 1 is a simplified schematic, in cross section, of a fuel rod;
Flgure 2 is an elevation view, in cross section, of a tritium removal and storage device;
Figure 3 is a view, in cross section, taken at III-III of Figure 2;
Figure 4 is an elevation view of a spring used in the upper plenum of a fuel rod;
Figure 5 is an elevation view of the device of Figure 3 contained within the spring of Figure 4;
Figure 6 is a schematic of a test furnace; and Flgure 7 is an elevation view, in partial cross section, of a test capsule.
DESCRIPTION OF THE PREFERRED EMBODIMENTS
During the operation of a nuclear reactor tritium i8 formed. Once tritium forms HTO in the reactor coolant, it is technically difficult and economically costly to - separate. Also, there are stringent regulatory restrictions today which limit the release of tritium to the environment.
For these reasons, it is highly des~rable to minimize the amount of tritium combining with the reactor coolant. In all nuclear reactors, tritium is produced as a byproduct of ternary fissions. This is the largest source of tritium production in many reactors. It is also formed as a product 3 of other reactions, such as reactions with boron-10, lith-45, 970 10816~73 ium-6 and llthium-7, and deuterium.
As the largest source of tritium production is within the fuel itself, typically comprising uranium, or plutonlum, or thorium, among others, it is desirable to remove the tritium by means in close proximity to this source. A typical form of nuclear reactor fuel is a stack Or solld sintered pellets 10 of uranium dioxide encased in a sealed metal cladding 12, as shown in Figure 1. End plugs 14 hermetlcally seal the cladding 12 at the top and bottom.
10 The most widely used cladding materials are stainless steel B and alloys of zirconium, such as Zircaloy-4. Tritium is formed ln the fuel pellet 10 matrix, and migrates in a gas-eous phase to the void volume between the cladding 12 and the pellets 10. Because of its small atomic size, a signi-- ficant portion of the tritium in the void volume may diffuse through the fuel rod cladding 12, and into the reactor cool-ant. Also, tr:ltium may react to replace hydrogen atoms in the fuel cladding 12 or react with the cladding 12. It has been found that tritium diffuses through stainless steel in 20 a reactor environment at a high rate, the rate being signi-ficantly hlgher than its diffusion through zirconium alloys.
Tritium also reacts with the zirconium alloy cladding to form an hydride, lessening the release of tritium to the reactor coolant. An ideal device which will remove and store thls ternary produced tritium should have the fol-lowlng characterlstics: (1) it should remove and store tritium in a gas phase within a fuel rod during the operat-ing life of the rod, (2) the removal function should not be limited by residual air, water vapor, or other gases nor-30 mally present in fuel rods, such as C0, C02 and CH4, among , .

45,970 10~1673 others, (3) it should reduce the reaction of tritium with the rod cladding, (4) it should be inexpensive to manufac-ture as compared to the cost associated with dealing with excess tritium in the reactor coolant, (5) the device should be easily adaptable to current and future ~uel rod designs, and (6) it should provide a relatively inexpensive source of trltium during fuel reprocessing, as compared to removal of tritium from an aqueous solution, for medical, tracer and other uses.
The apparatus disclosed herein meets all of these characteristics. The apparatus, for use in a fuel rod, consists of a two-layered composite of materials, and can be produced in almost any geometric shape desired. The inner -core 16 (Flgures 2 and 3) can be a number of materials, as long as the material meets the criteria of removing tritium in a gaseous phase within a reactor environment and retaining it through absorption or chemical reactlon until such time as it is specifically desired to remove the tritium. Tests performed and discussed below were based upon an inner layer ¦ ~ 20 of zirconium alloy, such as Zircaloy-4 which is the pre-ferred material of the inner core 16. Pure zirconium, as well as other zirconium alloys such as Zircaloy-2, among others, may also be used. The outer layer 18 of the device is an adherent layer of nickel bonded to the inner core 16.
The nickel layer 18 acts as a selective and protective barrier and allows passage of tritium as well as hydrogen and deuterium, at reactor operating temperatures. In a higher temperature environment, other materials, such as dissociated hydrocarbons, could pass through the nickel window if a sufficient quantity of these materials were .:

45, 970 ~08~G~73 available. Tests have shown, for the size device necessary for incorporation in fuel rods, on the order of 1.5 grams, that the nickel window should constitute roughly five to twenty percent by weight of the device, with a more ideal range between eight and twelve percent by weight. The nickel should be evenly distributed over all surfaces of the lnner core 16, such that about four to six percent by weight is on each slde of the inner core 16. Below this level, experimental results have shown that the removal rate is lessened. It further may allow buildup of an oxide layer on the device which also partially poisons its tritium removal function. This poisoning would be the effect if only a surface of zirconium alloy were placed inside the fuel rod, wlthout the overlaying protective nickel window. Although the apparatus will function above the preferred weight percent level, to lncrease reactor efficiency, lt is desirable to minimize the amount of neutron poisoning material in the reactor core. As there are typically in excess of 20,000 fuel rods in a typical reactor, even a small device in each rod will have an effect on neutron absorption. It is there-fore preferable not to exceed the eight to twelve percent by weight level. For the device for use in a fuel rod, an ; inner core 16 of thickness between 0.01 and 0.03 inch will be consistent with an outer layer of eight to twelve percent by weight. It should be noted that if the inner core is not completely covered with a nickel layer, the device will still operate to perform its removal function, but with decreased efficiency.
Since the device consists of two adherent layers, the bonding of these layers is critical, and must be care-_g_ . , . .. , . ~ .

45,970 108~6~3 fully controlled in manufacturing. The heat treatment iscrucial. The method disclosed herein includes cleaning the surface of the inner layer of zirconlum alloy to nuclear specifications. Allowable impurity levels of the zirconium alloy are as typically standard in the industry for fuel rods, and are defined in ASTM V-353. Subsequent to clean-ing, hlgh purity nickel is deposited upon the surface of the lnner layer by commercially well known manufacturing techni-ques. These techniques may include electroplating, vacuum deposition, or a liquid dip technique, among others, as long as the amount of the deposit is controlled. Controlled sputtering techniques may also be used. Subsequently, the zirConium alloy core 16 with nickel deposit 18 is thermally treated in a vacuum maintained at about 10 6 millimeters of mercury. It is heated to a temperature between 775C and 825C, and maintained for a minimum of three hours. It should not be heated more than several hours beyond this amount of time. This treatment activates the surfaces of the zirconium alloy by diffusion of the nickel into the zirconium alloy surface. This thermal vacuum implantatlon provldes the protective and selective layer of nickel 18, which is, as shown by testing discussed in the examples below, unreactive in the presence of water vapor and fission product gases, but permeable to tritium, hydrogen, and deu-terium in a reactor environment. The time and temperature relation of the heat treatment is critical as an excess of either would allow the material to form an homogeneous alloy, and an insufficiency would not provide sufficient bonding. As discussed above, an alloy would be poisoned by the other available gases within a fuel rod, thereby limit-t -10-i 45,970 1081f~73 ing its tritium removal and storage function.
Among the prime characteristics desired of a tri-tium removal and storage device for use in a nuclear fuel rod is that it does not add significant costs to the manu-facturing process, and that it does not in any way adversely affect reactor operation. An apparatus as hereafter des-cribed provides such desirable results.
Most fuel rods of the type discussed include a void plenum 20 (Figure 1) in the fuel rod, typically in the upper regions to allow for the buildup of fission product gases. The plenum 20 area may also be used for inclusion of mechanical components, most notably a retention spring 22 (Figure 4) or other retention device to maintain proper axial positlon of the fuel pellet 10 stack and allow for fuel axial expansion. An elongated annular shaped tritium removal and storage devlce 24 may easily be placed wlthin the spring 22, as shown ln Figure 5, and function to remove and store ternary fission produced tritium during the operating life of the fuel. The device 24 shown is approximately two 20 inches in length with a 0.2 inch outer diameter and a 0.03 inch wall thickness. This device 24 may be placed within the spring 22 in a fuel rod without complication, during fuel manufacture. A typical spring 22 as used in pressurized water reactor fuel rods is approximately seven inches long with a 0.35 inch outside diameter and a 0.22 inch inside diameter. An end cap 26 affixed to one or both ends of the spring 22 to retain the device 24 in the plenum 20 area may also be used. It may be a stainless steel disc with~ or without, a central aperture 28 to provide a free path for !

transport of tritium to the device 24. For example, the --11-- , '``~

45, 970 1081t~73 device 24 may be placed inside the spring 22 and then two end caps 26 spot welded to each end of the sprlng 22. The spring 22 would then be placed in the fuel rod as is presently done, with perhaps the added step of mere visual inspection to ascertain that each spring 22 does contain a tritium removal device 24. Alternatively, the device 24 could be placed above the spring 22, or in rods not using a spring or other retention devlce, it could be placed in the plenum with means, such as a small plate, separating the device 24 from immediate contact with the fuel pellets 10.
In accordance with the invention, a series of tests were performed to ascertain the ability of the inven-tlon to remove and store tritium. The tests were arranged to simulate a reactor environment, including placing a tritium removal and retention test device in competition with the zirconium alloy cladding for the tritlum. Early tests also simulated the abillty of the inventlon for tritium removal and retention in competition with several other mediums.
It should be noted that in all of tests, deuterium, whlch can be more easily obtained, was used as opposed to tritium, which is a typical laboratory technique. Deuterium is easier to work with in a laboratory environment and posed less of a health concern than would tritium. Tritium and deuterium are similarly sensitive to surface barriers and ; isotopic exchange reactions. Also, well recognized in the art, is that similar recovery and detection techniques may be used for tritium and deuterium. As among any isotopes of a given element, the kinetic relationships of tritium and deuterium are similar. Further, briefly statedg the dif-,,~

f 45,970 ~081~73 fusion coefficient o~ deuterium and tritium through materials as taught herein is similar, with tritium having a somewhat lower coefficient than deuterium.
EXAMPLE I
The first laboratory test involved a competition among eight differing devices. All of the samples were of slmilar mass. They were cleaned with acetone and then dried and weighed prlor to being inserted into a quartz furnace tube 60 (Figure 6). The samples were either in a thin foil sheet ~orm, approximately 10 mils thick, or powders, as lB noted below, and the Zircaloy-4 cladding samples were cuts from actual fuel cladding. The powders were contained in high purlty platinum crucibles which, when subsequently analyzed, had essentially no deuterium content. The furnace tube was then evacuated, and placed in a furnace 62. The eight samples were suspended within the tube 60 by a quartz sample holder 64. The samples were then heated to 650C
while the gas pressure was observed and the furnace 62 composition analyzed by mass spectrometry. When the gas atmosphere in the furnace 62 showed little or no change~ the furnace 62 temperature was reduced. When the temperature of the furnace tube 60 reached 310C, deuterium gas at a pressure of 1.4 millimeters of mercury was added to the furnace tube 60, corresponding 'co about 1.2 cubic centimeters. The pressure was monitored continuously by a metal capacitance manometer and gradually decreased to 0.44 millimeters of ~ mercury after forty-two hours. The furnace 62 was then ; cooled to room temperature and mass spectrometric analysis ; performed on the gas atmosphere. It showed that 0.16 cub~c centimeters of deuterium remained in the system. The samples . ~ :

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45, 970 108~673 were then weighed, the deuterium extracted from each sample by a hot vacuum extraction technique and mass spectrometrD
analysls. This involved heating each sample to about 1050~C, which is above the temperature range (800~C-850C) at which hydrogen and its isotopes dissociate from zirconium and Zlrcaloy-~. The amount of deuterium was quantitatIvely determined by mass spectrometry. Before each experiment discussed herein, the experlmental system was calibrated wlth National Bureau Or Standards (NBS) Hydrogen Standards.
1~ The results are shown in Table I. The letters "A"
through "H" representing each sample correspond t~ the letters on Figure 6 showing the relative locatlon of the samples ln the furnace tube 60. Sample "A" represents a zirconium-titanium powder, with 6.2% by weight nickel;
sample "B" a zirconium-titanium powder with a 3.9% by weight nickel; samples "C" and "D" were Zircaloy-~ cladding; sample "E", as discussed herein, a Zircaloy-~ core with a 5.7% by weight outer nickel layer; sample "F" a zirconium metal core with a paladium outer layer; sample "G'l a zlrconium-titanium alloy with a paladium coating, and sample "H" a zirconium core with a ten weight percent vanadium coating.
As shown from the three data columns of Table I, the Zircaloy-4 core with a nickel outer layer proved far superior to the other samples in removing and retaining the deuterium, even in competition with a wide variety of other samples:

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' 45, 970 ~081~73 TABLE_I

Sample (ppm) (cc)(cc/gm sample) A 747 .24874.180 B 669 . 21203.744 C 4.8 .oo56 .026 D 2.9 .0032 .016 E ?243 528612.556 11.5 .oo58 .o64 G 34.2 .0102 .191 H 2.8 .oo36 .016 EXAMPLE II
A second competltion test was run, using the same experimental procedure as discussed in reference to Example . I. Among the samples here, however, were included three comprising oores of Zircaloy-~, wlth varying weight percentage nlckel outer layers. Samples "B-2", "C-2", and "E-2" com-prised a ten (10%) percent, a 5.7%, and a 3.3% nickel layer, respectlvely. Sample "A-2" was Zircaloy-4 cladding material;
20 sample "D-2" a zlrconium-titanium powder with 6.2 weight percent nickel; sample "F-2" a zirconium-titanium powder with 3.9 weight precent nickel; sample "G-2" Zircaloy~in a thin foil (.005" thick) form; and sample "H-2" Zircaloy-~7 claddlng materlal.

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~ ' 45,970 ~081~73 TABLE II

Sample (ppm) (cc) (cc/gm sample) A-2 1.2 .0013 .0067 B-2 122 02~3 6832 C-2 45.1 0123 2526 D-2 4.6 .0018 .0258 E-2 2.4 ooo6 0134 F-2 2.3 .0007 .0129 G-2 11.2 .0026 .0627 H-2 0.6 .0007 .0034 As shown from Table II, the samples including an inner core Or Zircaloy-~7and outer layers of nickel dld quite well in deuterium adsorption. Further, it is most evident that the deuterium removal ability significantly increased with increasing weight percentage of nickel.
EXAMPLE III
A third competition test was run, utilizing the . ~/77 same procedure, and again the Zircaloy-4 core with a ten 20 weight percent nickel outer layer showed far superior. In Table III, sample "A-3" was Zircaloy-4 cladding; "B-3"
Zircaloy-/~7 foil with a ten percent nickel outer layer;
sample "C-3" zirconium-titanlum powder with 7.75 weight percent copper; sample "D-3" zircor~ium-titanium powder with 12.1 weight percent nickel; sample "E-3" zirconium-titanium powder with twelve weight percent copper; sample "F-3"
zirconium-titanium powder with 6.5 weight percent nickel;
sample "H-3" Zircaloy-~ foil; sample "I-3" Zircaloy ~ clad-ding.

45,970 ~81673 TABLE III

Sample tppm) (cc)(cc/gm sample) A-3 0.9 .001 0.005 B-3 407 100 2.279 C-3 29 .003 0.160 D-3 57 . oo8 o . 319 E-3 11 .002 0.062 F-3 19 .003 0.106 H-3 17 .003 0.095 I-3 1 .001 o.oo6 ::
EXAMPLE IV
A fourth competition test was similar to those discussed above. Here, however, all the samples were an-nealed at 660C for fifteen hours in a vacuum before the deuterium pressure addition. Again, the Zlrcaloy-4 with a ten percent by weight nickel outer layer proved by far to be ~ superior, and the adsorption significantly increased.
; Sample "A-4" was a Zircaloy-~ cladding sample; sample "B-4"
the Zircaloy-~ with nickel outer layer; sample "C-4" a zirconium-titanium powder with 7.75 weight percent copper;
sample "D-4" a zirconium-titanium powder with 12.1 weight , . . .
; percent nickel; sample "E-4" a zirconium-titanium powder ~ ~ .
with twelve weight percent copper; sample "F-4" a zirconium-titanium powder with 6.5 weight percent nickel; sample "G-4"
Zircaloy-4 foil; and sample "H-4" Zircaloy-~ cladding material.

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45, 970 1(~81~73 TABLE IV

Sample(ppm) (cc) (cc/gm sample) A-4 4.3 .0049 .0241 B-4 1413 .325 7.92 C-4 40 .0051 .224 : D-4 14 .0018 .079 E-4 16 .0021 .090 F-4 8 .0012 .045 G-4 6 .0015 .034 H-4 2 .0026 .011 EXAMPLE V
Later tests were performed and arranged to simu-late a reactor environment, including placing a tritium removal and retention test device 30 in competition with B Zircaloy-4 cladding. A test apparatus was arranged, and is shown ln Flgure 7. The test apparatus, referred to here-inafter as the "test capsule" 40, included fuel rod test Tr,q cladding 32 of Zircaloy-4. The capsule 40 was approximately 20 ll-l/2 inches ln length. Also included in the test capsule 40 were end plugs 34 o~ Zircaloy-4, a test tritium removal and storage devlce 30, a deuterium gas generator 36, and a glass tube spacer 38. The test device 30 was a rod prepared as discussed above, with a nickel layer of twelve percent by weight, and was approximately one and one-half inches in length and 0.2 inches in outer diameter. As shown, it was placed in the upper end of the test capsule 40. At the lower end of the test capsule 40 was placed the deuterium generator 36. The generator 36 was approximately one inch 45,g70 ~0816'73 in length with a 10 mil wall thickness of nickel and a 3/16 inch outside diameter. The generator 36 was made by taking a 3/16 inch diameter high purity nickel rod, and drilling the inside of the rod to give the desired wall thickness.
The bottom surface of the nickel rod was not drilled through.
Then, a controlled amount of deuteriated water (D20) was placed in the nickel shell. Also placed within the shell was high purity lron (Fe) wire, in coil form. While main-taining the lower portion of the deuterium generator 36 in a liquid nitrogen solution to solidify the deuteriated water, the upper portion of the nickel shell was welded shut.
After cooling, the generator 36 was then placed in the lower portion of the test capsule 40, which previously had been ~ealed by welding on one of the end plugs 30. The tube spacer 38 was a sealed glass tube approximately 7-1/8 inch long that made a loose sliding fit with the test rod clad-ding 32. The test device 30 was then inserted in the cap-sule 40, and the upper end plug 34 was welded in place, sealing the capsule at approximately one atmosphere of helium. A second test capsule was constructed, the only difference being the inclusion of a capillary tube placed ad~acent the test device 30. The capillary tube contained 260 micrograms (~4gm) of water. The tube ruptured at test temperature, releasing high temperature water vapor.
To run the test, the capsule 40 was placed in a gradient furnace that heated the cladding wall 32 opposite I the glass spaoer 38 and the deuterium generator 36 to slightly higher temperature than the test device 30. The device 30 t` ran at approximately 320C while the cladding 32 wall tem-30 perature varied from 380C to 320C. The higher temperature 1, --19--. ~

45, 970 1081~73 area was between the deuterium generator 36 and the device 30. The iron wire reacted with the deuteriated water to form a combination of Fe3O4 and Fe2O3 at approximately 300C, and liberated the deuterium, which freely passed through the nickel wall of the generator 36. The glass spacer 38 formed a small annulus for transport of the deu-terium to the test device 30, simulating the annulus between the fuel pellet 10 stack and the cladding 12 inside diameter in an actual fuel rod. The test was run in a controlled argon atmosphere that was monitored for escaping deuterium;
none was observed. The test capsule was held at temperature for seven days, and then cooled to room temperature.
Multiple analyzes were then performed upon the test capsules. Puncture and recovery of the internal gas atmosphere showed the only gases present to be helium and traces of hydrocarbons. Hydrogen and deuterium analyzes were then per~ormed on the test device 30 and at selected locations of the test cladding 32 represented by the arrows on Figure 7. The results are summarized in Table V. The letter "H" denotes the capsule with the 260 ~4gm of water addition.

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45,970 ~)81~73 TABLE V
Deuterium Hydrogen Capsule 1 lH 1 lH
Device ppm, wt. 7.0 5.3 22.9 35 ~4gm 30.4 23.0 96.8 149 percent 51.6 55.1 22.9 36.2 Claddin~
ppm, wt. o.8 0.53 9.2 7.5 ~gm 28.1 18.6323.0 262.0 percent 47.7 44.6 76.6 63.6 Generator ppm,~ wt. 0.3 0.1 1.4 0.55 ~gm o.4 0.2 2.0 o.8 percent 0. 8 0.35 0.5 0.2 As shown from Table V, the test device 30 con-tained about 52 percent of the initial deuterium. Less than ~ -1 percent of the deuterium remained in the generator 36.
The tests further showed that the added moisture had very .
20 little effect on the ability of the devlce 30 to remove the deuterium. In fact, it increased the removal and retention of deuterium, by the device 30, by several percent. This is believed due to buildup of an oxide film on the inner sur-face of the test cladding wall 32. The film could be seen by visual inspection, and was especially evident in the upper area of the cladding 32, where the water was released.
j There was no such film on the device 30 itself, as there was no reaction with the protective adherent nickel layer. As there is typically excess moisture on the surface and within ; 30 the fuel pellets 10 during manufacture, this same effect can ;~ be expected to be experienced during operation of the fuel within a reactor. An oxide film will be built up on the inner surface of the fuel cladding 12 early in the operating ;~ life of the fuel, thereby forming somewhat of a barrier to ~ - 21 -~, . .

,, 45,970 108~673 the interaction of tritium with the cladding 12. This will increase the efficiency of the tritium removal and retention device.
As a further result, the device 24 may perform a safety related function during plant operation. In the un-likely event that the cladding 12 of a fuel rod fails, reactor coolant water reacts with the inner surface of the fuel rod. The tritium removal and retention device 24 not only is lnert to the coolant but also retains its inventory in the presence of steam formed by the reactor coolant. In the unlikely event of fuel rod failure, the device will act to absorb free hydrogen, and will not act to catalyze the incoming coolant water, as might an alloy type device.
Another benefit of the device 24 disclosed herein is its ability to provide a source of tritium relatively less expensive than obtaining tritium from an aqueous solu-tion. Tritium has been used as a tracer element in many functions. It is also used in medical treatment. After a fuel rod containing the disclosed device is removed from a reactor, the device 24 can be easily removed and separately processed. Heating the device to a temperature in the range of 1100C in a vacuum maintained at 10 6 mm Hg releases the , entrained tritium, and also any entrained hydrogen, in a ~; gaseous phase. Separation of the tritium from this medium is significantly easier than separation from water.
It is therefore seen that the device disclosed herein provides a means to remove and store gaseous tritium.

It is particularly applicable to use in nuclear fuel rods, where its function is not reduced by residual water vapor or other fission product gases within a rod. It further limits ~'`'' ..

45, 970 10l3~673 the reaction of tritium with the fuel rod cladding, and canbe easily manufactured and incorporated in existing fuel rod types. It poses no additional problems in the unlikely event of fuel rod failure, and may provide tritium for medical, tracer, and other uses. It is apparent that many modifications and variations are possible in view of the above teachings. It therefore is to be understood that within the scope of the appended claims, the invention may be practiced other than as specifically described.

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Claims (9)

The embodiments of the invention in which an exclusive property or privilege is claimed are defined as follows:

1. A fuel rod for use in a nuclear reactor com-prising a plurality of pellets composed of nuclear material, a tubular cladding enclosing said pellets, means hermetically sealing said rod, said cladding having a clearance space with respect to the pellets, a plenum area within said rod, a tritium removal and storage device within said plenum, said removal and storage device comprising an inner core of zirconium alloy and an adherent outer layer of a nickel alloy bonded to the exposed surfaces of said core, said outer layer being above five percent by weight of said removal and storage device.

2 A fuel rod for use in a nuclear reactor comprising a plurality of pellets composed of nuclear material, a tubular cladding enclosing said pellets, means hermetically sealing said rod, said cladding having a clearance space with respect to the pellets, an upper plenum above said pellets, said plenum surrounding a retention device, a tritium removal and storage device within said retention device, said removal and storage device comprising an inner core of a material selected from the group consisting of zirconium and alloys of zirconium and an adherent outer layer of a material selected from the group consisting of nickel and alloys of nickel bonded to all exposed surfaces of said core, said outer layer being above five percent by weight of said removal and storage device.

3. The rod fuel of claim 2 wherein said retention device comprises a spring with at least one affixed end cap.

4. A nuclear reactor core comprising a plurality of fuel rods, at least one of said rods including nuclear material, a tubular cladding enclosing said nuclear material, means for hermetically sealing said rod, a plenum area within said rod, a tritium removal and storage device disposed within said plenum, said device comprising an inner core of a material selected from the group consisting of zirconium and alloys of zirconium and an adherent outer layer of a material selected from the group consisting of nickel and alloys of nickel, said outer layer being bonded to sub-stantially all exposed surfaces of said inner core and having a thickness of between 0.01 and 0.03 inch.

5. A nuclear fuel assembly for use in a nuclear reactor comprising a plurality of fuel rods, at least one of said rods including a plurality of pellets of nuclear material, a tubular cladding enclosing said pellets, means hermetically sealing said rod, said cladding having a clear-ance space with respect to said pellets, a plenum area within said rod, a tritium removal and storage device disposed within said plenum, said device having an inner core of a material selected from the group consisting of zirconium and alloys of zirconium and an adherent outer layer of a material selected from the group consisting of nickel and alloys of nickel bonded to a substantial portion of the exposed sur-faces of said inner core.

6. In a method of producing a nuclear fuel rod the improvement comprising placing a plurality of nuclear fuel pellets within a tubular cladding so as to provide a clearance space between said pellets and cladding and a plenum area within said cladding, placing a tritium removal and storage device within said plenum, said device having an inner core of a material selected from the group consis-ting of zirconium and alloys of zirconium and an outer layer of a material selected from the group consisting of nickel and alloys of nickel bonded to said core, said outer layer being above five percent by weight of said device, and hermetically sealing said pellets and device within said cladding.

7. The fuel rod of claim 2 wherein said outer layer is between eight and twelve percent by weight of said device.

8. A nuclear fuel rod comprising a plurality of pellets of nuclear material, a tubular cladding enclosing said pellets, means hermetically sealing said cladding, said clad-ding having a clearance space with respect to said pellets, a plenum area within said rod, a tritium removal and storage device disposed within said plenum, said device having an inner core of a material selected from the group consisting essentially of zirconium and alloys of zirconium and an adherent outer layer of a material selected from the group consisting essentially of nickel and alloys of nickel bonded to a substantial portion of the exposed surfaces of said inner core.

9. A core for a nuclear reactor, said core including a plurality of nuclear fuel assemblies, said assemblies includ-ing a plurality of fuel rods, at least one of said rods comprising a plurality of pellets of nuclear material, a tubular cladding enclosing said pellets, means hermetically sealing said rod, said cladding having a clearance space with respect to said pellets, a plenum area within said rod, a tritium removal and storage device disposed within said plenum, said device having an inner core of a material select-ed from the group consisting essentially of zirconium and alloys of zirconium and an adherent outer layer of a material selected from the group consisting essentially of nickel and alloys of nickel bonded to a substantial portion of the exposed surfaces of said inner core.