ML19312D769
| ML19312D769 | |
| Person / Time | |
|---|---|
| Site: | 07105507 |
| Issue date: | 02/13/1980 |
| From: | Ross D ENERGY, DEPT. OF |
| To: | Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 15473, NUDOCS 8003250314 | |
| Download: ML19312D769 (15) | |
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'.-l Department of Energy eir Washington, D.C. 20545 p
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Mr. C. E. MacDonald, Chief
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U.S. Nuclear Regulatory Comission Washington, D.C.
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Dear Mr. MacDonald:
- l For your information, enclosed are the following certificates and supporting technical documents for revision 7 to USA /5507/BLF (DOE-0R) for the HFIR Spent Fuel Shipping Cask:
a) 8 copies of the revised certificate of compliance b) 8 copies of the engineering evaluation c) 8 copies of the technical support to the SARP revisions The packaging is not expected to be used by licensees.
Please advise us of the results of your review.
Sincerely, e
Donald M. Ross, Chief Occupation 61 Safety Branch Operational and Environmental aey iv sion 3 Enclosures cc:
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U.S. OEPARTMENT OF ENERGY tit rt)
CERTIFICATE OF COMPLIANCE I'
- I For Radioactive Materiais Packages la. Certificate Numoer Ib. Revisen No.
Ic. Package identification No.
- 14. Page No.
Ie Total No. Pages.
3507 7
USA /5507/BLF (^0E-OR) 1 3
- 2. PREAMBLE 2a.
This certificate is issued to satisfy Sections 173.393a,173.394,173/l35, and 173.396 of the Deuertment of Transportation Hazardous Materiais Regulations (49 CFR 170-189).
Ob.
The pecxaging and contents described in item 5 below, meets the safety standards set fortn in Sucoert C of Titie 10, Code of Federal Reguiations, Part 71," Packaging of Radioactive Material fur Transport and Trar'socrtaten of Radioartive Matertal Under Cartain Conditens."
2c.
This certificate does not relieve the consignor from compliance with any requirement of the regulations of the US. Cepartment of Transportaten or other applicable regulatory agencies, including the governrnent of any country through or into which the package will be trensoorted.
- 3. This certificate is issued on the basis of a safety analysis repcrt of the package design or aco4scation-(1) Prepared by IName and addressl:
(2) Title and identification of reocrt or aponication:
(3) Cate:
Safety Analysis Report for Pack-November 1977 a.
Oak Ridge National Laboratory aging (SARP) - HFIR Spent Fuel Element Post Office Box X Shipping Cask Oak Ridge, TN 37830 Report No.: ORNL/ENG/T)f-12 (Continued on next page)
- 4. CONolTIoNS This certificate is conditional upon the fuefitting of the requ.rements of Subpart 0 of to CFR 71, as aponcao.e, and the conditiors specified in item 5 below.
- 5. Description of Packas.ng and Authortred Contents, Modei N., moor, Fissile Class. Otrier Canditions, and
References:
(a) Packaging:
(1) Model: EFIR* Spent Fuel Element Shipping Cask (2)
Description:
99Mo tar-Packaging for irradiated fuel elements and irradiated UCC primary get capsules. The inner cavity is 17 7/8 in. diameter x 32 in. deep. The internal configuration is determined by the type of fuel elements being shipped as follows:
(i) For HFIR fuel elements - a removable post consisting of aluminum wrapped with cadmium is centered in the cavity. The fuel aseembly (1 inner element and 1 outer element) is positioned in the cavity by a concentric fuel basket fabricated of stainless steel clad cadmium.
(ii) For ORR (Oak Ridge Research Reactor) type fuel elements tihich also in-clude BSR (Bulk Sbielding Reactor) fuel elements, for PRR (Puerto Rico Research Reactor) fuel elements,. and for CP-5 (Chicago Pile No. 5) fuel assemblies - the fuel elements are positioned in the cavity by a maga-99Mo tar-zine fabricated of stainless steel clad cadmium. The primary get capsules will be positioned in the same magazine. Adapters vill be utilizad to hold the capsules in the magazine. The center post is removed.
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I [ d et l 6b. Exo. ration Date:
S FCR THE U.S.OEPARTMENT OF ENERGY 7a. Address (of DOE isrurag OMce) 7b. Signature, Narne, and Titte (of COE Acoronag Offica,J US Department of Energ7 W.u.aimam Names' Post Office Bor E William H. Travis, Director Oak Ridge, TN 37830 Safety & Enviroamental Control Divisica 15G73 t
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Page 2. Certificate of Compliance, No. USA /5507/BLF (DOE-OR), Revision 7 3 (2)
(Continued) b.
Memorandum from R. M. Moser to William H. Travis, December 30, 1975.
c.
Memorandum from R. M. Moser to William H. Travis, October 27, 1976.
d.
Letter from M. E. Ramsey to J. A. Lenhard, July 26, 1979.
- 5. (Continued)
(iii) For PRNC (Puerto' Rico Nuclear Center) TRIGA/ FLIP type fuel assemblies - the fuel assemblies are positioned in the cavity by a steel, aluminum and elastomer basket. The center post is removed.
Shielding consists of a nominal 8 3/4 in thickness of lead between 1/2 inch thick stainless' steel inner and outer liners. Overall dimensions are 46 3/8 in diameter x 72 1/2 in high. Access to the cavity is through a top plug having 14 - 1 1/2 in, alloy steel studs. The gross weight is 23,000 lb.
(3) Drawings:
The cask an. modifications are described in Oak Ridge National Labora-tory Drawings: M-10191-EL-001-D through -006-D, -007-C, and -008-D-through 012-D.
The PRNC TRIGA/ FLIP fuel basket is described in Argonne National Labora-tory Drawing No.: W-0170-0090-DE-00.
(b) Contents:
(1) Type and Form of Material Solid, large quantity of radioactive materials, fissile encased in metal cladding as irradiated fuel elements OR as irradiated primary 9? o target capsules.
M (2) Maximum quantity of_ fissile material per package containing up to 9,500 g (1)' OneHFIRirradiatedfuelassembl{Uenrichmentupto93%.
of 235U as uranium oxide at a 23 Seventeen ORR-t irradiated fuel elements each 'containing (ii) up to 265 g of g. 50 as uranium-aluminum alloy at a 235U en-235U content of 4.505 g.
richment up to 93% for a maximum
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Page 3 - Certificate of Compliance, No. USA /5507/BLF (DCE-OR), Revision 7 (iii) Seventeen-irradiated PRR fuel elements each containing up to 235 192 g of U as uranium oxide as a 235U enrichment up to 20%
for a mav4 mum 2350 content of 3,264 g.
(iv) Seventeen irradiated ANL-CP-5 fuel assemblies each containing up to 170 grams of 235 235U U as uranium-aluminum alloy at a 235 enrichment up to 93% for a maximum U content of 2,890 grams.
(v) Forty-eight PRNC TRIGA/ FLIP irradiated fuel assemblies cooled y
for at least 30 days, each containing up to 125 grams 35U as 235U enrichment of auranium-zirconiumhgUcontentof6,000 grams.
dride mixture, with a 70%, for a maximum 23 99 (vi) Four irradiated primary Mo target capsules, cooled for eight hours after an irradiation period of less than 30 days, each 235U at 93% 235 to 20 grams of containingyUcontentof80 grams.
U enrichment, for a maximum The heat load is 1 3.75 kw.
(c) Fissile Class III O
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ORO Review of Prepesed A=endment for use of EFIR Spent Fuel Element Cask William A. Pryor This review is made of the letter dated July 26, 1979, from M. E. Ra=sey to J. A. Lenhard, subject, " Proposed Amendment to Certificate of Compli-ance No. 5507, HFIR Spent Fuel Element Cask".
The proposed use of the EFIR spent fuel element cask is acceptable under DOEM-0529 and 5201 based upon the following:
1.
There are no modifications to the cask or to the magazine for holding 99 o target capsules.
the UCC M
2.
Cask leading and unloading procedures are unchanged.
3.
The total fissile =aterial (assuming no burnup) is 80 g of 235U at 93% 235U enrich =ent in the four capsules versus a - vi=um of h.5 kg of 235U at 93% 235U enrich =ent for ORR fuel elements or 6 kg of 235u at 70% 235U enrich =ent for PRNC TRIGA/ FLIP.
For nuclear criticality safety, this loading is Fissile Class I.
4.
The total fission product heat load for four capsules of 1200 vt is considerably less than the approved 3200 vt for the HFIR spent fuel element.
5 External radiation levels for the cask and exclusive-use vehicle vill be measured and vill be within DOT li=its prior to release from ORNL.
6.
The leak testing of the espsules prior to irradiation and the maxi-
- num calculated internal teaperature of the fuel elements in the cask along with the maintenance of integrity of the silicone rubber gasket of the cask as indicated in basic-SARP, appears satisfactory to pre vent release of radioactive gasas.
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r-2 Conclusion Approval of the HFIR spent fuel element cask for shipment of these cap-sules is recommended.
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Po$T OFFICE Box X OAK RIDGE, TENNESSEE 37830 July 26', 1979 Department of Energy, Oak Ridge Operations Attention: Mr. J. A. Lenhard, Assistant Manager for Energy Research and Development Post Office Box E Oak Ridge, Tennessee 37830 Gentlemen:
Proposed A=endment to Certificate of Co=pliance No. 5507, HFIR Spent-Fuel-Element Cask We propose to use the HFIR Spent-Fuel-Element Shipping Cask for the ship-ment in August, 1979, of one irradiated primary 9'Mo target capsule of the i'CC-Tuxedo design described in the attached Patent No. 3,940,318.
This shipment is a part of the irradiation services for UCC authori:ed in the June 13, 1979, letter from Haythorne to Voth. We quest that an amendment to Certificate of Compliance No. 5507 be issued to permit the shipment of a maximum of four of the 99Mo target capsules of the UCC-Tuxedo
-design or a similar number of generic target capsules which are designed for an internal pressure of 900 psi and according to the ASME code for Class B nuclear vessels at 340*C.
Each capsule will contain a maximum 235U and 3000 Ci df 99 of 20 g of 93:
Mo plus other associated fission products at reactor discharge.
The target capsule will be irradiated in the ORR for the required pro-duction period, which will be less than 30 days. A U-shaped strap over the tubing fitting and tack welded to the end of the capsule may be added if deemed necessary for the ORR irradiation. Shipment of the target capsule from ORNL will be made after a decay period of at least eight hours, at which time the maximum heat generation rate in one capsule from fission product radioisotope decay is 300 W,..
The Safety Analysis Report for Packaging: The ORNL HFIR Spent-Fuel-Element Shipping Cask, ORNL/ENG-TM-12 states that a heat generation rate of 3200 W from a spent HFIR fuel element is acceptable under normal and accident c$nditions of transport.
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DOE, Mr. J. A. Lenhard 2
Shielding calculations were made by UCC-Tuxedo to define the radiation field outside the HFIR Spent-Fuel-Element Shipping Cask when loaded with 39 one Mo target capsule which has been irradiated for one week and which contains 3000 Ci of 99Mo at reactor discharge. The SDC shielc'ing code with updated attenuation values. from NBS-29 was used. The results of the calculations are tabulated below:
Radiation External to the HFIR Cask When Containing One >>Mo Target Capsule Decay Time (hr)
Dose Rate (millirem /hr)
Cask-Surface 3' from Cask 6i_' from Truck 8
363 43 8.8 24 132 16 3.2 32 116 14 2.8 40 107 13 2.6 These values are within the limits of DOT Hazardous Materials Regulations, 173.393(j), for exclusive-use shipments.
In the case of shipments of as many as four target' capsules, appropriate decay periods will be used to insure that radiation levels are within the 1Lnits imposed by the appli-cable shipping regulations. Since the'tsrgets will be loaded into the cask in the ORR canal, the cask can be he3.d in the canal so that no undue radiation exposure to operating personnel vill occur.
The quantities of important fission producta associated with the maximum 99 quantity of 3o in one target capsule irradiated for seven days or thirty days and af ter eight hour., decay are tabulated below. The 7-day period would be the optimum length for 99Mo production; however, reactor operating requirements or other considerations could increase the time in reactor, and 30 days was chosen as the maximum irradiation time.
Fission Product Activities in One Target Capsule After 8 Hours Decay, Ci 7-Day 30-Day Radionuclide T 1/2 Irradiation' Irradiation
~99Mo 67 h 2670 3221
'131I 8.05 d 802 1637 132I 2.3 h 1947 1947 133I 20.8.h 3114 3126 1357 -
- 6. 7 h' 1600 1600 133mXe 2.3 d 78 88
-133 Xe 5.36 d 2127 3473 135m 15 m 501.
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Fission Product Activities in One Target Capsule After 8 Hours Decay, Ci (Contd) 7-Day 30-Day Radionuclide T 1/2 Irradiation Irradiation 135Xe L9.2 h 1236 1236 85Kr
'10.3 y 0.3 1.28 87Kr 78 m 21 21 88Kr 2.8 h 290 290 89Sr 51 d 263 969 90Sr 28 y 2
8.6 93Zr 64 d 276 1051 95Nb 3.5 d 20 27 IU3Ru 39.4 d 211 748 106Ru 368 d 3
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The Safety Analysis Report' prepared by UCC-Tuxedo estimates the fractional release of noble gases from the UO2 film to be 0.01 at 700*C and 0.03 at 900'C. The target capsule is sealed with.1 atm of He gas inside, amounting to 12 millimoles of He.
If all of the noble gases are assumed to be released to the free space of the capsule,.this would amount to an addi-tional 0.64 m1111 moles of gas to the existing 12 millimoles of He or an increase of 0.053 atm over the original 1 atm pressure at room temperature.
In a supplemental letter to'their Safety Analysis Report for the reactor.
irradiation of the target, UCC-Tuxedo stated that a very extensive litera-ture exists on the release of volatiles from uranium oxide. They further stated that these references have been supplemented by measurements made on plated uranium oxide capsules. EIt has i'en found in agreement with the literature that the fractional release noble gases is less than 2.5% of the inventory and of iodine is 1%.
-ost of the iodine release is due to-fission product recoil, and wi plateout only about 0.5% of the iodine is in the gas phase. For th< two categories of fission products of safety significance during transport, iodine and noble gases, the quantities available in the gas phase of one target capsule at earliest shipping time (8 hrs decay) are as follows:
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DOE, Mr. J. A.'Lenhard 4
Fission Product Activities in the Gas Phase of One Target Capsule After 8 Hours Decay, Ci 7-Day 30-Day Irradiation
_ Irradiation 131I 0.8 1.6 1327-1,9
-1,9 1331 3,1 3,1 135I 1.6 1.6 Total Iodine 7.4 8.2 133mXe 1.9 2.2 133Xe 53.1 86.8 135m7e 12.5 12.5 1357e 30.9 30.9 Total Xenon 98.4 132.4 85Kr 87Kr-0.5 0.5 88Kr 7.3 7.3 Total Krypton 7.8 7.8 This capsule has been shown to be leak-free during several hundred irradiations in the UCC-Tuxedo reactor. Capsules pisted with uranium oxide have been temperature cycled between room temperature and 500*C-and-vibrated with ~1ess than 1% of the film appearing as loose powder granules. Each sealed capsule is subjected to a helium leak detector test and must conform to a specified maximum permissible leak of 2 x 10-8 3
sed em /sec (maximum sensitivity of the helium leak detector).
The integrity of the. stainless steel seal plug and welds has been verified in tests of up to 250-hour duration at 300*C and in short duration tests at-500*C. _The capsule is designed-for a maximum internal working pressure of 900 psi (ASHE code for Class B nuclear vessels at 340*C).
The HFIR' Spent-Fuel-Element Shipping Cask has been shown to protect transported fuel. elements from stresses involved ~1n fire and impact conditions (CRNL/ENG/TM-12). The_ calculated maximum temperature in a H71R fuel element in the cask sitting in direct sunlight and in 130*F ambient air is 213*C, During an experimental temperature measurement using a RFIR element having a total decay heat output of 3.2 kW, the maximum inner temperature of the cask reached 92*C after 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />. The maximum calculated temperature-in a HFIR fuel element'during a standard fire is 302*C. _ A comparison of 'the maximum heat generation rate in four
DOE, Mr. J. A.-Lenhard 5
99Mo' target capsules (1200 Wf) to the heat generation rate of the HFIR fuel element (3200 W ) leads to the conclusion that the maximum temperature MotargetcIpsulewillbewellwithintestedlimits.
of the 39 The silicone rubber gasket of the HFIR Fuel-Element Cask has been shown in the SARP to maintain its integrity at anticipated normal and unusual temperature con-dicions.. If a leak developed in the target capsule during transport, a portion of the iodine activities and practically all of the noble gas activities would diffuse into the cask cavity and would be retained by the gasketed closure with no increase in external radiation. In addition, a substantial fraction of the iodine isotopes would plate out on the cooler internal cask surfaces.
The proposed shipments of 99Mo target capsules described above have been reviewed and approved by the ORNL Transportation Committee as stated in the attached memorandum from E. M. King, dated July 17, 1979. We request your early review and issuance of an appropriate a=endment to Certificate of Compliance No. 5507 so that the first shipment of an irradiated 99Mo target capsule ccn be made in August, 1979.
Very truly yours, y i. 6,!:*,.,.us M. E. Ramsey Assistant Dire tor for Services MER:EL:es Enclosures cc/ enc:
J. A. Cox (4)
J. H. Evans R. F. Hibbs C. C. Hopkins E. M. King H. Postma W. R. Ragland W. E. Terry File - RC
United States Patent ti91 tiil 3,940,318 Arino et al.
(45) Feb.24,1976
[541 PREPARATION OF A PRIMARY TARGET FOR THE PRODUCTION OF FISSION Primary Examincr-Leland A. Sebastlari PRODUCTS IN A NUCLEAR REACTOR Attorn'y. A car.or Firm-William Ray.nond Moran (75) Inventors: Hirofumi Arino, New Windsor, N.Y.; Frank J. Cosolito, Ringwood;
[57]
ABSTRACT Kenneth D. George, Boonton, both A primary target for the production of fission products of N.J.; Alfred K. Thornton, New in a nuclear reactor, such as uranium or plutonium fis.
Hampton, N.Y.
sion products,is comprised of an enclosed, cylindrical
.[73) Assignce: Unica Carbide Corporation, New vessel, preferably comprised of stainless steel, having a York, N.Y.
thm, continuous, untform layer of fissionable material, integrally bonded to its inner walls and a port permit.
(22) Filed:
Nov. 6,1973 ting access to the interior of the vessel. A process is
[21] Appl. No.: 413,319 also provided for deposising uranium material on to the inner walls of the vessel. Upon irradiation of the Related U.S. Application Data target with neutrons from a nuclear reactor, radioac.
[62] Dives.ua of Ser. No. 100.918. Dec. 23,1970.
tive fission products, such as molybdenum.99. are formed, and thereafter separated from the target by (52)
U.S. C L....-
. 204/l.5; 29/458; 176/14; the introduction of an acidic solution through the port 176/16; 423/2; 424/1 to dissolve the irradiated inner layer. The irradiation
[511 Int. C1 B0lk 3/00; C0lg 43i00 and dissolution are thus effected in the sarne vessel
[53) Field of Search............. 204/1.5; 29/458 without the necessity of transferring the fissionable material and fission products to a separate cherr.ical (56]
References Cited reactor. Subsequently, the desired isotopes are ex.
UNfTED STATES PATENTS tracted and purified. Molybdenum.99 decays to tech.
3.243.352 3/1966 DowJouresques..
..-... 174/47 X netium.99m which is a valuabic rnedical diagnostic 3.318.695 5/1967 Goslee et at
.176/67 X radioisotope.
3.373.M 3 3/1948 Hurley at aL; 20all.$ X 3.691.087 9/1972 Hurley et al.
.....- 204/t.3 x Ca
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. U.S. Patent Feb.24,1976
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2 uranium target. When molybdenum is irradiated in a PREPARATION OF A PRIMARY TARGET FOR reactor,"Mo with a high degree of radionuclide punty THE PRODUCTION OF FISSION PRODUCTS IN A is obtained by the (n, y) reaction. Furthermore, the NULLEAR REACTOR chemical processing of the irradiated target is simple.
5 This method is currently widely used by radiopharma.
This is a division of aprlication Ser. No. 100,913, ceutical manufacturers.
filed Dec. 23,1970.
However when the Mo target is irradiated in the This invention relate, to a primary target useful for -
reactor, m an extremely small portion of the Mo is the production of fission products in a nuclear reactor, convertu.o radioactive "Mo by the (n. 7) reaction.
In one aspect, this invention relates to a primary target to Therefore, the specific activity of "Mo, the ratio of useful for the production of radioactive fission prod.
"Mo activity to the total weight of elemental Mo, is ucts such as "Mo in high concentrations and with a small. Moreover, the active absorption sites on alumina
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high degree of purity. In an additional spect, this in-are virtually consumed by inactive Mo, thus requiring a vention is directed to a primary target having a thin larger adsorption capacity to load a high "Mo activity.
continuous, uniform layer of a fissionable material,15 The specific activity of (n. y) "Mo is proportional to integrally bonded to its inner walls and a port permit-the neutron flux available in the reactor. Although ting access to the interior of the vessel. In another several investigations have also been reported on ad-aspect, the invention is directed to a primary target sorbents with a high adsorption capacity for Mo, the which serves as the same vessel for irradiation and vast amounts of inactive molybdenum in this case ad-chemical dissolution. In a further aspect, the invention 20 versely limits the activity of a Mc" generator.
is directed to a process for depositing uranium material Hence, for optimum use "Mo must be of an excep-on the inner walls of the vessel.
tiona!!y high order of purity, and it should hase a rela.
Recent medical investigation has shown that Mc"is tively high specific activity. As indicated above, one of an extremely useful tool for diagnosis. High purity the ways that has been studied for producing "Mo has Mc* is used primarily as a radioisotope in a variety of :S been by irradiating uranium in a nuclear reactor. How-medical research and diagnosis. It is well suited for ever, one of the problems that is encountered when liver, lung, blood pool and tumor scanning, and is pre.
producing "Mo in this manner is that more than 50 ferred over other radioactive isotopes because of its elements and more than 110 radioactive isotopes are short half. life which results in reduced exposure of the formed by nuclear fission. Therefore, the recovery of organs to radiation. In addition to medical uses Mc" 30 any single radioactive species from such a mixture can can also be employed in industrial apphcations, such as be a formidable task. Methods that have been em-in the measurement of flow rates, process control, ra.
played heretofore for recovering "Mo from irradiated diometric chemistry, and the like. Since the radioiso-uranium provide a yield product which often is not tope sought to be used has such a short half-life, it is sufficiently pure for medical use because it contains common practice to ship the users of the isotope the 35 significant amounts of radioactive iodine and ruthe-parent element;in this case radioactive molybdenum-nium.
- 99. The user then extracts the technetium from the More recent attempts to prepare high specific activ.
molybdenum-99 as his needs require.
ity readioactive materials by irradiation of uranium is For medical diagnosis, radioactive Mc" is usually to prepare a target comprised of uranium as an alloy in
' obtained from a generator containing the parent ele 40 a sandwich plate of aluminum.Through irradiation of ment. The generator is prepared by adsorbing "Mo on this sandwich, fission products containing the desired a column such as alumina. The column is then sent to radioactive materials are generated. Through subse-a hospital where the physician washes the column and quent chemical dissolution and separations, the desired obtains the technetium-99m solution which is then radioactive products are obtained in a relatively pure administered to a patient orally or by intravenous injec 45 form.
tion.The technetium-99m localizes in brain,Iung.hver, However, the use of an aluminum sandwich plate as spleen, bone, and the like depending on the prepara-a matrix for the uranium presents serious disadvantages tion of the Me" solution. The exact location can then in the production technique. The need to dissolve the be detected by typical scanning. This technique which alumine n matrix in order to obtain the uranium re-does not require an operation for diagnosis has become 50 quires a considerable period of time, for example, sev.
very popular in recent years and hence, the need for eral hours of the production process. During this time the Mc" generator is rapidly expanding.
the radioactive materials are decaying and therefore The first Me" generator was developed at Brook-final product is being lost. Moreover, the presence of haven National 1.aboratory. Uranium was irradiated in dissolved aluminum in the solution further complicates a reactor and the "Mo that was produced by nuclear 55 the separation steps and renders it difficult to obtain j
fission process was separated by alumina chromatogra-pure products. The relatively high volume of solution i
phy. The purified "Mo was again adsorbed on an alu-necded for the dissolution of the Isrge mass of alumi.
mina column from an acidic medium and the "I'c"in num results in corresponding large volumes of radioac.
the column was recovered by dilute hyorochloric actd.
tise waste solution wnich is difficult and expensive to Thn method, however, is currently not used for medi. 60 dispose of. Hence this methed requires a strict control cine because radionuclide purity of Me" solution from of varicus parameters such as temperature and concen-the Brookhaven column is not considered high enough tratinn and hence is slow and largely unsatisfactory-for medical use. It contained significant amounts of Moreeser, the purity of the radioisotope molybdenum.
radioisotope impunnes such as '*'Ru. iodine isotopes.
99 obtained from this method is not always satisfactory and the like.
65 for the current medical market.
More recently, as disclosed m U S. Pat. No.
According!v. it is an object of this invention to pro-3,382.152, a medical re" generator was des eloped by vide a novel primary target useN for the production of 1
using reactor arradiated molybdenum in place of the fission products in a nuclear reactor. Another object of un A 1}@
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4 this inventiosi is to provide a primary target useful for Both the bottom and top of the primary target are the production of molybdenum.99 in high yields and sealed with stainless steel closures which are welded in
' having a high specific activity. A further object of this place. The top c!osure is fitted with a port which can be inventmn is to provide a target which is convenient to sealed to provide an entirely closed system during es.
handle and from which the radioisotopes can readdy be $ posure to the neutron source as well as at various stages recovered. A sull further object is to provide a process during separation and recovery of the desired isotopes.
' for the preparation of a primary target.These and other The bottom and top closures, including the port and objects will readily become apparent to those skilled in welded seals, must be at least the sarne integnty as the the light of the teachings herein set forth, ra!!s of the container in order to ensure a safe opera.
In one aspect, the present :nvention is directed to a 10 (ion dunng radiation and subsequent proccuing.
primary target for the production of fission products, in a preferred embodiment the pnmary target is fab-such as uranium or plutonium,in a nuclear reactor and ricated from annealed, seamless, stainless steel tubing the process for its preparation. The primary target is approsimately 18 inches in length and having an outer comprised of an enclosed, cylindrical, stainless steel diameter from I to 2 inches ant,a wall thickness of vessel, preferably having a thite, continuous, uniform 15 from about 0.03 to about 0.10 inches. The, top ts layer of a finionable material, integrally bonded to its equipped with a port permitting access to the mterior inner walls and a port permitting access to the interior of the veuel. The port is composed entirely of metal, of the veuel. The primary target of this invention re.
preferably stainlesa steel, and must be capable of with-standmg the pressures and temperatures created during tains the advantage of the aluminum sandwich method, 20 exposure of the primary target to neutrons. It has been namely, good heat transfer, but avoids its disadvan.
tages. The fissionable material is disposed as a thin observed that,temperat, res of up to about 300'C. are layer, adherent to the inner surface of the cylindrical generated dunng irradiation. The primary target should vessel.The small thickness of the layer which may be of be capable of withstanding temperatures of at least the order of one thousandth of an inch thick, and its about 30m for,at least one hour.
intimate contact with the veuel results in good heat "
.As hereinafter mdicated, the primary target will co,n.
transfer from the deposit to the coolant that is in
- '" " predetermined amount of fissionable matenal contact with the exterior surface of the venel depou.ted on its inner wall. In practice, for an 18 mch pnmary target.of I inch diameter, a has been observed Through the use of the cylindrical venet the irradi-can be produced from a cy@lindncal uram,adioactbny that a maximum d aW 10 urks d r ated material has its surface esposed for prompt and um coating afrective dissolution in the post irradiation phase of the which is 13 inches in !
process. The dinclution sofution can be introduced appr aimately 20 mdl ength and has a thickness of igrams of uranium per square through the port in the volume needed to dissolve the centimeter. Such depesits, we,gh from about 7 to about i
irradiated material.Through suitable choice of solvent grarns. anium neues up to abt M mW.
the depmit can be dissolved with little affect to the E " F " " * "'"" * ***'
- *'I venel itself. The far3e surface area etPosed results in targets have also been employed omd are calculated to rapid Jiuolution, which may be,only a few minutes, yield approximately 25,000 curies of radioactivity and therefore conserves procesung time and avoids based upon a 15 inch coating of the length of the inner lou of product through radioactive decay. The result
- wall of a 1 inch outer diameter stainless steel tube.
ing solution contains an insignificant quantity of dis 40 Such deposits weigh from about 18 to about 23 grams.
solved venet, which considerably simplifies the subse*
In another aspect, the primary target serves as the same quent chemical proccumg and enables sophisticated vessel for irradiation and chemical dissolution of the separative methods to be used that result in highly-pure
- uranium, products.
In a further embcdiment, the present Invention is With reference to the drawing. the single FIGURE 4s directed to a process for the deposition of the uranium shows a perspective view of the primary target 10 material onto the inner walls of the primary target. In which is a cylindrical vessel closed at the bottom and practice, the uranium can b1 deposited as uranium top. The top is equipped with a port 12 permitting metal, uranium oxide,or in any other form as long as an access to the interior of the vessel The cutaway sec.
adherent bond is formed with the inner walls of the tions shown in the drawing permit observation i f the 30 target. It has been observed that uranium can be elec-layer of fissionable material 14 which is deposited on, trolytically deposited from an aqueous electrolytic bath and integrally bonded to the interior wall of the target.
containing uranium salts enriched in the fissionable The layer extends from a predetermined length from isotope and chemical additives. The thickness of the l
points 16 to la along the inner wall of the venei.
deposit can be up to at least 50 milligrams per square l
In practice, the target material can be comprised of 35 centimeter. When subject to $00*C. for one hour or most any metal or alloy to which the fissiorable mate-more in a nitrogen atmosphere,it was observed that the rial can be bonded and subsequently chemically re-deposit maintained its adhesiort and integrity even moved by acid treatment with litrie or no attack on the when subjected to severe vibrations.
target vessel itself. For example, the primary target can As previously indicated, the preferred vevel material be fabricated of stainless steel, nickel, nickel alloys, 6o is seamless, stainleu steel tubing in lengths of up to 18 risconium, zine coated alummum, and the like. Stain-inches and diameters of from one to two inches. The less steel, has been successfully ernployed as the target tube should be smooth, free from any flaws and havs, a material and is the preferred choice. The walls of the thickness of from armut 0 03 to about 0.10 inches.
target venet must he smooth, free from any flaws. This Pnor to electroderosition, the tube must be thoroughly
- ensures an intimate contact of the fissionable matenal 65 cleaned with acid in order to ensure good bondmg of with the inner walls of the vessel and optimum heat the uranium to the etainless steet. This can be onve.
trartsfer to the coolant during espmure of the pnmary niently effec *ed by stoppenng one end of the titbe with
- target to radiatum.
a rubber stopper, filling the tube with 23 per cent sulfb-h O
6 A
1 ) A X i[.,
5
~
3,940,318 5
6 ric acid at about 95'C., and allowing it to stand for at minutes. This removes the uranium deposit at the ends least 20 minutes. A second acid treatment may be em.
of the tube, leaving approximately 15 inches of plated ployed if desired.
area. After washing and drying, the target is heated in it has been observed that the preferential method to an oven at 500*C. for one hour under an inert atmo-ensure opttmum bonding of the uranium to the tube is 3 sphere, such as nitrogen. The target is then sealed at to first deposit a thin layer on to the inner walls of the both ends by welding a bottom and top closure to the tube. This permits observation to assure that the elec-
- tube, troplated uranium is adhering to the tube and also After the primary target containing the uranium de-serves to obtain a tare weight of uranium-235. In prac-posit has been sealed and carefully tested for leaks, it is tice, the layer deposited in this preplating stage, can be to then ready to be placed in the nuclear reactor. The either uranium 235 or uranium 238. However, the primary target is then exposed to radiation in accor.
remaining layer subsequently deposited is comprised of dance with known techniques and employing the usual uranium 235, safeguards and controls. After exposure for approxi-The electrolytic bath employed both in the preplating mately 100 to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> the target is removed from the and plating steps in an aqueous solution containing at 33 reactor and transferred to a facility for separation and least one uranyl compound in which the uranium has processing of the fission products.
been enrichedin uranium-235. A wide variety of uranyl Although the electrodeposition process of this inven.
compounds can be employed as long as they are soluble tion is particularly pplicable to the preparation of m water. Illustrauve ura.iyl compounds include, among primary targets having cylindrical configurations,it can others, uranyl chloride, uranyl formate, uranyl nitrate, 20 also be employed to electrodeposit uranium onto met-uranyl oxalate, uranyl sulfate, and the like. The pH of als of other shapes. For example, flat plates of stainless the bath is adjusted to a pH of from about 4 to about 9, steel or other metals can be plated with uranium by this and more preferable to a pH of 7.2.
process.
The concentration of the uranyl compound in solu-Other methods can also te employed, if desired, to tion is not necessarily critical and can vary depending 25 deposit the fissionable material,i.e., uranium or pluto-upon the time employed in electroplating. In practice, nium, onto the inner walls of the vessel. For example, however, optimum results are obtained from an elec-the metal can be sputtered, ion plated, or evaporated, trolytic bath comprised of 0.042 hlolar UO (NO ) 6 onto metal surfaces by known techniques.
3 H 0 and 0.125 hlolar (NH.) C,0.. H O which has The following examples are illustrative; 2
been adjusted to a pH of 7.2.
30 After the tube has been treated with acid,it is rinsed EXAhlPLE I with distilled water and placed,in a vertical position in Preparation of Primary Target a water bath. A carbon anode is then mwrted into the tube, centered, and fixed in place. A cathode electrical An 18-inch long tube of I inch outer diameter and connector is placed over the outside diameter of the 3s constructed of annealed, seamless No. 304 stainless target and both anode and cathode cannected to a steel tubing (hllL-T.8504A specification) was c! caned power source. Electrolyte is introduced into the tube in sulfuric acid solution and washed. Uranium enriched from the bottom and flows through the tube and out the to 93% in "U was electroplated over a 15. inch length top at a rate of from about 150 to about 180 milliliters inside the capsule in the form of a uniform thin film of per hour. The electrodeposition is conducted at a con. 40 uranium oxide. The electroplating was effected by first stant voltage of 1.5 volts and at a current within the preplating a thin-film or uranium onto the inner surface range of from about 0.3 to about 1.0 amperes. AI.
of the tube from an aqueous bath containing 0.042 though electrodeposition can be effected at tempera.
molar uranyl nitrate and 0.125 molar ammonium oxa-tures within the range of from about 50' to about late, the pH having adjusted to 7.2 with NH,OH. Elec-100*C., it is preferred to electrodeposit at the higher 45 trodeposition was effected for 60 minutes at a current temperature. In practice when the water jacket sur.
of 0.9 amperes,1.5 volts; and a temperature of 93' =
rounding the tube is maintained at 93*C. = 1*C., excel.
1*C. Thereafter, the cylinder was removed from the lent prepiating is obtained in one hour at a current of plating assembly, washed with water, dried, and 0.9 amperes. Upon completion of the preplating, the weighed. The final uranium deposit was made from a target is removed from the water bath, rinsed, dried 50 similar electrolytic be as that used in the preplating and weighed. The preplated target can be stored indefi.
step. The temperature employed was 93*2: 1*C and a nitely. The final plating of the uranium is done in a fixed voltage of 1.5 volts. The current was cycled by similar manner using the same apparatus employed in means of a clock mechanism starting with 0.3 amperes, the preplating step. Care must be taken in inserting and 0.6 amperes, then 0.9 amperes, then 0.3 amperes, etc.
removing the carbon anode so as not to damage the 55 every 15 minutes.' The electrolyte was circulated plated surface. When the bath temperature reaches through the cylinder at a flow rate of 200 milliliters per 93*C. = 1*C., electroplating is commenced at a current hour. The electrodeposition rate was approximately 1.2 of 0.3 amperes. In order to minimize bubble formation grams uranium oxide per hour. After about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> the and to achieve good bonding the current is cycled at 15 cylinder was removed from the platmg assembly, minute intervals starting with 0.3 amperes for 15 min. 60 washed and dried. The ends of the cylinder were utes, the current is increased to 0.6 amperes for the dipped in nitric acid to remove about 1 % inches of the nett 15 mmutes, and then to 0.9 amperes for 15 min-uranium deposit to grve an arbitranly selected length of utes. The cyc!c is then repeated starting at 0.3 amperes.
15 inches of uranium deposit in the tube. The resulting Electroplatmg is contmued for a period of approxi.
film thickness was 20 mg "U per em of tube surface, 8
mately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Thereafter, the target is removed from 65 for a total deposit of 7 gmSU. The total uranium mass the bath. washed and dried with forced hot air.
was determined gravimetrically. The plated tube was Each end of the tube is dipped in concentrated metic then baked at 500*C. in mtrogen. The adherence of the acid at 35*C. to a depth of about 1% inches for a few film was checked with a vibration test. The film re-I"D 9@ YM A
D *%
J R h
=
3,940,318 7
8 i
mained adherent despite temperature cycling between for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. Thereafter, the pnmary target was re.
room temperature and 500*C.; the latter being well rnoved to a hot cell facility, the swage. type fitting ahove ibt cipected radiation ternperature of 330*C.
opened and the primary target connected to a self-scal-Tute,nted with uranium have been temperature-ing entrance port. The uranium oxide was dmolved by cycled and vibrated with less than 1 percent of the film 5 introducing though the port a mixture of 15 cc of con-appearing as loose powder granules.
centrated H,SO.and 60 cc of 10 per cent H 0s. Any off 2
The two end caps of the tube were hellarced in place.
gases in the primary target were trapped in a liquid The swage. type fittings that comprise the capsule seal nitrogen cooled trap. The mixed fission product solu-and entry port are No. 316 S/S. The maximum allow-tion was passed through an "Mo adsorption column (I 10 able internal working pressure is 900 psi (ASME Code cm x 5 cm) containing 2 cc of silver.cnatedcharcoal for Class B nudear vessel at 340*C.).
(20-50 mesh) and 2 cc of charcoal. The column was The plated vessel was then filled to about one atmo-wa3hed with 60 cc of dilute suliunc acid and 60 cc of sphere of helium, scaled, and then leak. tested with a water successively. The "Mo retained in the colurnn mass. spectrometer type leak detector. The maximum was eluted with 20 cc of 0.2 M NaOH. The etuant was 15 permissible leak is 10-'sec/sec. The integrity of the passed through another punfication column that con-stainless steel seal plug (and the welds) has been veri-tains 2 cc of silver. coated. charcoal at the upper part of fled up to 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> at 300*C., and in short. duration the column and 2 cc of zirconium phosphate at the tests at 500*C., and also in the 214 hour0.00248 days <br />0.0594 hours <br />3.53836e-4 weeks <br />8.1427e-5 months <br /> Instrumented bottom. To the "Mo product solution thus obtained,5 Target Experiment during which the radiation monitor.-
cc of hcl was added to make the "Mo solution isotonic ing system indicated no primary capsule leakage.
saline solution. The solution contained about 100 cu.
ries of "Mo. A 0.1 cc portion of the "Mo-isotonic EXAMPLE !!
saline solution was added to a Woelm alumina column irradiation and Recovery of Molybdenum.99 (0.6 x 3 cm) and the column was washed with isotonic The reactor irradiation assembly employed was com-25 saline solution. The adsorption of "Mo was >99.999 prised of a sealed primary capsule containing uranium.
per cent. The "Tc-*, after it accumulated in the col-235 and enclosed within a close. fitting secondary con.
umn, was eluted with 10 cc of isotonic saline solution tainer. Heat generated in the primary was conducted and recovered >90 per cent. The "Mo content in the through the narrow gas. gap between it and the second.
"Tc*' cluant was in the order of 10-* percent. No other ary. Gas lines entenng the top and bottom of the sec-30 radioactive contaminants were detected either by a ondary allowed a helium atmosphere to be established GeLi or Nat (TI) crystal coupled with a gunma. ray within the container and a slow sweep of gas to be analyzer. The alumina breakthrough in the "Tc" elu-taken to monitoring equipment located on the reactor ant was very small, <! ppm. The total heavy metal bridge. Pressure, flowrate, and radioactivity of the gas content was <1 ppm.
were monitored. Exit gas was filtered before venting 35 Although the invention has been illustrated by the into the reactor-building exhaust duct via a solenoid-preceeding examples, it is not to be ccnstrued as being controlled shutorf valve. The secondary container was limited to the materials employed therein, but rather centered in a stringer tabe within the reactor core and the invention is directed to the generic area as herein-was cooled by primary water flowing in the annutus so before disclosed. Various modifications and embodi-fermed. The sasembly was designed to contain about 40 ments can be rnade without departing from the spirit 400 curies of "Mo at removal from the reactor, and scope thereof.
The primary target capsule, as prepared in Esample I What is claimed is:
above, was then placed in a secondary capsule which
- 1. A process fer the preparation cf a primary target was fabricated from No. 304 stainless steel sanitary for the production of fission products in a nuc! ear reac.
tubing. A lead weight was provided for ease in place-45 tor, said target being compIs.d of (a) an enclosed ment and for ensuring that the assembly would not float metallic cylindt; cal vessel scaled at its top and bottom, in water. Two gas lines (% and % inc. OD No. 304 S/S)
(b) means for permitting access to the interior of said were provided, one in the top cap and one near the vessel, and (c) a thin, continuous, uniform layer of a lower end 'of the capsule. These lines supplied the he-fissionable uranium material, integrally bonded to the lium gas which served as the heat. transfer medium 50 inner cylindrical wal:s of said vessel, said process com-between primary and secondary necessary for limiting prising the steps of:
the primary temperature to the design value of 330*C,
- a. electrodepositing a uraniurn material onto the The upper end-cap consisted of a stainless steel"CA-inner cylindrical walls of said vessel by electrolyz.
JON type VCO coupling TIG. welded to the capsule ing a syste n comprised of:
body. This coupling employed a silver. plated stainless 35
- i. an anode, I
steel O-ring for its seal. The O. ring was discarded after ii. said vessel as the cathode, use. A!! welds and penetrations in the secondary cap-iii. an aqueous electrolytic bath containing at least sule body were helium leak. checked.
one uranyl salt enriched in uranium 235 and l'
The seenndary capsule containing the primary target having a pH of from abemt 4 to about R said capsule was placed in a core stringer tube. This alumi. 60 electrolyzing being effected at a temperature of num (No. 60nl) tube provided the 0.25 inch gap from about 50* to about 100*C; and at a current nee 6d for the desired cooling water speed of 3.5 ft/sec which is periodically and sequentially cycled past the secondary capsule. Primary reactor cooling from about 0.3 to about 0.6 to about 0.9 am.
water with normal gravity flow was used. Meastwe-peres, and ments in a test stand showed that at least 3.9 ft/sec is 65
- b. sealing said top and bottom of said vessel.
obtained.
- 2. The process of claim I wherein said uranyl salt is The stnnger tube, containutg the secondary and pri-uranyl mtrate.
mary taryct, was then lowered into a nuclear reacter
- 3. The process of claim I wherein said uranyl salt is and irradrated at a neutron flux of 3 x 10'8 niem sec.
uranyl sulfate.
2 e
2