. * . 1 .. 1 - . . a + 1 . . . : - I OF ORNL P 1724 L .. - . Y 1 • : . . 1 4 . . . - . ... 50 156 163 11:: 1.6 MICROCOPY RESOLUTION TEST CHART NATIONAL BUREAU OF STANDARDS – 1963 * OLNU P-124 : CLASSIFICATION ( ! APPLICABLE) ORNI - AIC HOV 18 1955 Conf.651041-/ FACE. NO. 14 1 BOTTOM OF I FIRST LINE OF TEXT OR' CHAPTER TITLE · ORNL - AEC - OFFICIAL Oiicu * 4 - THE BRITISH CERAMIC SOCIETY MEETING ON NUCLEAR AND ENGINEERING CERAMICS | 25th - 27th October, 1965 Atomic Energy Research Establishment, Harwell, Berks., England . .. . . . * * 21 * "Properties and Prospects of Thoria-Base Nuclear Fuels"* LEFT MA MARGIN -------- RIGHT MARGIE ..... DES . By A.R. Olsen, D. A. Douglas, Y. Hirose, J. L. Scott, J. W. Ullmannt Metals and Ceramics Division, Oak Ridge National laboratory, Oak Ridge, Tennessee, USA . . MAXIM 1... 25 . . .... .... .. . .. CENTER LINE RKLEASED FOR ANNOUNCEMENT IN MUCLEAR SCIENCE ABSTRACTS . - - - - - - - LEGAL NOTICE This report was prepared as an account of Govorament sponsored work. Neither the United Stites, nor the Commission, por any person acting on beball of the Commission: A. Makes any warranty or representation, expresiod or innplied, with rospect to the accu- racy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, mothod, or procon disclosed in this report may art infringe privately owned rights; or B. Assumes any liabilities with rospect to the use of, or for damages robuuag from the use of any information, apparatus, method, or procoss disclosod in this roport, Ao usod in the above, "por son acting op beball of the Commisslon" Includes way on- ployon or contractor of the Commission, or employee of such contractor, to the extent that such employee or contractor of the Commission, or employce of such conti actor prepares disseminater, or provides ACCOGU to, any information pursuant to bio omyloyment or contract with the Commission, or dio omployment with such contractor, - - - - - ... - - - Bu ö .. --- --- ---- ---- 13 REMAINING LINOS ог түрі. - *Research sponsored by the U. 8. Atomic Energy Commission under contract with the Union Carbide Corporation. . men ; o ir is i n 2010). "";' emical Technology Division, .: :, ... · VEND TYPING 1 TL ORNI - AEC - Qucu * -- ORNL - AEC - OFFICIAL olan ........... u . 1 CLASSIFICATION Tonirani (IF APPLICABLE) . . .. > '-'. Cl. 19:SIFICATION "IF APPLICA) PAGI. NO. . COMEC - Orricial i. PROPERTIES AND PROSPECTS OF THORIA-BASE NUCLEAR FUELS . . : .A.. R. Olsen Y. Hirose. D. A. Douglas J. L. Scott. . . J. W. Ullnann .. ... 5 . . n ABSTRACT . . ... ... ... .. . .. 10 The fertile material, thoria, offers unique properties as a source oſ nuclear fuel. Reactor control problems are minimized and an extended irradiation life 18 possible as a result of the nearly constant fission- . ! able material from 2330 bred into ThO2-U02 solid solutions. The sol-gel process provides a low-cost method of reprocessing which can be combined with remote fabrication of high-quality fuel rods for economical fuel recycie. RIGHT MARGIN .!...DESIRED .MAXIMUM 1 . Irradiation tests with vibratorily-compacted thoria-base fuels at .. .im ... . . - - ... - linear heat ratings up to 1000 W/cm , and burnups in excess of 100,000 ma/tonne have shown that this material must be considered as a unique fuei and not just as a UO2 substitute. As a result of its more refractory character, thoria has been shown to be capable of delivering higher linear heat ratings than urania without central melting. High- burnus, irradiations have also demonstrated the retention of crystal structure, good dimensional stability, and reasonably low fission-gas release. 19 Altnough the chemically processed sol-gel ceramic fuels are almost . ? theoretically dense, the large surface area of fine particles in the si agglomerates and the current methods of production permit the adsorption " or entrapment of significant quantities of gas. Such gases have not ?: affected tests to date; but this represents an area where a significant - .. ..... .. . - . - - - . - - 1 .- 1 ---- ORKL - AEC - OFFICIAL CLASSIFICATION .:,:,liti APPLICAOLE): 1..ASSIFICATION llo APPLICABLET :ii. PONCE NO. " ..OOTTOM OF 7 ORNI - AEC - OFFICIAL ORNL-AEC -oiričiau ....... :: improvement can be made. Thoria is' not as subject to stoichiometry į OR CHAPTER TITLfi.. ::;. effects as urania, but the influence of the solid solution oxygen con- tent on thermal conductivity and microstructural changes requires Adaltional out-of-roactor as well &b Irradiation testo. · Thorium-base ceramic fuels as pressed and sintered bodies, as - vibratorily-compacted powders, and as spheroidai particles have a proven potential in the nuclear industry as the current interest in ThO2-U02 advanced converters attests. This potential combined with the possible economies of the gol-gel process and remote fabrication establishes a real need for additional characterization and improve- . FIGHT MARGINGI ment of a long-used basic ceramic material. i... . . m ore . mis-0ESI , ...DOSTRED 7.A1AXIMUR - . -..'..... 11 INTRODUCTION ... 25 Thoria has long been a basic compound used by ceramists in non- nuclear applications because of its nigh melting point, approximately 3300°C, and its stability. Thorium dioxide has a fluorite crystal structure which is stable up to the melting point. Ii does not absorb excess, oxygen to a measurable extent, and although it blackens on pro- longed healing above 1800°C in a vacuum, the loss of oxygen is not measurable by current chemical analysis or lattice parameter measure- ments. The nuclear potential of thoria derives from the fact that *thorium 18 abundant in the earth's crust and as a fertile material yields fissile 2330.' The utilizetion of thorium in power reactors as ThO2 or in any other form is, of course, largely dependent on economics. The economical potential of thorium utilization has been discussed by Lane ( 1) and - -. . - - - • - . - CC C.. - - - - - - - - - - - WEMAINING LINES ; OF TYPE .. ---- , cm - -- " - . ---- - - - ENO TYPING *• * o prin post with a ... .. - ORNL - AEC - OFFICIAL ... - ORNI - AEC - OFFICIAL --. CLASSIFICATION (IF APPLICAOLE). Simone ---- -- - - - - - - - - : CLASSIFICATION TIF APPLICABLES: - L'AGE. HO. .-. . . OANLAEC - OFFICIAL TEXT TI . Teli.. others. (2-3) In this paper we will present the results of Irradiation tests and complementary supporting research and development that is being done to establish the performanse characteristics of vibratorily. 1 compacted sol-gel derived ThO2-U02 fuels. The program includes a B comparison of such material with ThO2-U02 derived from other processes. . When these characteristics have been established the reactor engineers i will have a set of limitations within which they may design a reactor system that will yield the lowest overall power cost. ". . a _ mereno- * IRRADIATION TESTS :RIGHT MAUGIH . ---- MAXIMUI .. - . Irradiation testing of metal-clad bulk-oxide fuels of mixed ThO2 and UO2 was started at the Oak Ridge National Laboratory in 1961. The testing, which is a part of the Thorium Utilization Fuel Cycle Program, has concentrated on fuels produced by the sol-gel process and fabricated into rods by the use of vibratory compaction. . Most of the tests have been made with noninstrumented rods in the : Materials Test Reactor (MIR), Engineering Test Reactor (ETR), and the Chalk River National Research Experimental Reactor (NRX). Three tests were conducted in the Oak Ridge Research Reactor (ORR), one a trefoil *cluster in a pressurized-water Icop operating at 250°C and 750 psi, and the other two as instrumented rods in the Poolside Facility with 540 and ...9.700°C cladding Temperatures. : ** The initial phases of the program involved short stainless steel .5 tubes, while the last. two groups fabricated have used Zircaloy-2 'clado a l 3;ding and a design that was developed for the semiremote fabrication of 1 293u-containing rods in the Kiiorod Facility. (8,one group of rods .-. .-. . .. -... C - OFFICIAL i .. " ENI TYPING O ....... ........... .....CH CLASSIFICATION (IF APPLICABLE) : : CLASSIFICATION ALION (IF APPUCALLES ORN1- PAGE: NO. . 5 - . - А воттом оf nisT LINE OF TEXT TER TITLE O RNL-AEC - officia . .. OFFICIAL . si W --- - .: 10 C - . - . FION MARGIN -. -- - ....DESIRED MAY IPSUM - --- - currently under Irradiation is enriched with 2330 and was fabricated in thie facility: :: The sol-gel process for producing the fuel with a particle density: 1 : of 99% of theoretical has been described by Wymer and Coobs (6) in an early paper at this meeting and in previous reports. (9,10) Since then process has been inder, development during the time that these irrad- iations have been conducted, there have been improvements in the pro- dess and in the resulting Oxides. In most of these tests, the weight percent of 235U has been from 4 to 5% in order to reduce the time :''! required to obtain a significant exposure, even though the equilib- rium concentration of ilssile atoms in a time-averaged perturbed flux of 1.1 x 1024 neutrons cm"? secºl is approximately 1.7 at. % of heavy element. Postirradiation examinations Include, dimensional analysis for swelling or bowing, fission-gas release-measurements, gamma scanning, burnup determinations, x-ray diffraction analysis, and metallographic examinations. ? The Irradiation program to date 18 sumarized in Table 1.: The detailed results of the tests completed to date have been reported previously. (9:16) The results proved that the chemically produced ceramic has the basic requirements of a nuclear fuel and that it com . pares favorably with arc-fused material. The experiments also showed : *that vibratorily-compacted powder uel rods perform well when comparedi OF TYPE .. 5 with fuel rods containing pressed-and-sintered pellets at moderate a linear heat ratings up to 300 w/cm. Currently, we are investigating - . ?.) ---- ::: : : 10 0 . -' TAINING LINES . . r 21; 50' . ORNI - AEC -O .. ORNL - AEC - OFFICIAL I .. . . - CLASSIFICATION : tiili (IF APPLICABLE) ! i . - .- - -. th EXI A . ..'nin cimet. .. ... .min....... ... . . .. ni !.. .. Table 1. Summary of Thorium Fuel Cycle Program Irradiations of Powder Packed Rods . mana - mana Lalasar No. of Rodo.. Desi patioa Peak Baron Type of Ozido : . Duosity (7 cl theoretical) Heat Pad Rod Dimanais Dimensions (cm) en op wil hud/tona. Status (w/cm) metal) · Objective : -----..... ... .. --- : YTRI 7 Arcfused aad Sol-Gol E 86 to 87 27 0.08 0.06 390 15,000 to 100.000 .. MTRD 2 Sol-Gal s a 88 to 89 6 camined. Provide base-line data to - din mactor we in comparing sol. god mad arcofased oxide In reactor . Obtain higher heat ratiag by increasing caricho i... 5 7 in 0.8 je 0.06 600 100,000 : . . * VTR-II 6 Sol Gol 35 neat. ... 86 to 89 : 30 820 0 . 100.000 :. . kn reactor Compare oxide caiciaing .5.16 ate nephen md higher . beat ratings obtained . by increasing diameter Being es- Same us for MTR-II ....116 ADUL.sasukin .PRL saia Sal God 352 NOI.LVOL:118SV 7. * 1.1 0:06 960 220 32,000 .8 0.06 160 16,000 zamland Sol Gel A . wad B Sol-Gel C Provide base-line dati MRTI 4 57 ::;;. . med en 240 08 0.06 5,000 .. NRX- 11 6 · based Sol-Gel S 48 to 89 - 99 _28. 270 Exsniand Study effect of increased : : .length. . ... Examined Study effect of tocreased ......izvirni leagth . .. Examined Study Tho, Pao, and . ..:: lower packed density NRED :: .. sol. Gal 74 to 76² 29. 08 23,000 . . 0.06. 260 22,000 .. .......... . sorin . ... Home. ......... "Y in.. .. a -:. : IT- . - - - . 1 . 0 · TVI21330-OV - INIO : W 1310- 33V - INUO . . Table 1 (continued) - . Linens No. of Type of . : . Density (% of theoretic " Fuel Rod Dimensions (cm) Beat Ratlag Put Beraup (1d/tomno Status jestima Designation Bado Longtha OD Vell ORR Loop 3 Sol-Gal 26 84 to 5 35 1.2 aoe 500 2,100 Brenined ORR Praloide 2 SolGel D Brentand 18 . - - 6 BNL Sol.. 48 1.3 009 630 a reactor 30,000 to 30,000 to 100.000 Study be pressurised water at 2609C eod 1750 pri . Neuwettuctive thermal coadactivity ning . . central themocouple la Nek at 318 poi; 540 mod 705°C Study effect a remote fabrication and oxide cilatatag . Study Tho, Weaket auto terid with godaally Lacrestag buat rating man provide pro tactatur l iesiog product antusial for chantal mucoalag 7 STR-OI. 88 Sol-Ged .48 1.3 0.09 770 10,000 to 70.000 In reactor . - THO, "Au rods were ded with type 304 stainless steel except for groupe ETR-II.mod - II and two of the three ORR Loop specimens; these were de la Zis Sep packed . ORNL - AEC - OFFICIAL ORNL - AEC - OFFICIAL (1.115.CRT10;ATION lor nisi TAULCI ·linit 10. EXT." ANL - MC -0181CIAL i..! in the effects of extended burnup, higher heat ratings, and processing me variables such as semiremote fabrication and sol-gel calcining atmo-. - .com . . spheres. . i m on.com ------ ! RIGHT MANGINI - DESMED TAXIIUMI .... Hfects of Extended Burnup in, The effects of extended burnup on mi"..ostructural changes, swell- ing, and fission-gas-release rates are best seen in the data from the MI'R-I group and a group of rods containing pellets that were irradiated i for power reactor fuel reprocessing studies. Some of the pertinent .. data are shown in 'Table 2. . The fission-gas-release rates have been : moderate. There is no evidence of a saturation phenomenon such as a sudden increase in gas release or any significant changes in the fuel rod dimensions. The gas-release rates appear to correlate best with the exposure time. This cen be a function of the time at temperature and/or the number of temperature cycles. During the ANS meeting in Gatlinburg, June 21-25, 1965, W. B. Lewis stated that recent in-reactor tests on UO2 fuels showed distinct steps in gas release with each thermal cycle. Although we have not checked this proposal against the detailed history of these experiments, the effect is not surprising if one accepts the gas-trapping theory proposed by Carroll and Sisman (17) and considers the structural changes under Irradiation and thermal cycling. Macroscopically, the vibratorily-compacted and the pellet fuels appear the same after irrediation. The fuel is severely frag- mented with interlocking pieces separated by both longitudinal and transverse cracks. If the gases are trapped at grain boundaries and. dislocation sites during irradiation, the fragmentation that occurs on thermal cycling will release part of this gas. Erat the fuel - - - - - - . . .-.. .- -- -- . ... IVIDIJO. bir sana i tesinde yeni .. i . !... ....." .; . CLASSIFICATION (IF APPLICAM.E). ::: '. -INIO Fortwo s 110- V-IN 19121810 DIV-INIO OF TYPE...... FAINING LINES :. .EFT MARGIN--.. im 11. CHAPTER TIILI 1:57 LINE OF TEXTI !UPTOP 097 - I . .. .0) .2 "...i was . . ... . -- - - . - * * - - - Table 2. Comparison of Various ThO2-UO2 Fuels at Farious levels of Burnup 501.. . . : ation Time tor full- power days) Maximum Burnup (fissions/cm3) veraged Peak Linear te 85kr Heat Ratingo Release (W/cm) Fuel . Arc-fused Tho2–6.5% voz (e) x 1020? 2.5. .... 299 - 8.6 :- 376 ; 707 297 (a) 14.4 . . . Sol-gel Thoz-6.59 voz (a) · 2.9 (If Apm CLA:317 . ..267 341 286 33.8 33.7 31.6 37.8 32.8 34.6 b.8. 48.2 -- 2.4 7.2 6.4 0.5 13.2 17.0 22.8 12.6 (c) (c) - (of pigo! RA!! - . . - . ----311 - 140: 110. 119 375 . . . 8.2 ****** 691 23*75*1.5 - 16.5 905 26.4 660 21.1 . 497. 8.1 406 NOTIV:11:,:V 1:) el Thoa 4.5% . : 420 pellet(b) Pressed-and-sintered Tho, Ilets(b) i : 461 ☺.32.an . s 270 .) - - IN3 11.0 394. :52. 42.6 Vibratorily compacted to a density 85 to 87% of theoreticali. Pellets pressed from coprecipitated powders and sintered tɔ 93% of theoretical density. CGas samples diluted with air in sampling and partially lost. . - . - - -- ---...--. --- -. - . ...-. . -.-. --. ---...- . - - - - - - 1 he ---- - -- - . - - - - - - - - - - ------ - ---- . . MAXIMUM '.- DESIREO MARC;!N RIGHT LONLAEC - OFFICIAL ORNL - AEC - OFFICIAL - - -- - - ------- ----- ----- ---- . CLASSIFICATION ( 11 11 ,109!ILE:) - . PAGE NO. .' . 10-15 - - - - TEXT Tel -.ORNL-AEC - OFICIAL -- - ti --- RIGHT MARCIN ..MAXIMULI - - TEXT temperatures of these irradiations sintering occurs in the central portions; thus, with poter changes some of the cracka are cyclically 1. Pormed and healed. Gas release is also a function of fuel temperature, And above some temperature ossentially all of the gu# 18. released. The temperature at which this occurs 18 about 1700°C for UO2 and not yet defined for ThO2-U02, but probably above 2000°C. Another time- and temperature-controlled phenomenon 18 grain growth or sintering. The development and growth of equiaxed grains were most pronounced in the sol-gel materials in the MIR-T group. The unirradi- ated material is uniform with no microscopically detectable structure, but known to contain submicroscopic porosity. On irradiation, this : porosity first appeared to agglomerate. Continued irradiation developed a distinct grain structure in the center of the fuel rod with equiaxed grains 6 to 8u in size at 8.2 x 1040 fissions/cm3 (3.75 reactor full- power days) and up to 6qu in size at 16.5 x 1020 fiøsions/cm3 : : (691 reactor full-power days). A more or less circumferentially oriented crack, caused by thermal cycling, was found at different locations moving withi Increasing burnup or exposure time from the center toward the outer surface of the fuel. This crack appears to be associated with the extent of fuel sintering at operating temperature and was observed in the pressed- and-sintered pellet fuels as well as the s vibratorily-compacted fuels. . Lattice-parameter measurements on these rods, listed in Table 3, show the stability of the lattice structure and indicate a considerable i capacity for fission products. :) 2,.. 50 . . . . . .. ..... -. ...... .inc . . --.. Woo . . --..-. e is? . . O iii. .. .... ... ENO VYPING ........ ....... - ORMAEC - OFFICIAL .. - CLASSIFICATION TIF APPLICAOLE) ... - .. O TVIJ1310- IV-IN? LEFT IVI1360-23 THIO or TYPE ...then 53N17 NIHIVi vil - - 5 CHAPTER TITLE .. NE OF TEXT SOTTOM OF . O-NW on dowo .. SD.- Table 3. Effect of Irradiation on Lattice Parameters of ThO2-002 Fuels 1 . i Unit Cellb (A) Preirradiation Postirradiation . - .. . Buraup Group Rod : Fuel Materiala (fissions/cm3) :: x 1020. MTR-I Ü-2 Arc-fused O 8 .6 I U-3 Arc-fused 14: 4 3 2-7 Sol-gel ::8.2 2-8. Sol-gel . . : :- 16.5 Pellet 645 Pressed and sintered : CENTE12. 8 5 ........ 729 Pressed and sintered 21.1 : 712 Pressed and sintered ... 26.4 : Th02-.59 U02, uranium enriched in 235Tºto 93%. Nelson-Riley Function. : '5.594 0.001 5.599 + 0.009 5.594 1 0.001 5.592 + 0.002 5.593 + 0.001 . 5.594 + 0.003 5.593 + 0.001 5.586 £ 0.004 .590 + 0.001:~** 5.590 + 0.002 5.590 $ 0.001 : 5.584 + 0.002 5.590° 1 0.001. 5.587 $ 0.002 . LIF APPLICABLE) CLASSIFICATION .. - F'AGE. NO. · "lish APPLICABLE)" CLASSIFICATION - - - - - . . - . . .. -- -- - .- - - D...... 1 ENO'TYPING : ... . .",". - . ......... ---- -- - - - - --- --- --- --- . MAXIMUM - VESIREO MARGIN RIGHT ORNL - AEC - OFFICIAL - CLASSIFICATION lof APPLICADLE) - - - PAGE NO. · ORNL - AEC - OFFICIAL The Thô24.5% UO2 fuels show no evidence of breakaway swelling at burnups as high as 2.6 x 1021 fissions/cm”, 120,000 Mwa/tonne of thorium plus uranium, where three-fourths of the fiøsion energy has been derived from the 2330. bred into the fuel during irradiation. : According to estimates (16) of swelling on these pellet fuels the change in volume was limited to 0.46% AV/1020 fissions/cmº in the pellet rod with 21.1 x 2020 fissions/cmº. This value can be compared with the 0.9% AV/1020 fissions/cm® proposed by Anderson(18) for Uo2 based on 1 similar irradiation tests. Theoretically, both of these swelling limits appear to be essentially correct as we shall see in the laters 10 -- RIGHT MARGIN ito. - discussion. ADESIRED MAXIPTUM Effects of Higher Heat Ratings and Sol-Gel Calcining Atmospheres •1. 2 -- - . - . · The MIR-III and the ETR-I groups contain a single sol-gel ThO2-002 fuel preparation which was dried, divided into three segments, and calcined at 1150°C separately under different atmospheres (air, nitro- gen, and the standard (Ar% Hz). The MIR-III group 18 still being Irradiated but the BIR-I group has been examined. Table 4 lists some of the irradiation conditions for this latter group. The experiment was intended to produce central melting in these fuel rods. The post- Irradiation gamma scans and inicrostructures indicate that this was not 0 achieved possibly because during the first 9 1/2 hr of exposure the reactor operated at 50% of full power. Such operation would provide an skice of 50 w/cm which the previous ORR loop tests indicated was high enough to produce sintering and columar grain growth, thus increasing a 2 the thermal conductivity and lowering the central temperatures at rulli .....love insie popo..... .............. is...END. TYPING.. 1 cc . . in N O- ORNL - AEC - OFFICIAL .. .meniemiebedin .. . .. ... me. . CLASSIFICATION lor aprICABLE) .. - -, - :.. - -- h . TVIJ1310- DJV-N10 ,VII10 !0-dir 310-23V-INIO. : . : i. . OF TYPE... RIJAINING I INES LEFT MARGIN- SA CHAPTER TITLE -- HST LIM OF TEXT: oorlom OF :;! - v . - W . . . . . . .. - - - - - li O ... A w a 67 in ico . L-ii----......... - - - - - - - 02 15 Lo apo...en ardom. ' , Table 4. Conditions for the Irradiation of Vibratorily-Compacted Sol-Gel Th026% 10214) Mel Rods at Higher Heat Ratings (ETR-I) . 1 .. . Maximum Burnup - Fuel Rod Calcining Number · Atmosphere Vibrated Density (% TD) Peak Linear Heat Rating E (W/cm) (Mwa/tonne Th + U) (fissions/cm²) . X 1020 I . - (IF APPLICABLE) CLASSIFICATION ***" - -- -- - - - - - -5- 18 Air 88.1 870 20,400 . . 4.3 Ar 4% H2. 88.4 865 20,000 . 4.3 -8 1 8 5.1 -- C910? - - ------22,000 20: Aird. •88.3 870 . . 20,300 ... Suranium enriched in 2350 to 93%. Based on a calculated theoretical density of 10.04 g/cm3. Crime averageu value for the total exposure of 140.5 test reactor full-power days A binary distribution of fuel particle sizes, the other three rods utilized a ternary mixture, ...ci ...... . ; PAGE NO.: (of ATPLICABLE) CLASSIFICATION. 4.3 ; : ENO : ** Y .. .... ......... - - - - - - . . . . - -- . -- - - - - - - . - -- --- - - MAXIMUM ...DESIRE ::TMA!?GIN RIGHT ORNI - A FICIAL ORNL-AEC - OFFICIAL - - - - - . - - - - - - - - -- - - - - - - :::.:.: Sobo ta.at * CLASSIFICATION TIF APPLICABLES PAGE NO. .... 14 ORNL - AEC - OFFICIA pow - - - .- . *.- ...i ... power operation. Col.umar grain growth was extensive and for the first i time in this series of irradiation tests on Th03-1102 central 'voids were formed. The extent of these vo:ids was 3 cm long at the high burnup end. of fuel rod 10 and essentially the full fuel length of rods 8, 18, and X 20 (27 cm). The composite radial micrographs of three rods are shown in Fig. 1. Examination of these microstructures permitted the calcu lation of Skdə values for columnar grain growth and equsixed grain 1.54 growth shown in Table 5 together with data from the earlier loop and ** poolside experiments for comparison. These data can be compared with - vibratoriiy-compacted UO2 fuel rods 18 where voids were formed at " RIGHT: so kde of approximacely 50 W/cm. The fission-gas release based upon .. DESIRED · the 85Kr analysis is moderate when compared with UO2 at similar heat i į ratings but does indicate a possible effect of calcining atmosphere on the release rates. A final judgment will depend on the examination of the MIR-III group. These tests indicate that vibratorily-compacted sol-gel fuels will operate at linear heat ratings of at least 1000 W/cm and that the limit has not been reached. MARGINT MAXIMUM . ... - .. .- • - - . - .. ..... .... et.... ... ..., -- -- -.. . . Mixed Progeny. (ThO2-Pu02). Fuėl Irradiations Three rods containing sol-gel ThO2 mixed with PuO2 before calcining ! and loaded by tam packing have been examined after exposure at linear in heat i'etings up to 245 W/cm to a burnup of 29,000 ma/tonne of rium plus plutonium. The fission-gas-release rates were less than 5% and 5. the microstructures were similar to "THO2-VO2 exposed under the same al conditions. Additional tests of this type of fuél are planned. 311. 2 0 :0" 7. si 1 . . -- OON WAGON 2.0 "Y" :: END. TYPIN :.. ....achowe <.......... ORNL - AEC - OFFICIAL ........ . CLASSIFICATION TIF APPLICABLE) . . ' , ood.. '. " duo.'.... .... . M i ........ 110-IV-INIO Iridisso- Daj - - .. ; . O TYPE........ ING.LINES T MARGIN CHAPTER TITLE - ¿T LINE OF: TEXT . BOTTOM OF 2 . --- .-. Table 5. Effects of Higher Heat Ratings on Irradiated Sol-Ge] ThO2-U02 Fuels . . * ...... 8567 (°C) ...... Heat Ratings for Various Micro- structural Changes& ::.. "Fuel Rod Cladding Identifica- Surface Release Barnup Experiment tion (%) (Mwa/tonte Th + U) ORR LOOP L-I-A 260 46.5 31.2 12.3 1,600 --- L-l-B 260 - 57.0 54.2 35.4 Lost 2,100 L--C . 260 49.6 :. 38.2 3.9 . : 1,230 ORR Poolside 03-5 . 58.5 35.5 18.3. •5,220 --- 06-5 - --- ---..40.1 --- CENTER LINS - m b -- Lost -- 1,040 ETR-I.. 18 100 76.5 49.3 48.2 38.6 . 20,400 10 : 100 79.7 38.1 33.3 27.8 20,000 75.3 42.2 39.3. 21.4 22,000 .. .. :: (IF APPLICAOLE) CLASSIFICATION .. PAGE NO. TIF APIPLICABLE) CLASSIFICATION , : : . heat ratings are time-averaged values for the various integrals: is the central tempera- Jure, limit, Ty, is the limit of void formation, T., is the limit of columnar grain growth, To is the limit of discernible equiaxed grain growth. 'Although equiaxed grain growth was observed the extent was not clearly defined. . 1 . . i... L. Lir . .- . . - - :! . . . - - - - - . DESIRED MARGIN li RIGHT: i. MAXIMUM . . . . . . . - . .. -- -- - - .. . - - - - ---- - ORNL - AEC - OFFICIAL synne JORNLAIC - OFICIAL . .. - .. - . CLASSIFICATION ir APPLICABLE) - . I . ; . PAGE NO. * 167 ... - ORNL - ACC 2oricimit .. i:*:. '-. EXT : o i Li .. -- Semiremotely Fabricated (233U-ThjO2 Tests ** During the past year, remotely fabricated rods of the ETR-II, group containing ThO2-3% 233002 in Zircaloy cladding were inserted in the reactor. They are performing well after eight months exposure at : a linear heat rating greater than 600-W/cm. A similar group of rods containing pure ThO2 (EIR-III) are also under irradiation. This group will be subjected to an increasing linear heat rating which will reach approximately 1000 W/cm after 350 reactor full-power days when the accumulated burnup will be approximately 28,000 mwa/tonne. . These . i -- -- II ---- Trods are in the ETR at a maximum unperturbed flux of 2.6 x 1024 neutrons cm-2 sec S.. . -- - HT MARGIN DESIRED MAXIMUM .. .. . CALCUIATIONS AND COMPLEMENTARY TESTS : -.-• --- -- --- .: Thorium Conversion and Fission-Rate Calculations To evaluate the irradiation tests it has been necessary to deter- : mine the number of fissions occurring in the original 2350 and those occurring in the bred 2330. Since most of the irradiations have involved a number of steps in the unperturbed thermal neutron flux to maintain a nearly constant linear heat rating, it was necessary to calculate the production and consumption of 2334, 234U, 235U, 2360, and 233Pa as ... a function of a variety of effective flux levels. Since the actual 10 is unperturbed flux varies throughout each reactor cycle, and the neutron .yti spectrum varies with each exposure location in the reactor, a .. sophisticated analysis was deemed impractical. Consequently, an exist- ing code, CRUNCH, (20) with a set of effective cross sections based on ij.experience in the 'MIR and ETR reactors was used to calculate the ..- - * - - - - . - .. : . . 1 . ** - - - END TYPING '. ' and rid ing, i,.. .... .... -- .. i CLÁSSIFICATION !!F APPLICABLE) CLASSIFICATION PIR APPLICABLE; PNT NO. . [ 17 OINI - AC - OFFICIAL o BOTTOM OF NIRST LINE OF TEXT P. CHAPTER TITLE - production depletion and fissioning of the various isotopes during the d i ORNL - AEC - Osricia :13 - . -- ::: . . ... - . Irradiation of thorium. In the reflector positions of the MIR and EIR reactors, the epithermal flux is approximately 8% of the thermal flux and the effective cross sections chosen for this flux are tabulated in . Table 6. The use of the resulting curves requires the determination of the effective flux for each step in the irradiation. Again simplifying assumptions are used. The perturbation factor for the fuel rod is calculated by the method of Lewis(21) for the beginning and end of . .10) . . ti - - -- - LEFT MARG - each flux step, and the arithmetic average is used to correct the RIGHT: MARGIN - - - DESIRED .MAXIMUM --- - ........ ---- .. 'a - unperturbed flux level. Because of the variable .operation of these experimental reactors withí significant periods of downtime, an additional correction factor to accommodate the 233 Pa decay is applied. ' Thus, the perturbed flux is multiplied by the reactor full-power days Lars ! of operation and divided by the calendar days involved for each flux i step. . This latter simplification provides a better value for the fissile 2330 bút leads to underestimates of the 234U and 236U concen- trations. This can be seen in Fig. 2 where the resulting calculated isotopic concentrations are compared with the postirradiation mass spectrographic analyses of a number of experiments at different levels of burnup. Nevertheless, this permits a more accurate calculation of the fissions in 2350 and 2930, thus providing a better calculation of į .. as the percentage of fission gas released and permitting a better calcu- OF Type slation of the semitheoretical swelling of a fuel mixture for comparison with the experimental data. .: . --. ; 10.1 --. - - - --- .. - - .. RIMAINING LINES . . IAL 270., Iris . : END TYPING ORNL - AEC - OFFICIAL NI - AEC - OFF! .... CLASSIFICATION LIF APPLICAULE): -roup . - . 4 - - , i i IT!1i10niin) 11 ill l'111. All! :.. ONLÓ AEC - QLEIC Table 6. Constants Used for Modified CRUNCH Calculations . . . - - .: Effective Microscopic Cross Section . .. (10-24 cm2) il . jo:. Isotope Capture : Fission 232mm [ 11.53 233 .;., 13503 150 55 525 Half-Life 1.41 x 1010 22.1'min 27.4 days 1.62 x 205 2:48 x 105 y 7.13 x 108 y 2.39 * 107 y 6.75 days 233 pa 233U : 2340 2350 : 2360 23'0. 146 110. ; ii RIGHT MARCIN 585 .. 30 .: DESIRE.CO - 210 :: . 585 MAXIMUM - 1. i ! SNI 1. 25 : *: monde de cicado en ! . . . 1. '30 . . -.-. - ...- - - . - ........ -- . ---- - ----- ...... ---- 1 . -W.. . .... . is comico --.non . 5 . ở * 1 . indir -. s, a sº - --. . - OFFICIAL arm - AEC - . -ộc • N “ in ...END. TYPING: IN CLASSIFICATION TIF APPLICABLE) CLASSIFICATION Tlf nipli 4:.) PAGE NO. 1.197 BOTTOM OF FIRST LINE OF TEXT 1 OUR CHAPTER TITLE: ; Reactivity Effects ORNL-AEC - Orniciat - OFFICIAL - - - - - -- - - - - In an effort to establish some idea of the influence of irradi- ation on the reactivity of a thorium-base fuel the data calculated by i the CRUNCH code, were combined with the adsorption cross sections and į! eta values for 2350 and 233u to determine the change in effective fuel is reactivity with increasing burnup. Thus, an effective regeneration : factor, neppe for the fuel including all fertile, fissile, and fission- product influences was calculated. This, then, provides a parameter which 18 independent of the other considerations normally used in the - - . . . . .-. - - : - - . - - . .. - - . - -- LEFT MARGINEA "four-factor formula" for the calculation of the infinite multipli- · RIGHT 1 MA!!(:IN 01: S100 cation constant, k, but does define the influence of flux and exposure - MAXIMUM . -... on the reactivity of the fuel. Figure 3 is a presentation of this parameter for a thoria-base fuel originally (0.984 23279-0.016 233U)02. This composition is shown since it represents approximately the equilib- rium concentration. Acaitional fissile material can be added either as separate seed fuel or to the mixture in order to provide a critical array without changing the basic form of the curves. 1. The nepe curves point out the relatively small changes in reactivity for ThO2-base fuels with increasing burnup. This could reduce reac- tivity control problems and with fuel management lower the peak-to- average power distribution in a reactor while still permitting burnups well in excess of 50,000 mwa/tonne of thorium plus 'uranium. .-. -. . . -. - . -* .. • .... . ..... i A REMAINING LINES O OF TYPE..... ORNL - AEC - OFFICIAL 6 Flosion-Product Swelling Calculations The semitheoretical swelling calculations have long been used by : ::..? Ifradiation experimenters.. Brinkman(22) proposed a concept for calcu- END TYPING ¿ lating the swelling of metal fuels. If one assumes that the fissile, C. .. CASSINICATION !!!? ORNL - AEC - OFFICIAL ' d ' .. ., TIF niri ICADI.F.) : e. CLASSIFICATION lil ALT1.11.111.11 comments Mission .corte mousse · Marof NO. 12017 - AEC - OSSICIAL ther - ..... ..... o . exertion s i fertile, and fission-product atoms in the limiting case under maximum restraint each occupies a volume equal to its volume in the pure sol.id form and that the effective volume of an atom in any compound is equal 1:1; to its volume in the metallic lattice then an effective volume change for any nuclear reaction can be calculated. Using the weighted average volume for fission-product pairs described by Brinkman the effective: volume changes shown in Table 7 were calculated. 1. Justification for the use of these values lies in their remarkable agreement with experimental evidence. For example, using 200 Mev/fission, * there would be 2.44 x 1020 fissions/cm2 in a theoretically dense UOz: fuel which had 1 at. % burnup. Multiplying the fissions per cubic centimeters by the effective volume chąnge, one obtains 1.54% AV/at. % burnup which compares favorably with the 1.65% AV/at. % burnup predicted by Anderson(28) from a review of a mass of experimental irradiation data at moderately low temperatures or with a high degree of cladding restraint. Similar comparisons can be made for UC2 and uranium metal awe RIGHT MANGI! ::08 Siren MAXIMUM .. there ...... - . -.. . .-. W :, irradiations with even better correlations. . When one considers the effect of capture' on gross swelling in the two common fertile materials it is apparent that the conversion of.: . 2380 to 239Pu has very little effect, while the 232 In to 2330 conversion : '. has a significant effect. "This is particularly true for the oxides i ..91 where the theoretical concentration of heavy element atoms in ThO2 18 1:9; 93% of the concentration in UO2. Thus, for a 78 decrease in fertile s! atom density there 1. a potential 20% decrease in volume expansion at 91 the same conversion and fiosion levels, as shown by the last two items ! l in Table 7. When this is combined with the higher capture cross section . --.. --.. - 2 in - .-. - . CLASSIFICATION in ApruicalLET .185.1!ION 117. Apigny IP proto, [21] ORNL-ACC - Orviciul torri! OF 1 LINE 0% Iri ilAMPIR TIL.fi " Table 7. Effective Volume O or Atomic Reaction - 5 .. .. Reaction : - - .--. *- --. - ,-.. -- Atomic Volume . (cm /atom) Volume Change: 1 Final Initial (cm/reaction) Ktion) ; 1: : x 10-23 x 10-23. : * 10-23 8.368 : 2.028 . : +6.340. 8.368 : 2.046 | +6.322 1 :1 8.368 · 2.000 : : +6.368 ::. • 2.000 (2.072) -0.072 2.512 .3.304 : 0.792., 1. . 2 2.028 .028 3.302 3.304 -1.276 2016 ... 8.368 3.304 8.368 : -2.072 i +6.296 :::.. :00:51 - MAXIMUM . . . . : :10 2330 pilsion 2350 fission . 239 Pu fission 2384 - 239Pu Th -233P& Th 2330 i T MARGIN--.. in + 233U fission... *** 238U 239P4 + fission .-;-- i 15 - - RIGHT MARGINT ---- ... +5.066 ::.. w . -,- • ............. ...... .. CENTER LINE -.- 2 -. 1. . .. .. - -- simeseni "! - . -- . .. i .--..-..... --- - - .. ..! 'Bli 7.- 43 6: - -.- - .NING LINES 'F TYPE --- .--.- - ORNI - .nw -- -- OKNI - AEC - OFFICIAL. ---- -END TYPING ... . ., 1 - - OFFICIAL: ::CLASSIFICATION' oliF APPLICAOLET** . : (LASSIFICATION TIF APPLICABLED PAGE NO. OF OF TEXT TITLE -- of thorium in a thermal neutron spectrum, it is apparent that the ...thoria-base fuels can be taken to significantly higher levels of exposure before fiel expansion endangers the integrity of the metal cladding. ......-- ORNL REC-orricia - - - -- --- RICHT MA:611 3 einen Thermal Conductivity Thermal conductivity data on ThO2-U02 mixtures are limited. This is particularly true for the sol-gel materials. At ORNL We are con-. ducting an investigation of the thermal conductivities of a variety of sol-gel compositions fabricated into pellets. Only preliminary data are available as yet. The thermal conductivity was determined at ...30, 80, and 120°C in an axial flow apparatus on air-firech pellets with a density of 90% theoretical. The conductivities for (Th–8% U)02+x were 10.033, 0.032, and 0.032 w cm-1 °C, respectively. "These data are . somewhat lower than those reported by Kingery(23) several years ago but are of too preliminary a nature to be considered conclusive. The effective thermal conductivity of the compacted powders was measured in the only two instrumented fuel rods irradiated to date. The Th02-2.9% 102 powder compacted at 85% density gave effective in thermal conductivities of 0.021 W cm-20;-1 at an average fuel tempera- ture of 990°C and of 0.017 w cm2 °C-2 at an average fuel temperature in; of 1325°C. These conductivities compare favorably with UO2 conduca.' tivities in the same temperature range 0.03 i 0,027w2cm-2.0difor, 7.ti' 61-95%. denge pellets. .. .- - .- . : wla nas : ? Sin Ol........... ............. CLASSIFICATION (IF APPLICABLE) ; US CLASSIFICATION lil APPLICANT.:). PAGE 910). [ 23 ORNI - AIC - OF MOTTOM OF 115T LINE OR TEXT LOR CHAPTER TITLE A Gas-Evolution Experimento .ORNL - ARC-OFFICIAL . - One of the areas of concern with powder-compacted fuels and with in -. ! CIAL. -. the chemically produced 801-gel material has been th e inclusion of gas .. - - . - - .. . --,. - -- - --- ... - LEFT MARGIN... RIGHT - such as hydrogen, oxygen, nitrogen, or water which could be released under irradiation to cause high internal pressure and premature failure of the cladding. The most common specification for 102 pellet fuels in " the United States today is a gas release of 0.05 cm;'g of fuel at . temperatures of 1200°C. Although gas release valises lower than this . can be obtained with large sol-gel fragments, the usual particle size :: distribution used for vibratory compaction yields 0.2 to 0:3 cm?/8 of MARGIN fuel. . . . . . DESWED 4 MAXI-AUM The active surfaces of the fired gels readily sorb gases, partic- ularly CO2, which can be released under irradiation but atmosphere controls at critical stages in the process ma, control this sorption. This phenomenon is being stuåled extensively at this time. ... There has been, in addition, considerable concern over the possi- bility of trapping large quantities of gases in the structure of the gel during its thermal decomposition. Removal of such gases would be 'very difficult. A recent experiment has shown the quantity of such entrapped gases to be extremely small. The technique employed was modified vacuum-fusion technique. The :9 fluxing agents Bed and. Al2O3 were used to lower the melting point of the oxide from 3300 to 1810°C. The binary BeO-Al2O3 eutectic was pre- melted to eliminate sorbed gases associated with the fiuxing agents. · Single large chunks of sol-gel oxide about 1 to 88 in weight were melted in an Induction-heated molybdenum bucket No Rostmelting . . . . . . . . . na com a vo öz TIAINING LINES O OF TY!'C.. 1:0) ORNI - AEC - OFFICIAL ORNL - AEC - OFFICIAL ..... .. ...... ..... ... .. . .. i ' . : .. CLASSIFICATION PF APPLICABLE). - . + : CLASSIFICATION (11 111116.1!31.1.) "AGE NO. [241 OM OF NE OF TEXT 1 -- - .. . . . 10 . . .-. . wo. Wom ARGINS RIGHT' MARGIN : new, V.-Y., DESIRED 1. MAXIMIUM " is. Pernai ER TITLE . metallography showed that complete melting of the sol-gel thoria. chunks had occurred after 40 min in the molten Al2O3-Beo eutectic at.. : 1900°C, Gases evolved during the melting of the ternary eutectic were pumped via a Toeppler pump into a calibrated volume. The. gag pressure W@8 reasured periodically with a modified McLeod gage. Results of the melting experiments are compared with outgassing data from fine and coarse powders of sol-gel ThO2 at 1200°C in Table 8., It was observed that the gas release from bulk sol-gel ThO2 fired in Ar-4% H2 was negligible when compared to the outgassing of fine powders. Air firing of a duplicate sample had no appreciable effect on the gas 1 release, although calcining initially in air appears to increase the amount of sorbed gas. -- 25 The firing atmosphere had a marked effect on the quantity of adsorbed gas for particles in the intermediate size range but had "a: minor effect on -200 mesh material. Calculations indicate that 0.2 cm'lg STP of Pission gases are: produced per atomic percent burnup of (Th-U)02. Thus, the release of adsorbed gases within a fuel element may have the same effect on the internal preasure of a fuel element as 15,000 to 20,000 Mwa/tonne :1:!: fuel burnup with 100% fission-gas release, although the Irradiation tests to date have not indicated any such effect. Further work is ...! in progress to more clearly define the extent of the gas adsorption problem and eliminate it. I t i ... 25 r .. .. . 101 miel. .. -. . -- .. -- - .- ... * r - * . . ' -- . ., C LINES EYPE une point .. rom ...! 's . . i END TYPING i . 'ORNBAEC - OFFICIAL : . ... CLASSIFICATION ilor AMPLICABLE! . i :. euri.rona in . : . WIJI390 - 93V - INIO OF TYPE - ITAINING LINES INI? . grup MARGI HAWTF.N TITLE. OTTOM OK lidt.10 . . 1 S . ' .: 1.' is gor - * awa :. Namun . . . - -- -- _ E l 13 i 1 Table 8. Gas Release from Sol-Gel Tho De oe ve nedes Dubois sua .. si ei 43 ... Firing Temperature Density 7°c) (g/cm²) · BET Surface (m²/8) Mesh Size Sample Designation Gas Release (cm/& STP) at 1200°c On Melting - Firing Atmosphere 1150 35-) Alt . : 1150 2. 35-2 Arm/% H2 - 0.337 -0.394 0.007 0.319 0.022 0.330 Are yo 12. PAGE NO. IF ANFICABLE CLASSIFICATION' a. : 1. 35-3 : -9.99 ... 6 +16 0.027 -200 - 10.570 - +] 6 0.004.14 -200 0.820 6 +16 0.005 -200 0.790. +2.5 -0.0003 +2.5 0.0003. +2.5 0.003 +2.5 0.001 +2.5 0.002 H-1 - مسلم خشمسه لل مسلسنس نظفند · 1.APPLICABLES LASSIFICATION Nitrogen. ..i Art E2 . (Air-fired sample of H-1) A ir -- Ar% H2 . ; N2 H-2 PI- B PI-2 PL-3. 1150 9.49 2250 9.49 2150 73? ****9.91 2150... 9.95 1150 : 0.0030 0.0028 0.0135 0.006 0.0101 0.003 0.002 0.002 ..-. ..... ...etc.. .! . 1 • ... - - - - : - . , '.. . . ENO TYPING .! .. : ... T. MAXIMUM DESIRED MANGINI RIGHT ORNL-NEC - OFFICIAL * - 1-AEC EC - OFFICIAL * 2 ** CLASSIFICATION (IF APPLICANT.61 ingen MinGf NO, or OF TEXT ? R TITL.. DISCUSSION w . The available data on the irradiation testing of thoria-base bulk ! oxide fuels are not complete. Additional work is required on the determination of thermal conductivity, gas sorption, and in establishing 1 -.,... the upper limits of operating characteristics. Without these the most. economical reactor designs cannot be derived. The high melting point and stable structure of ThO2 when combined 1. with its unique breeding potential in thermal reactors indicate possible SIN ---- economies. T:RIGHT MARGINI DESIRED) MAXIMUM . . . . :. --- : Although the data on ThO2-base fuels are st111 somewhat limited, & preliminary comparison with UO2 fuels is in order. The data presented ..::" in Table 9 compare powder-compacted. YO2 and Thó2-U02 fuels using the en mandi : Skde values for specific structural changes most commonly used in con- :...paring. bulk oxide fuels. (24) The toz work was done as part of the Maritime Reactor Program and was reported previously. (19–25) The ETR-I data were included to provide a value for the Škde for void formation in the thoria-base fuels... Although there are uncertainties in the peak-to-average ratios of y heat flux in these experiments, it is apparent that the thoria-base. fuels will withstand higher thermal ratings with reduced microstructural changes and lower gas-release rates. Out-of-reactor thermal simulation tests are planned to establish the heat flux and temperature ranges required to produce columnar grain growth in the thorià-base fuels so that more precise quantitative comparisons can be made. 1:59: In most economic evaluations the two principal cost factors in. I' thorium fuel utilization are Inventory charges and fabrication costs. . . . - - TES - - ! . END. TYPING : ASC-05 - . . . CLASSIFICATION lor Apipl.ICABLE) -... ::- . . .: : M . i ini . . -, . ..... ..TC-8: aping, ! 3:10:33 O F Typ !! INING 1. 1611310-31-INIO for p MARGE OM ER TITLE 1.LING PC TEXT B. .. -0 :... . .. .. . Table 9. :: ... . .:.: Comparison - -- - -- - ...!! - --- . - -- - --- O2 and ThO2-U02 Vibratorily Compacted Fuel Rods : - Heat.Ratings for Various Micro- structural Changesb (w/cm) 85kr czejding Rating sa kao kdo jesh kdo po kdo Release Release . 86.9 6,290 . . ..85.6 ..32.5 : 20.7 34.8 21.1 ..41.2 20.0 30.9.: 22.8 ".-31.3 . 21.2. .. 23.9 Inside .: : . burun Surface Linear · Fuel Mwa/tonne Temperature Heat Experi- Fuel - Density Heavy of cladding Rating ment Materiala (%7D) Element ORR Loop_U : 711 .::. 87.1 4,780 353 434 :: 87.1 . ' 5,140 361 465 :: 86.9. 6,290 · 383 .: 566 " 4,810 347 404 801 : 85.8... 5,040 351 - 423 8P1 . 85.5 ... 6,880 387 578 · ORR Loop ThO2-6% UO2 . : LA .. . 85.2. . 1,600. 341 381 368 LIC .: 84.1 . 1,730 * 348 .: 410 ETR-I · ThO2-5% UO2 . . . 88.1 .. 20,400 211 870 .. .10 :... 89.4. 20,000 209 .865 v 4-8 . .... . 85.1 22.000 218. 914 12.4 47.0 72.0 77.0 25.0 25.0 1, 11F APPLICABLE) CLASSIFICATION 17.5 16.1 20.3 · :los av PLICA .C.MOTIFICATION CENTER L!!;5. - PAGE NO. - LIB . . :: : .2.3 84.1 2,100 499 30.3 No void 39.7 No void. 36. 32.6 No void . 20.5 17.7 20.7 !! -18 . . !! ENIO TY 63.0 66.4. 61.3 35.8 24.6 28.2 34.7 19.8. 25.3 ... 38.0 28.0 22.0 2 002 fuel was arc-fused crushed and vibratorily compacted; oxygen-to-uranium ratio 2.002:2.003; uranium. enriched in 2350 to 56%. ThO2-U02 fuel was sol-gel material crushed and vibratorily compacted; i 'uranium enriched in 2354 to 93%. - "The heat ratings are time averaged values for the various integrals: reis center, to surface of fuel, is void to surface of fuel; pos is the limit of columnar grain growth to surface of fuel; and se is the limit of discernible equiaxed grain growth to surface of fuel. 127 . *.- MAXIMUM DESIRED MARGIN RIGHT - -- - S onNEC - OFFICIAL . . ORNL - AEC - OFFICIAL 2- in. . MT S . . - - - : CLASSIFICATION (l! APPLICABLE) MAGT: HO. [28] ... EXT! LED " The fact that thorium has a higher capture cross section for . -.. · - thermal neutrons than 238U requires a greater Inventory of. flssile;. material, all other factors being the same. Also, since the fløsile .it material mixed with thorium 18 in separated form, rather than as natural! or slightly enriched uranium, it makes the inventory still more expensive Fabrication costs on the other hand are certainly a function of the fuel form and the mode of encapsulation and assembly. For fuel rods, .: Lotts and Douglas (26) have shown that fabrication costs in dollars per i - kilogram can be reduced by increasing the diameter of the cladding ! om tube,_Bettisi(27) experlence Indicates a similar conclusion in that the militar fabrication cost per centimeter of rod remains nearly constant over a wide range of diameters. Thus, the fabrication cost in dollars per kilogram is equal to this constant, in dollars per centimeter, multi- : plied by the length of rod per kilogram of heavy metal. If the con- trolling factor in fuel power is the linear heat rating imposed by 'such '. things as central melting of the fuel or ission-gas-release rates, then :! It can be shown that fabrication costs increase linearly with specific ...power. This is done by substituting the identity equation: i kW/kg = (kW/cm) (cm/kg) into the previous equality to provide -- ..'..DESIR(1) ..MAXIMUM - - -- -- .. - - - - .. .. . - . - . .-- .. - - - 10 . . . .. 10. .:.. $/kg = (cm) (kW/kg). 1... .- -.-.-.-.- 71...ri This same equation indicates that any allowable increase in the linear aj neat rating will lead to a reduction in the fabrication costs and/or . lower Inventory. į . ini ................... . ENW TYPING: ORNI - AEC - OFFICIAL : : . CLASSIFICATION :.... lor APPLICABLE)" :* CI. 197911 ICATION (I APLICA!! ! MAGL. 110. | 29 и потом ОF fin, LINE OF TEXT OTCHAPTER TITLE . . -- + spomies indian . . i. more Thus, the establishment of higher linear heat ratings safely attainable in ThO2-base fuels may Indeed contribute to improved costs, although the surface heat fluxes will have to be optimized with other | variables such as pumping power. Although most economic optimization studies, such as those by Rosenthal et al. (?) that compare thörlum- and uranium-fueled systems, generally indicate higher exposures for the thorium-fueled systems in the range of 20,000 to 40,000 Mia/tonne of thorium plus uranium, the peak-to-average exposure requires a significantly higher safe exposure. level. The tests to date Indicate such a safe level in excess of.. - 100,000 Ma/tonne of uranium plus thorfun with very little change in EFT MARGIN RIGHT MAFGIN . --- . ... DESIRED) : MAXIMUM ... . . · fuel reactivity. . . . . !-- ?5 These properties of thoria- fue Is when combined with the IOW . . - • costs associated with the sol-gel process indicate that such fuels have excellent prospects for providing an economical nuclear fuel for power reactors, particularly if the reactor design is based on the thoria rather than urania limitations. - - - : - - -.- - - CONCLUSIONS: .... " wir-.-.-.. - . . . .,. , - -- - As a result of the testing program to date, the following con- clusions can be drawn: 1. Thoria-base fuels with uraniurn contents less than 10% will 7 withstand exposures in excess of 100,000 mia/tonne of thorium plus 5: uranium with no evidence of breakaway swelling or sudden increases in fløston-gas release at linear heat ratinge between 300 and 450W/cm kde öf.32.16.48 W/cm). END TYPING ! - - MA:HING LINES TYPE -.phoon - - NL - AEC Oss 1. 2;. in . . . I ...' nd ... ORNI -AE . ... .. CLASSIFICATION, IF APPLICANILE) ***** ' IAL ** ** ... . ... CLASSIFICATION (IF APPLICABLE) Parir 10. [30;*. WL-AEC - Official OF F TEXT TIT - - . 2. Vibratorily-compacted sol-gel derived thoria-base fuels com- ! pare favorably with vibratorily-compacted arc-fused and pelletized fuels. ........ .. men jawaban 3. Vibratorily-compacted sol-gel thoria-base fuel will operate at a lineer heat rating of 1.000 W/cm (5.9kde of 100 W/cm) to burnup levels! :.1 of 20,000 ma/tonne of thorium plus uranium with no evidence of central ? | melting and fission-gas-release rates less than 30%.*** in ** 1.1" 4.. The sol-gel process does not lead to gas entrapment within the particles although the surface sorption characteristics are not fully ::. Metro - :.'. i SIN ---- RIGHT MARGIN iii. defined. ..;0 5. Semitheoretical calculations of the minimum possible swelling - ... DESIRED MAXIMUR . . . . . . - --- --- - -- : -- -- - i . - - -..- . . . in UO2 and Th02-V02 match closely the experimental evidence and Indi- cate at least a 20% lower swelling rate for bred 2330 fissions than for bred 239Pu fissions. 6. Postirradiation analyses agree well with calculated levels of fissile atom content up to 120,000 mwa/tonne of thorlum plus uranium . burnup and indicate relatively small changes in fuel reactivity even at flux levels as high as 1.0 x 1024 neutrons/cm2. . 7. The structural stability and higher thermal capability of . ii: |thoria-base fuels, when compared with UO2, indicate potential savings in ? Inventory and fabrication costs provided the reactor design is based • on these fuel limitations. ' . معده بسندهما 1.AZ swed ; 10), .. . 7 . - NES ACKNOWLEDGMENTS -. . --- ..21 to The efforts of "many people are represented in the work reviewed 1 this paper, and it is irqpossible to give full:credit to each. In the Puel Irradiations Program special note 18 due to s: d. clinton of the ORNL - AEC - OFFICIAL . - ENOTYPING: . --.. Tin CLASSIFICATION ... (IF APPLICABLE): ... .: C:1 1::SIFICATION Ill APPIAH!!! I'Nil No. ( 31!. ORNI - AIC - OFFICIAL OOTTOM OF 1 ST LINE: OF TEXT !! O'CHAPTER, TITLE: .-.- .:ORNL - AEC - QELICIAL Chemical Technology Division for preirradiation calculations and from - - - --- -- - -. - - . . - - - - - --- . -- - 10) the Metals and Ceramics Division to W. S. Ernst, J. D. Sense, J. W. Tackett, and R. W. McClung for fuel-rod fabrication, welding, and nondestructive testing; E. J. Manthos for metallography; and to W. C. Thurber and 8. A. Rabin who supervised many of the early experi- ments. The CRUNCH calculations were made by E. D. Arnold of the Chemical Technology Division. The thermal conductivity measurements are being made by J. P. Moore on pellets fabricated by J. M. Robbins, both of the Metals and Ceramics Division. R. A. Bowman performed the utai 7 oxide melting experiments to determine the quantities of trapped gases while the other gas-release data were obtained by the Analytical Chemis- try Division. To these and all of the others who contributed, i the aüthors wish to express their appréciation, FT MARGIN-W RIGHT MARCIN - - - .. - DO:511170 im MomXTURAL - - - - - -- --- . - - - -.c. ch.dw. . -.*:* * :* - ...-. i CENTER - - - -- --- - - . - - .. - - . -- . . - .. ... : . -----... .. .:... - 'IINING LINES O'F TYPE ... Moment 5 1:1 ::. . ? : ....... -.- 21... 50 . . OINT-AEC - OFFICIAL END TYPING - AEC - OFFICIAL . ........ .. CLASSIFICATION .: (IF APPLICABLE) . .. : CL. ASSIFICATION lll Arn.CABIT) L'Arti NO. ORNLOAEC - Orucian -------Basterminations serisine meccaniche . .... ... . .... .... .. .... TEXT! ITLE REFERENCES - A - - - 1. Lane, J. A., "The Economic Incentive for Thorlwn Reactor Develo;- . 1: ment," paper presented at the IAEA Panel on "Utilization of Thorium 1 in Power Reactors," Vienna, Austria, June 14-18, 1965. . I 10 ! 2. Rosenthal, M. W., Bauman, 8. F.; Bennett, L. L., Carlsmith, R. s., and Vondy, D. R., "The Technical and Economic Characteristics of Thorium Reactors," paper presented at the IAEA Panel on "Utilization i 15 of Thorium in Power Reactors," Vienna, Austria, June 14-18, 1965. RIGHT MARGIN DESIRED!: MAXIMUM! ----------... Duret, M. F., and Halsall, M. J., "A Preliminary Assessment of it Thorium as a fuel for Thermal Reactors," paper presented at the IAEA Panel on "Utilization of Thorium in Power Reactors," Vienna, 11. Austria, June 14–18, 1965." 23 .' . io . .. 46. Crowther, R. L., "Usè of Thorium in Boiling Water Reactors," paper presented at the 6th Annual Nuclear Congress. Symposium on "Uranium- Thorium Cycle," Rome, Italy, June 13–15, 1961. 17 1. r 110 . 5. Rosenthal, M. W., Adams, R. E., Bennett, L. L., Carter, W. L., ".. Douglas, D. A., Jr., Hoskins, Ri E., Lawson, C. G.; Lotts, A. L., 4. Olson, R. C., Perry, A. M., Roberts, J. T., Salmon, R., and". Vondy, D. R., A Comparative Evaluation of Advanced Converters, ... 15 USAEC Report ORNL-3686, Oak Ridge National Laboratory, January 1965. ** ... .- . 101. .. - -- WORM - - - . . 1.6. Van Cleve, J.J E., and Lotts, A.' L."Proceedings of the 12thcoorifer. 1.50ence on Remote Systems Technology" (Hingdale, Illinois: American Nuclear Society, 1964) p. 257. . . ORNL - AEC - OFFICIAL END TYPING . . • ... 1 . ..... ..in ORNI -A CLASSIFICATION :: "... (IF APPLICABLE) . - . 11.AC3!:10: ANTON Ifi 111 1 '11i - - '' niet te 33 - - - ORNI - AEC - OFFICIAL .BOTTOM OF 3T LINE, OF TEXT I - --ORNL-ACC - Orlicia - -- -- 3. - - - - - .... - - CHAPTER TITLE.. 7. Haws, C. C., Matherne, J. L., Miles, F. W., and Van Cleve, J. E., Summary of the Kilorod Project – A Semiremote 10 kg/day Demon- · s stration of 233N02-ThO2 fuel-Element Fabrication by the ORNL Sol- Gel Vibratory-Compacted Method, USAEC Report ORNL-3581, Oak Ridge National Laboratory, September 1965. - ...... ! .: ... ....... po 10 . - ... rar . . .. - ... - - - - ... .... Wymer, R. G. and Coobs, J. H., ""Preparation, Coating, Evaluation; and Irradiation Testing of Sol-Gel Oxide Microspherer," this conference. . - . - -' . . . .Om EFT. GIN mendi RIGHT MARGIN . . 11:095119 Ferguson, D. E., Dean, 0. C., and Douglas, D. A., "The Sol-Gel Process for the Remote Preparation and Fabrication of Recycle Fuels," :-MAXIMULA 1 paper presented at the 3rd United Nations International Conference --- 25 on the Peaceful Uses of Atomic Energy, Geneva, Switzerland, August 31-September 9, 1964:5" > ara ' L ' '. - ...- con 100 Wymer, R. G., and Dồuglas, D. A., Status and Progress Report for ::: Thorium Fuel Cycle Development for Period Ending December 31, 1963, USAEC Report ORNL-3611, Oak Ridge National Laboratory, July 1965, pp. 81-112. .. - - + - 35 - - - - - - - - . - - . . . - - - - . - 10. 11. Wymer, R. G., and Douglas, D. A., Status and Progress Report for Thorium Fuel Cycle Development for Period Ending December 31, 1964, 1, USAEC Report'ORNL-3831, Oak Ridge National Laboratory (in press). . . . - .. . To 15 . EMAINING LINES OOF TYPE....Homes $ . ;D 12. Rabin, 8. A., Status and Progress Report for Thorium Fuel Cycle Development for Period Ending December 31, 1962, USAEC Report . só ORNL-3385, Oak Ridge National taboratory, October 1963, pp. 102–128." ENOTYPING oRt.AEC - OFFICIAL m aintenance... wasani.reddito!..;) AEC - OFFICIAL - **1 * . . : CLASSIFICATION (IF APPLICAULE) : - - . - . - : - ** . : CLASSIFICATION HA AF!!!! :(. Ali ; TAG: NO. 34 TEXT TEET.' 13. Rabin, s. A., Osborne, M. F., Clinton, 8. D., and Ullmann, J. W., . "Thorium Fuel Cycle Irradiation Program at the Cak Ridge National 5. Laboratory," Proceedings of the Thorium Fuel Cycle Symosium, ✓ Gatlinburg, Tennessee, December 5, 1962, USAEC Report -TID-7650, '1. Bk II, p. 643, AEC Technical Information Division. ...ORNL - AEC - OFFICIAL -- - i 10 -- - .. - -.---...--. 14. Rabin, s. A., Clinton, s. D., and Ullmann, J. W., "Irradiation of Nonsintered ThO2-U02 and ThO2-PuO2 Fuel Rods for Power Reactor - 15 Application," paper presente 1 at the Symposium on Powder Filled ..Uranium Dioxide Fuel Elements, Worcester, Massachusetts, 20 November 5-6, 1.963. ..in-: 316illi Mlipii!!! --- - . . 001:56:1; ..MY!!! 15. Olsen, A. R., Trauger, D. B., Harms, W. O., Adams, R. E., and 1 - 25 Douglas, D. A., "Irradiation Behavior of Thorium-Uranium Alloys and Compounds, " paper presented at the TAEA Panel on "Utilization i of Thorium in Power Reactors," Vienna, Austria, June 14-18, 1965. i .. 16.' Rabin, S. A., Ullmann, J. W., Long, E. I., Jr., Osborne, M. F., and Goldman, A. E., Irradiation Behavior of High-Burnup ThO24.5% UO2 Fuel Rods, USAEC Report ORNL-3837, Oak Ridge National Laboratory, October 1965. . ... : . . . 17. Carroll, R. M. , and Sisman, 0.; "fission-Gac Release During Fission- ing in U02," paper presented at the 1965 Annual Meeting of the 15. American Nuclear Society, Gatlinburg, Tennessee, June 21-24, 1965. l , 3 :: 670 18. Anderson, T. D., Nuci. Safety 6, (2), 165, 1964–1965. . ?.. ! ;. . -...- vw OTMT - AEC - OFFICIAL Tiwi .... ENO TYPING CLASSIFICATION .: TIF APPLICABLE) ORHI 01.19,1FICATION Ill MTL II All! "161 10). 35 (IOTTOM OF 75T LINF. OF 'TEXT ! O. CHAPTER TITLE: pon. ILO AEC - OFFICIAL ..ORNL - AEC - - - - - 19. Haynes, V. O., Thurber, W. C., and Long, E. J., Jr., "Fuel Irradiation Tests," Maritime Reactor Ann. Progr. Rept. Nov. 30, 5 1963, USAEC Report ORNL-3775, Oak Ridge National laboratory, pp. 61–77. - - - - - - . . . . . . . . 20. Ietzke, M P., and Claiborne, H. C., CRUNCH – An IBM 704 Code for Calculating N Successive First-Oråer Reactions, USAEC Report ORNL-2958, Cak Ridge National laboratory, October 1960. - .-. :: 21. Lewis, W. B., Nucleonics 13, (10), 82; 1955. LEFT MARGIN--...- .. . : 226 Brinkman, J. A.; "Nuclear Metallurgy" (New York: American Insti- .. BRIGHT MARCIN DESI! 9X1:10M Metallurgical and Petroleum Engineers, i - Vol. VI, pp. 1-11. 3 . - 25 si ... - - - 23. Kingery, W. D., J. Am. Ceram. Soc., 5, (12), 617, 1959. . - , : i 246 Robertson, J. A. L., ſkdo in Fuel Irradiations, Canadian Report CRFD-835, April 1959. :: ; 25. Osborne, M. F., and Long, E. L., Jr., Post Irradiation Examination : of Maritime-ORR Loop Experiments 5, 7 and 8, USAEC Report ORI, ORNL-TM-921, Oak Ridge National Laboratory, October 1964.' : O . .. . 10:26. Lotts, A. L., and Douglas, D. A., Jr., "Refebrication Technology for the Thorium-Uranium-233 Fuel Cycle," paper presented at the 161, TAEA PaneI on "Utilizat :on of Thorium in Power Reactors," Vienna,? ! Austria, june 14-18, 1965. TAINING T.INES or TYPE....... END TYPING doo - ORNI – AEC - OFFICIAL www. AEC - OFFICIAL wow....... , : .CLASSIFICATION (IF APPLICABLE)... 1 . lib 15FICATION ll! ANTIC 101) .. PAST. NO. - ...... ORNL - ACC - OFFICIAL ini. C - 27. Large Power Reactor Program, Interim Report, L'SAEC Report . WAPD-LPR-141, Westinghouse Fectrie Corporation; Bettis Atomic 5. Power Laboratory, July 1963, p. A-5-1. .. : . 10 .. . . . .. 15 . . . . .746..:: . . - - RIGHT MARGIN - : DESIRED pracy on VAXT . ..1mkin". TER LINE ----.. . ---- an entend wees. I - ---- - -- - - .--. .. . .' ... more... ......mi, v .. .. ... ..-. . . -:* - .. . . inir.i. . . :. *** . *---* . : : CIAL go i. mo ::;ENO TYPING ...---******* .. . . A. . . . .. ::CLASSIFICATION ; (IF APPLICABLE),.. . . TO r. : . ORNI - AEC - OFFICIAL FIGURE LIST ...... . Fig. 1 (Photo 80962) Composite Radial Micrographs at Maximum Burnup Locations of Vibratorily-Compacted 801-001 Th02-6% UOPuol Rod. Oled in : Type 304L Stainless Steel. : Fig. 2 (ORNL-DWG 65-9381) Efect of Exposure on the Reactivity of an (0.984 T-0.016 233U)02 Prel. :.. fig. 3 (ORNL-DWT 65-9382) Change in Uranium Isotopic Concentration as a Pinction of Irradiation Level. : . . . .. . . .. . ii. ... . 0 ... .: , . ... .. .... Taigiasi ini dan men.. : ... . ..... son. 17 . .,. '. ORNL - AEC - OFFICIAL ron a .. ORNI - AC - OFFICIAL . ... PHOTO POWI . " re ZOOM - 2.64 mm AIR FIRED FUEL IRRADIATED TO 4.3 x 10" dissions/cm' AT A LINEAR HEAT RATING OF 870 w/cm. - - . . - . . . . . . ini la 2,84 mm ...----------- . . ... .............. .........-...-- - ------... . . . FUEL FIRED IN Ar-4% H, ANO IRRADIATED TO 4.3 x 1080 pissions /cm AT A LINEAR HEAT RATING OF 868 w/CM. R-23747 TOUTES DY Y . A . Cen: - . - . . ri. - - . San A NITROGEN FIRED FUEL IRRADIATED TO 4.8 * 100 fisslons /ono AT A LINEAR HEAT RATING OF 910 w/cm. . S OINI - AEC - OFFICIAL - .-- - - --. mym. . . . . • . -- ; Y-66378 ORNL DWG 65-9381 1.20 sisa--- · ORNL - AEC - OFFICIAL in .. . .. 2337233 + 235 235 Th, U, and Fission Products 7233 = 2.29 ist s 7235 = 2.08 1.10 = 2.00 . : . * = 1.00 x o Teff. ; * = 6.00 x 1013 : 1 . 1$ = 3.00 x 1013 . WILL ..! $ = 1.00 x 1014 C -0.80 E - i 103 UILULUI 104 . ": ' 105 229 BURN-UP MWD/MT TH-U233 ORNL - AEC - OFFICIAL Effect of Exposure on the Reactivity of an (0.984 Th - 0.016 233 U) 0, Fuel. : : !' . ------ Niin. cra. . , -- -".. . . . - op..*-.. : -- .. . : 100r -. Y-66377 ORNL DWG 65-9382 TTTTTTTTTT Burn up Capsule Material (Mwd/T Th & U Charged) Calculated Measured Z-5 Sol-Gel E (TH-3.96% U) OZ 14,000 ETR-I Sol-Gel 35 (Th-502% U! O2 22,000 Z-? Sol-Gel E (TH-3.96% U) 02 40,000 U-3 Arc-Fused (Th-3.96% U) O2 71,000 U 236 2-8 Sol-Gel E (Th-3.96% U) O2 81,000 Pellet (Th-3.92% U) O2 120,000 No. U233 1234 U235 712 U233 O ISOTOPIC ANALYSIS (%) --- -- - . - -- 1,235 .. OLZETOLIILI IT 0 10 20 30 40 50 60 70 80 90 100 110 120 x 100 BURN UP (Mwd/T Th and U) : .. . Change in Uranium Isotopic Concentration as o Function of irradiation Level... minee ** .. r 1.3 . ORNI - AEC - OFFICIAL Titi******** C. . 2 i Tryginim termine $ . . . . - --- ...... .... ' . ' END DATE FILMED 12/23/65 ------* -... -rine .... . ... - . - - --- - -- - - - - .. . . - -