NOV 23 1971 __ {3 o g 2 _ ..-..dfaf M ik A < ey Fom OAK RIDGE NATIONAL LABORATORY operated by _ UNION CARBIDE CORPORATION * NUCLEAR DIVISION for the U.S. ATOMIC ENERGY COMMISSION \ Ve ORNL- TM- 358"~ } THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED BY THE ENRICHED URANIUM, Pu=238U, AND 233y—_Th FUEL CYCLES M. J. Bell and R, S, Dillon NOTICE This document contains information of a preliminary nature and was prepared primorily for internal use at the Ock Ridge Notional Loboratery. It is subject to revision or correction and therefore does not represent a final report. T S — m———— - . —— R e — .. e e — el - e — This report was prepared as an account of work sponsored by the United States Gowvernment. Neither the United States nor the United States Atomic Energy Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights. s - \ < G o . \.‘J ¥ = - A & R o MOV ,g'f 9 4d A OAK RIDGE ‘NATIONAL I.ABORATORY operated by UNION CARBIDE CORPORATION NUCLEAR DIVISION for the U.S. ATOMIC ENERGY COMMISSION ORNL- TM- ' THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED BY THE ENRICHED URANIUM, Pu=238U, AND 233y-Th FUEL CYCLES M. J. Bell and R, S, Dillon NOTICE This document contains information of a preliminary nature and was prepared primarily for internal use ot the Oak Ridge National Laboratory. It is subject to revision or correction and therefore does not represent a final report. 3548" R 5 e . e e w5 R . : a i o 1 S o “ € : v ORNL-TM-3548 Contract No. W-T4OS5-eng-26 CHEMICAL TECHNOLOGY DIVISION THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED RBY THE ENRICHED URANIUM, Pu-238(, AND 233U-Th FUEL CYCLES M. J. Bell and R. S. Dillon™ NOVEMBER 1971 *Present address: AECOP, Oak Ridge, Tennessee. OAK RIDGE NATIONAL .LABORATORY Oak Ridge, Tennessee 37830 operated by UNION CARBIDE CORPORATION f for the U.S. ATOMIC ENERGY COMMISSION iii CONTENTS Abstract . . . ¢ f e e e e e e e e e e e e e e 1. Introduction . . . . . . « « + 4 . . 2. Reactor Operating Conditions and High-Level- Waste Radionuclide Inventories . . . . . . . . 3. Hazard as a PFunction of Age of High-Ievel Waste 4. Long-Term Relative Hazards of High-Level Waste . 5. Isotopic Dilution of High-Level Wastes . . . . 6. Relative Hazard of Transuranium Wastes . . . . 7. References . . . . ApPendiX « v . v w e e e e e e e e e e e e e e e 11 1k 16 18 19 v THE LONG-TERM HAZARD OF RADIOACTIVE WASTES PRODUCED BY ’ THE ENRICHED URANIUM, Pu-<3%U, AND <=33U-Th FUEL CYCLES M. J. Bell and R. S. Dillon ABSTRACT An evaluation has been made of the long-term hazards of ~ "high-level"” and "alpha" radioactive wastes generated by irra- diation of three fuels representative of those which will be used to generate electrical powér in the next several decades. In this evaluation the composition and radioactivity of the wastes generated by typical LWR, IMFBR, and molten-salt breeder “reactors utilizing the enriched uranium, Pu-228U, and 233U-Th fuel cycles,respectively, have been computed for times up to 30 million years after discharge. | . The volumesof water necessary to dilute the various types of high-level waste to the radiation concentration guides (RCG) for - ingestion in unrestricted areas have been computed as a function of age of the waste. It was found that the volumes required to dilute the three types of waste were similar when evaluated at the same age and exposure. The volume of water necessary to dilute any of the three wastes aged 1000 years and the associated salt in the proposed Federal Repository to the RCG for unrestricted uses is less than that required to dilute the ‘same amount of uranium ore tailings to the RCG. The high-level waste deposited in the Federal Repository will result in an alpha activity of less than 10 uCi/kg-at 10,000 years after burial when averaged over the total mass of the Repository. . The hazard of a number of types of waste contaminated to 10 uCi/kg (the upper limit assumed for surface burial) of initial parent alpha activity was also calculated as a function of time, - and the time of maximum relative hazard was determined, Trans- uranium or alpha wastes contaminated to -this level will present a maximum ingestion hazard similar to naturally occurring uranium ores. 1. INTRODUCTION ‘High-level and alpha wastes generated in nuclear fuel cycles for . production of electrical power will contain.isotopes which remain radio- active for miiiions,of years. The "high-level" wastes are principally the fission-pré@uét congentrates that arise from the recovery of fissile and fertile materials from.spent fuel. Typically, however, these wastes contain a variety of actinides that are made from transmutation of fuel material and, in addition, quantities of uranium, thorium, and plutonium that are not economically recoverable for recycle., "Alpha'" wastes — materials contaminated with substantial concentrations of long-lived alpha emitters — are produced primarily in plants for preparation of nuclear fuel materials. Sighificant variations will occur in the compositions of the wastes generated by various reactor concepts because of differences between types of fuel, neutron energy spectra and flux level, and length of irradiation, as well as'efficiéncy of utilization of fuel. Calculations have been made, as a function of decay time, of the compositions and radiocactivities of the wastes generated by typical light water reactors {(LWRs), liquid metal cooled fast breeder reactors (IMFBRs), and molten-salt breeder reactors (MSBRs), which are representative of enriched 237U, 229py, and 233U fuels, These radicactivities have been used,along with the radiation concentration guides for ingestion in unrestricted areas given in the Code of Federal Regulations (10 CFR 20), to evaluate the relative hazard of the wastes. The compositions of 460 fission products and 80 actinide elements and their decay products were included in the calculations, which were per- formed with the ORIGEN isotope generation and decay code.l’2 The authors wish to acknowledge the assistance of H. C. Claiborne, J. 0. Blomeke, and J, P. Nichols in reviewing this report and for their helpful suggestions in the course of the work, 2. REACTOR OPERATING CONDITIONS AND HIGH-LEVEL-WASTE - 'RADIONUCLIDE INVENTORIES The quantities present at time of processing in wastes resulting from 33,000 MWA of exposure in typical LWRs, LMFBRs,and MSBRs are given in Table 1 for those nuclides found to be of importance in the evaluation of the long-term safe disposal of radioactive wastes. The LWR considered has been described in ref., 3. It is fueled with 3.3% enriched uranium and is operated at an average specific power of 30 Mwymetric ton of heavy metal charged to the reactor to a burnup of 33,000 MWd/metric ton. The fuel is assumed to be processed at 150 days after discharge, with removal of 99.5% of the uranium and plutonium for recycle, Table. 1. Quantities (Kilograms) of- Long-Lived Hazardous Nuclides Present at Time of Processing for 33,000 MWd of Exposure in Reactors Utilizing the Enrlched Uranlum, Pu-2387, and 233U-Th Fuel Cycles Isotope Half-Life Enriched Uranium® P 2337.Th® % Sr 27.7 v - 0.5k - 0.302 0,918 1297 1.7 x 107 y - 0.231 o o.2l1 0.35 232Th 1.41 x 1010 y - . - 24L 233pg T .27.0 4 o o 0.0042 233y 1.62'x 105 y 0.0167 2347 2.47 x 105 y | 1.15 x 1075: 1.36 x 10-4 0.0041 235y 7.1 x 108 y 0.0k ‘ 0.00745 0.0011 2367 2.39 x 107 y 10,0226 - 5,52 x 1074 0.0011 2387 4.51 x10° y - 4,72 4.38 237Np 2.14 x 10F 'y 0.483 0.128 -~ 0.0317 238py . 864y 8.4 x 1074 0.00534 ' 0.303 . 239py - 24,390 y "~ 0.0265 0.288 © 0.0037 240 py 6580 y 0.0107 0.0997 1.1 x 10-% 241py 13.2 y 0.0050 0.026 - - 5.67 x 10~ 242py 3.79 x 105 0.0017 0.016 " | 2alpm - L33y 0. 0446 0.472 2438y 7950 v 0.0925 0.2h4g 2420 2163 4 - 'bf00582 0.0188 244Cn - 18.1y 0.0278 0.0145 3 3% enriched uranium irradiated at an average specific power of 30 MW/metric ton to a burnup of 33,000 MWd/metric ton in a typlcal PWR. Proce331ng at 150 days after discharge. bLWR discharge Pu and depleted U irradiated in core and blanket of AT Reference Oxide LMFBR at an.average specific power of 58 MW/metrlc ton to a burnup of 33, OOO MWd/metrlc ton. Processing at 30 days after discharge. Reference MSBR equilibrium fuel cycle with continuous chemical processing by fluorination-reductive extraction and the metal transfer process. The LMFBR considered was the Atomics International Reference Oxide Design.:‘}'5 The mixed fuel discharged from the core and blankets of this reactor has been irradiated to an average burnup of 33,000 MWd/metric ton of heavy metal charged to the reactor at an average specific power of 58,2 MW/metric ton. The fuel is assumed to be processed at 30 days after dis- charge with a 99.5% recovery of uranium and plutonium, The MSBR is a fluid fuel reactor that operates on the Th-233U fuel cycle.6 The present concept employs fluorination-reductive extracfion of the fuel salt to isolate 233Pa outside the reactor core with a 10-day removal time, This chemical processing step is also responsible for removing plutonium and a number of fission products from the fuel salt on & 10-day cycle. Strontium, barium, and the rare-earth fission products are removed from the fuel salt by an extraction process called the metal transfer process with removal times ranging from 16 to 51 days. In addi- tion, salt containing thorium is discarded to waéte on a hZOO—day cycle. This mode of operation, which results in thorium utilization of only - 13.7%, makes fairly inefficient use of fertile material relative to the LWR and IMFBR concepts. The reference MSER has a yield of 3.2% of the - reactor fissile inventory per year, and it was assumed that 1/2% of the uraniufi removed from the reactor for sale was lost to waste. Also, high- level wastes are removed from the system in batches every 220 days follow- ing fluorination to recover uranium which might be present in the waste streams. It was assumed that 1/2% of the uranium in the waste streams was not recovered by the fluorination, and that all the 233Pg which remained at the end of the 220 days was lost with the fission product waste, It should be noted that the LMFBR and MSBR are advanced concepts with thermal efficiencies of 40O to L5%, while typical PWRs achieve thermal efficiencies of about 32%. 'Consequently, an LWR generating the same electrical energy as an LMFBR or an MSER will produce 25 to 40% more waste than that given in Table 1 or in subsequent tables since they are all based on 33,000 MWd of heat production. 3. HAZARD AS A FUNCTION OF AGE OF HIGH-LEVEL WASTE The composition, radioactivity, and hazard of the three types of waste were computéd.as a function of age for tifies up to 30 million years. Tables 2-L4 show the radioactivity of a number of isotopes of special interest as a function of age. In these tables, the actinideé are grouped according to their decay chains, so that it is easier to observe daughters building up and gradually reaching equilibrium as ~ their precursors decay. Table 5 presents the values used for the RCG for the isotopes found to be most important in determining the long-term hazard. A measure of the ingestion hazard associated with a radionuclide is the quantity of water required to dilute the nuclide to the RCG for unrestricted use of the water; the larger the amount of Wéter required, the greater the hazard. - The volumes of water required for the three types of wastes are shown in Table 6,and the isotope'which is the principal hazard at a given time is shown in parentheses. The fission product 2°Sr is the principal ingestion hazard for the first few hundred years and the hazards are about the same for the three types of wastes. In the first 30 to 300 years after disposal, the measure of hazard associated with the wastes decreased from around 10! cubic meters of water to around 3 x 108 cubiC'meters,due primarily to thé decay of P05y, For the period 300 to 3000 years after disposal,the 222U fuel waste is somewhat less hazardous than the others due to the absence of transplutonium isotopes. The hazard associated with all the Wasfes rises slightly at about a quarter of a million years after disposal, which is due to peaking of the 22€Ra activity. The hazard associated with the 233U fuel wastes diminiéhes less quickly than the other wastes due to the presence of the relatively large amount of *32Th, the parent of the isotope 228Ra{which is the predéminant hazard in this waste after 10 years). Table 2. Typical Radiocactivity (Curies) of Long-Term Hazardous Nuclides in Waste from a Thermal Reactor Fueled with Enriched Uranium as a Function of Age for 33,000 MWd Exposure A - Age of Waste (years) Nuclide - 107 108 : 104 105 108 107 90 Sr 6480 1.5 x 1078 | 1297 0.038 " 0.038 0.038 0.038 0.036 0.025 241 A 145 o 3k | 1.9 x 10°% 243 A - 17.6 16.3 T2 . 2.09 x 1072 239py 1.68 | 2.06 - 4.06 0.57 234y 0.022 0.0h2 _ 0.041 , 0.032 L.0 x 1073 1.6 x 1073 226 Ry, 1.7 x 1077 © 5.3 x 1078 2.65 x 1078 0.021 | 5.2 x 1078 2.3 x 10-8 Total B~ 33,900 37 27.1 15.5 k.o 0.13 curies ' . _ . Total Q . 276 61 14.9 | 2.2 - 0.15 curies Table 3. Typical Radioactivity (Curies) of Long-Term Hazardous Nuclides in Waste from a Fast Reactor Fueled with Pu and 238U as a Function of Age for 33,000 MWd Exposure Age of Waste (years) Nuclide ' 10° 10° - 10t ' 108 | 108 107 90 5y 3.62 x 10® 8.27 x 1077 1297 - 0,035 0.035 ~ 0.035 0.03k4 0.033 0.023 341y 1.39 x 102 356 4.33 x 1074 243 ppy W74 L3.7 19.3 5.62 x 1072 239py 17.8 18.5 21.0 2.31 2347 0. 0847 ~0.187 0.183 0.1k2 0.0127 0.0015 226Rg 6.29 x 1077 - 2.22 x 107 1.17 x 1072 -0.0931 0.0181 0.0021 Total B~ 31,400 68.5 b2, 2 a7 4.2 0.29 curies _ ' Total 1760 Lu6 50.2 " 5.1 3.2 0.17 curies Table L. Typical Radioactivity (Curies) of Long-Term Hazardous Nuclides in Waste from a Thermal Reactor Fueled with 233U and Th as a Function of Age for 33,000 MWd of Operation Age of Waste (years) Nuclide 102 10° 10% - 108 10° 107 90 gy 11,000 2.5 x 106 1297 0.057 0.057 0.057 0.057 0.055 231pg 0.661 0.649 0.535 0.078 1.01 x 1076 1.00 x 105 223Rp, 0.635 0.649 0.535 0.078 1.01 x 1078 1.00 x 1076 2347 1.02 1.87 1.82 1.4 0.113 3 x 10711 226Rg .06 x 105 .50 x 1073 0.118 0.926 0.167 3 x 10712 232Th 0.0266 0.0266 0.0266 0.0266 0.0266 0.0266 228Rg, 0.0266 0.0266 0.0266 0.0266 0.0266 0.0266 237Np 0.022 0.022 0.022 0.022 0.016 8.8 x 107¢ 2337 0.199 0.198 0.190 0.129 0.0028 8.8 x 1074 229Th - 0.039 0.050 0.132 .. 0.136 0.0029 8.9 x 10-4 228 Ra 0.037 0.050 0.132 0.136 0.0029 8.9 x 107 Total B L4 400 17.7 16.7 13.2 L.h 0.2k curies . , _ . Total o 2360 11 9.6 12.9 2.1 0.22 curies 9 - Table 5.HfiThe;RCGs-Usedgin‘Evaluating.the Ingestioanazards from Radionuclides (from 10 CFR:20, Table II, Column:2) RCG Nuclide Half-Life - (ci/m?) 90 gy : 27.7 v 3 x 10-7 1297 1.7 x 107 y 6 x 1078 223Rg, | 11.4 g 7 x 1077 226Rg - 14.8 a 6 x 10~7 226 Ry, - 1620 @ 3 x 1078 22°Ra | 6.7y 3 x 1078 229 7340 ¥ 2 x 107® 230 Th | | 8 x 10% y 2 x 10-® 2321 1.41 x 101° y 2 x 10-® 231pg - 3.25 x 10% vy 9 x 1077 233pg - 27.0 d 1 x 107% 233y 1.62 x 105 y 3 x 107® 2347 2.47 x 108 y '3 x 10°° 2357 7.1x 108y 3 x 107 R367y 2.39 x 107 y 3 x 107° 238y 4,51 x 10° y 4L x 10°° 237 Iip 2.14 x 10° y 3 x 107° 238 py 86.4 v 5 x 1078 239y - | 24,390 y 5 x 1078 240 pyy - 6580 y 5 x 107 241py - 13.2 y .2 x 1074 242py 3.79 x 108 y 5 x 1078 24lpm 433 y 4 x 107¢ 2437y 163 d h x 107® 10 Table 6. Volume of Hz0 (Cubic Meters) Required to Dilute Wastes Resulting from 33,000 MWd of Exposure of Enriched 235U, 23%py, and 223U Fuels to Levels Permitted for Unrestricted Use (RCG from 10 CFR 20, Table II, Column 2) Age of Waste (years) 2362 239y P 2337C 30 1.26 x 10™ (®8r) 7.22 x 10%° (%§r) 2.11 x 101 (%0%5r) 100 2.24 x 1019 (®°gr) 1.32 x 101° (90gr) 3.75 x 10%° (®03r) 300 © T 2.00 x 108 (°°g8r) 3.93 x 108 (241an) 3.76 x 108 (%°3r) 1,000 1.55 x 107 (241Am) 1.09 x 108 (241pm) L.72 x 108 (223Ra | and 228Ra) 3,000 6.53 x 10° (242Am) 2.2h x 107 (243pm) 5.06 x 10 (223Ra and 228Ra) 10,000 h.26 x 108 (239Ppy) 1.26 x 107 (243Am) 9.34 x 10F (226Ra) 30,000 2.4 x 10° (239py) 6.88 x 10f (239py) 2.2 x 107 (226Ra) 100,000 2.14 x 108 (2°%Ra) 5}7h'x 10° (226Ra) 4,45 x 107 (228Ra) 300,000 2.38 x 10° (226Ra) 5.81 x 10° (226Rg) .68 x 107 (225Ra) 1,000,000 1.58 x 10f (1291) 2.03 x 10° (1297) 9.42 x 1CF (226Ra) 3,000,000 9.93 x 10° (2°1) 9.53 x 105 (12°71) 1.80 x 10° (228Ra) 10,000,000 5.28 x 105 (12°1) 4.88 x 105 (129T) 1.56 x 1CF (228Ra) 30,000,000 2.57 x 105 (297) 2 ¢ ] 1.20 x 10F (22°Ra) .38 x 105 (1291) 2.0 x 108 = volume of water (cubic meters) required to reduce ingestion hazard of the corresponding amount of uranium ore (7900 tons of ore contain- ing 0.17% U). volume of water (cubic meters) that results from dissolving the salt required to store waste equivalent to 33,000 MWd exposure to a termi- nal concentration of 500 ppmn. 1.1 x 107 1l 1.07 x lO6 = gpproximate volume of water (cubic meters) required to reduce ingestion hazard potential of the corresponding amount of earth containing naturally occurring U + Th in equilibrium with their daughters at the average concentration in the earth's crust. aReference PWR fueled with 3.3% enriched U, operated at a specific power of 30 MW]metrlc ton of heavy metal charged to reactor. Processing losses of 1/2% of U and Pu to waste are assumed. bAI Reference Oxide IMFBR mixed core and blankets fueled with LWR discharge Pu and diffusion plant tails. Average specific power of blend is 58.2 MW/metric ton of heavy metal charged to reactor. Processing losses of 1/2% of U and Pu to waste are assumed. - CReference MSBR with continuous protactinium isclation on a 10-day cycle and rare- earth removal by the metal transfer process. Thorium is discarded on a L4200-day cycle, and 1/2% of excess uranium is assumed to be lost to waste (see text). 11 Lk, LONG-TERM RELATIVE HAZARD OF HIGH-LEVEL WASTE The isotopes ®2°Ra and 228Ra are the predominant alpha-emitting radionuclides for ages greater than 100,000 years., These isotopes occur in nature as daughters of 238U and 222Th and,. if they are sufficiently dilute in the waste, will present a hazard similar to those of naturally occurring uranium. and thorium deposits. In the proposed Federal Reposi- tory in bedded_salt,.the waste equivalent to 33,000 MWA of exposure may be assumed to be associated with about 5500 metric tons of salt and 2400 tons of .shale, This is based on the current design of the high-level facility of the Repository, assuming that the accumulated waste from 3.2 x 10° MWd(t)* of exposure is dispersed in the 900-acre by 300-ft- thick section of the bedded salt. ©Since the LWR or IMFBR waste from 33,000 MWd(t) of exposure will have a volume and a mass of about 3.3 ft® and 140 kg, resPectlvely, the salt bed serves to dilute the wastes by approximately a factor of 37,000 in volume and 56,000 in weight. (The dilution would not be asAgreat for the present MSBR concept because of the larger qfiantity of waste generated per 33,000 MWd of operation. ) This dilution results in an average R387 concentration in the bedded layer of salt ahd‘shale of less than 1 ppm by weight for the 2357 and 239py wastes, and an average 232Th concentration of 30 ppm by weight for the 2337 fuel wastes The amount of 934U which is present in the waste 1n1t1ally, plus that Wthh is produced by decay of 242Cm and 238Ppy, is well above that whlch would be in equlllbrlum with 1 ppm of naturally occurring 238U, The quantltles of 2247 which ultlmately occur in the wastes correspond to those which would be in equilibrium with 15, 5k, and 680 wt ppm 238U in the bedded layer of salt and shale for the wastes from 235U, 23°%Py, and 2"3"3U fuels, respectively. The average concentratlons of uranium and thorium in the earth's crust are 6 and 12 ppm, respectlvely, and the volume of water requlred to reduce the ingestion hazard of the corre3pond1ng mass of earth (7900 metric tons) containing these concentrations of uranium and thorium in * _ The most recent forecast7 estimates that the high-level waste to be accumulated in the Federal Repository, up to the time of closing in year 2000, is 319,000 ft?,Whieh is equivalent to a total exposure of 3,19 x 10° MWd(t). Most of this waste will be from LWRs since most of the IMFBR waste will not yet have been delivered to the Repository. 12 equilibrium with their daughters is 1.07 x 10° m® (see Appendix). The " hazard associated with the LWR type waste (the predominant type in the ‘Repository) is only a factor of four higher 10,000 years after disposal, and the hazards from the other wastes are about an order of magnitude higher. However, the hazard associated with the 233U fuel waste rises again as the 226Ra concentration increases. The presence of 226Ra and 228Rg in high concentrations in the waste from the 232U-Th system is ‘the result of the absence of a step for recycling plutonium and the assumed inefficient use of thorium in the present reprocessing schemes. If the processing flowsheets are improved to eliminate these two problems, the 233U fuel wastes would remain very similar to the 235U and 22°9Pu wastes at ages greater than 10,000 years. Other thermal reactor concepts employing the Th-233U fuel cycle are the ngh Temperature Gas Cooled Reactor (HTGR) and the Light Water Breeder Reactor (LWBR). The wastes produced by these reactors would be similar to MSBR wastes since the higher 9°Sr’concentra.tions Would'be present initially, since 238Pu will not be separated from the waste and recycled, and because efficient recycle of the relatively inexpensive thorium is not envisioned.at present, Wastes from 228U fuels aged 30 years have a thermal'power 64% higher than 23°Pu wastes of the same burnup, and wastes aged 100 years have a thermal power h6% greater Since the peak tempera- tures in the mine are reached at about 50 years after burlal, the hlgher heat generation rate from ®33U fuel wastes will require a greater spacing between waste cans, with a subsequent increase in cost. The burial cost of the waste is a small fraction of the fuel cycle cost, however. An alternative method of assessing the relative hazard of the radio- active wastes is to compare them with naturally occurring radiocactive surface deposits. The southwestern United States has uranium ore reserves containing 0.2% U305 or greater, totaling almost 50 million tons. In the year 1968, domestic production of U,04 totaled over 12,000 tons, indicating that some 6 million tons of such ore were processed and & corresponding amount of uranium ore'tailings ‘was disposed of.‘ The volume of water required to dilute the radionuclides contained in 7900 metric tons of uranium ore tailings to the RCG for unrestricted use of the water is 13 2.0 x 108 m®, Consequently, the ingestion hazard associated with a quantity of uranium ore and uranium ore tailings equal to the amount of salt and shale asseciaied With 33,000 MW4 of Waste‘is greater than that associated with any of the three high-level wastes aged 1000 years. Another method of assessing tne relative hazard of a radioaetive waste is to compare it:with the hazard of naturally radioactive monazite sand. Thorlum and uranium are present 1n the mlneral monazite which occurs in beach sand in some areas of India, Brazil, Malaysia, and the southeastern United States.8 Beach sands in Malagasy Republic average 2 to 2, 5% monaz1te, the monazite containing 8.8% ThOz and 0.41% Us0g. Monazite . sands found on the southeast coast of the United States average 3.1% Th024and O.H?%,Uéoa. To reduce the ingestion hazard from radioactivity ifi these two sands for a mass‘equivalent to the waste and associated salt in the prOposed"Federal.Repository will require about 6 x 107 and 2 x 107 m?,of wafer for the Malagasy and United States sands, respectively. Henee, dissolving the buried 235U or 233U fuel wastes aged in the range of 1000 to 3000 years, along with the associated salt, would present a hazard similar te.dissolving the radium from the same mass of monazite- contalnlng beach sand (noting, of course, that monazite sand is among the least soluble of naturally occurring materials). Another consideration with respect to the disposal_of.high-ievel radioactive wastes in bedded salt is that to dissolve the naste for a 33,000 MWd exposure and associated salt to a potable concentration of 500 ppm NaCl by weight would require 1.1 x 107 m® of water. Thus, 235U and 233U fuel wastes aged 3000 years, if diluted to drinking water con- centration in NaCl, would be below the radiation concentration guide for ingestion by the general population. . Hence, if the salt mine is dissolved some thousands of years in the future with sufficient water to dilute the radionuclides to the RCG and the dissolved materials find their way to: drlnklng Water supplles, the water would be unacceptable as drinking _ water because of 1ts sodlum chlorlde content, 1k 5. .ISOTOPIC DILUTION OF HIGH-LEVEL WASTES It'has been suggested that a radioactive isotope may be dispoeed'of without further dilution provided that it is diluted to an acceptable specific activity with an adequate quantity of the stable element in the same chemical form and if the diluted material is also acceptable on the basis of external hazard and chemical toxicity.9 Consideration has been given to this dilution method for'reducihg the ingestion hazard associated with the high-level wastes. Table 7 lists the isotopes which present the greatest hazard at a given age of the #aste; their radicactivities, and the quantities of stable elements which would be required to dilute the wastes below the maximum permissible specific activities given‘in ref. 9. In ref., 9 it is assumed that the isotope 9°Sr must be diluted with stable strontium, rather than another bone-seeking element (sueh as Ca), but that the elements americium and plutonium may be diluted with rarée earths. It is also assumed that an "acceptably low" level of specific activity is the vaiue such that, if a person were to assimilate the element (or chemically similar species) only from the source of interest, the body burden of the radioisotope would not exceed the maximum permissible body burden for occupational exposure. USing these assumptions it can be calculated that 432 metric tons of stable strontium are required to dilute the 2°8r in 100-year-old 235U waste resulting from 33,000 MWd of exposure to an acceptably low specific activity. Since each 33,000 MWd of waste is associated with 8000 metric tons of salt and shale, a concentration of 5% by weight stable strontium would be required throughout the salt mine, which would be impractical. The concept may prove feasible for . diluting waste aged 300 years, since only 3 metric tons of stable strontium would be required for each 33,000 MWd exposure. If it is assumed that the behavior of americium and plutonium in the body is similar to that of rare earths, then 49 and 5.7 metric tons, respectively, of rare-earth elements are required to be added to the wastes to eliminate the ingestion hazards from 241Am and 239Py, Also, if 250 g of stable iodine is added to each 33,000 MWA of waste, the ingestion hazard from 12°I becomes acceptable for occupational exposure. Table 7. Mass of Stable Carrier Required to Dilute Radioisotopes in 238U Fuel High- Level Waste to Maximum Permissible Specific Activity in the Environment Mass of Carrier _ , - Maximuma Required to Dilute Age Isotope Radioactivity Carrier Critical Permissible to Specific Activity (years) (Ci/33,000 MWd) . Organ Specific Activity (metric tons) (Ci/metric ton) 33,000 MWd Exposure 100 90 gy 6480 Sr Bone 15 432 300 ogr W67 Sr " Bone | 15 o 3.1 1,000 241pm o 34.h rare earth Bone 0.7 L9 10,000 239py k406 rare earth Bone - 0.71 5.7 100,000 22§Ra 0.021 . b Bone b | ' - 1,000,000 1291 0.036 T Thyroid 80° ' 4.5 x 1074 *National Academy of Sciences-National Research Council, Publication 985, Disposal of Low-Level Radioactive Waste into Pacific Coastal Waters, Washington, D. C., 1967. bInadequate biological data to epply the specific activity concept to 226Ra. - “Calculated from data given in International Commission on Radiation Protection, Report of Committee II on Permissible Dose for Internal Radiation (1959), Pergamon Press, Oxford, 1960. ST 16 6. RELATIVE HAZARD OF TRANSURANIUM WASTES For disposal in the Federal Repository, Godbee afid,Niéholle have proposed a lower limit for transuranium or "alpha":waste of 10 uCi/kg of initial parent alpha activity. Alpha wastes (which mostly arise from fuel fabrication) of lower activity are to be disposed of by surface burial.‘ To evaluate the hazards associated with surface burial of alpha wastes,an investigation was made of the relative ingestion and inhalation hazards of a nufimer of isotopes and mixtures of isotopes expected to be present. Table. 8 compares the radioactive properties of uranium-and thorium in the average earth's crust, uranium ore, and‘firanium ore tailings with those of a number of transuranifim wastes contaminated to an initial parent alpha activity of 10 uCi/kg. Shown in the table -are the properties at the time of isolation of the paren£s~and at the time of maximum ingestion hazard;' At the time of maximum hazard, the wastes present an ingestion hazard similar to uranium ore tailings althofigh their total activities’, thermal powers, and inhalation hazards are about an order of magnitude greater than uranium ore tailings. The maximum ingestion hazgrd resulting from an initial parent alpha activity of 10 uCi/kg of either natural uranium or natural thorium is larger by an order of magnitude than any of the other wastes considered. For some transuranium materials, 238Pu, 242Cm, and *44Cm, the maximum ingestion hazard is associated with the initial 10 uCi/kg of parent alpha activity. The ingestion hazard associated with low-level plutonium wastes contami- nated to a specific activity of 10 uCi/kg is appreciably less than that associated with a kilogram of uranium ore tailings, and the ingestion hazard associated with plutonium wastes is comparable to the maximum hazard associated with other alpha wastes contaminated to the same initial specific activity. The ingestion hazards for various wastes are sufficiently similar that a criterion based on total alpha activity alone, regardless of isotopic composition, may be sufficient for classi- fying transuranium or alpha wastes. - . Table 8. Comparison of Radioactive Properties of Naturally Occurring Thorium and Uranium Deposits with the Assumed Lower Limit for Classification as Transuranium Waste (10 pCi/kg) Total Activity Cubic Meters of Water or Air at RCG® Alpha Activity Time Since /kg) Thermal . : Source of Parents Isolation of (1Ci/ke Power ~ E=C Kilogram Dissolved or Suspended (uCi/kg) Parents (y) Beta Alpha (nW/ke) Water® AirP Average Earth Crust® . U 0.0021 4.5 x 10° 0.012 0.016 0.00058 0.091 (22€Ra) 31,300 (23°Th) Th 0.0013 | k.5 x 10° 0.0052 0. 0078 0.00030 0.045 (228Ra) 9,250 (228Th) Total 0.0034 L5 x 10° 0.017 0.02k4 0.00088 0.136 (22%Ra) 41,000 (23°Th) Uranium Ored 0.59 L.5 x 10° 3.5 b7 0.16 25.6 (226Ra) 8.8 x 108 (230Th) Uranium Tailings® 0.029 4.5 x 10° 3.5 4.1 0.15- 25.6 (22%Ra) = 8.6 x 108 (=2307y) Natural U - 10 0 0 10.0 0.27 0.292 (234y) 2.90 x 10° (238y) Natural U 10 1P 57.1 76.2 2.76 433 (226Ra) 1.5 x 108 (239Th) Natural Th 10 o . 0 . 10.0 0.24 5.0 (232Th) 1 x 107 (232Th) Natural Th 10 25 Lo.o" 60.0 2.31 °© 346 (228Ra) 7.1 x 107 (228Th) TWR pul - 10 0 470 10 0.34 | L4 (241py) 3.1 x 108 (241py) LWR Pu 8.6 308 114 20.4 0.68 - 5.23 {(241Anm) 2.3 x 10° =238py) 238Ppy 10 0 0 10 - 0.33 . 2.0 (®38py) 1.4 x 10® (238epy) 236 py 10 0 0 10 0.35 h h 236 py 0.0068 308 0.63 1.89 . 0.075 0.23 (22%Ra) 1.9 x 10 (228Tnh) 2420y 10 0 0 10 0.37: 0.498 (242Cm) 2.5 x 10F (242Cnm) 2440 10 o - 0 10 0.35 0.143 (244Cm) 3.3 x 107 (244Cm) 2327 10 0 0 10 0.32 0.33 (232u) 1.1 x 107 (232y) 282y 9.1 108 18.1 54, 3 2.1k 6.6 (224Ra) 5.6 x 107 (228Th) 233y 10 0 0 10 0.29 0.333 (=22U) 2.5 x 10F (233y) 233y 8.8 3 x 1048 25.7 51.8 1.96. - 18.9 (225Ra) . 1.1 x 108 (229Th) 93% 238y 10 0 0 10 0.27 0.330 (238y) 2.5 x 10F (235y) 93% 236y 10 - 1088 uo.p 69.8 2.6 36 (223Ra) - 3.8 x 108 (231pa) 237Np 10 0 o - .10 0.29 © 3.33 (227Np) 1 x 108 (237Np) 237Np 9.1 3 x 1058 29.8 50. 3 1.86 18.3 (225Ra) 1.8 x 10% (237Np) %Based on radiation concentration guides for continuous exposure in unrestricted areas (10CFR20, Table II). bThe principal contributor to the hazard is shown in parentheses. | “Assumes U and Th concentrations of 6 and 12 ppm, respectively. dassumes ore containing 0.17% U. . _ “Uranium ore tailings are assumed to have 5% of the original uranium but all of the radioactive daughters. Tpu is assumed to have isotopic composition of 1% 238Pu, 59% 239pu, 249, 240py, 129 241Py, and 4% 242py, gTime since isolation of parents at which maximum ingestion hazard occurs. | hRCG values are not defined for 238py, LT 10. 18 7. REFERENCES D. E. Ferguson et al., Chem, Technol. Div. Ann. Progr. Rept. May 31, 1969, ORNL-4422, p. 89. M. J. Bell, ORIGEN - The ORNL Isotope Generatlon and Decay Code, ORNL.-4628 (1n preparation). M. J. Bell, Heavy Element Composition of Spent Power Reactor Fuels, ORNL-~TM- 2897 (May 1970). K. Buttrey, O. R. Hillig, P. M. Magee, and E. H. Ottewitte, Liquid Metal Fast Breeder Reactor - Task Force Fuel Cycle Study, NAA A-SR- MEMO-12604 (January 1968 ). :Staff of the Chemical Technology Division of ORWNL, Agueous Processing of IMFBR Fuels - Technical Assessment and Experimental Program Defi- nition, ORNL-4436 (June 1970). M. W. Rosenthal et al., MSBR Program Semiann. Progr. Rept., February 28, 1971, ORNL-U4BT6. F. L. Culler, J. O. Blomeke, and W. G. Belter, "Current Developments in Long-Term Radioactive Waste Management," paper A/CONS/L9/P839, presented at the Fourth United Nations Conference on the Peaceful Uses of Atomic Energy in Geneva, Switzerland, Sept. 6-16, 1971. U.S. Bureau of Mines, Bulletin 630, Mineral Facts and Problems, 1965 ed., "Thorium," pp. 947-959. National Academy of Sciences-National Research Council, Publication 985, Disposal of Low Level Radioactive Waste into Pa01f1c Coastal Waters, Washington, D. C., 1962, H. W. Godbee and J. P. Nichols, Sources of Transuranium Solid Waste and Their Influence on the Proposed National Radioactive Waste Repository, ORNL-TM-3277 (January 1971) (Official Use Only). APPENDIX RADIOCACTIVITY IN EQUILIBRIUM WITH NATURALLY OCCURRING URANIUM AND THORIUM AND THE ASSOCIATED INGESTION AND INHALATION HAZARD (Cubic Meters) Potentially Contaminated to RCG by Radiocactive 20 Table A.l1. Radiocactivity and Volumes of Air and Water Daughters in Equilibrium with 1.0 g of 232Th Radiocactivity Ingestion Hazard Inhalation Hazard Nuclide (uCi) (cubic meters of water) (cubic meters of air) L 0.109 ' 0.0547 1.09 x 108 228Rg, 0.109 - 3.65 1.09 x 108 228 pc 0.109 1.22 x 1073 1.82 x 102 228Th 0.109 0.0156 5.47 x 10° 224Rg, 0.109 0.0547 5.47 x 10° 220Rn 0.109 a 1.09 x 10t 216 pg ; 0;109 a. - “’ a 212pp 0.109 5.47 x 10-28 1.82 x 102 212Bj 0.109 2.73 x 10-% 3.65 x 10! 212pg 0.070 a a =208m 0.039 a a Total 1.09 3.78 7.71 x 108 -aRCG not available; isotope is ignored in calculation of hazard. ’, 21 -Table A.Z, Radioactivity and Volumes of Air and Water (Cubic Meters) Potentially Contaminated to RCG by Radioactive Daughters in Equilibrium with 1.0 g of Natural Uranium Redioactivity Ingestion Hazard Inhalation Hazard Nuclide (rCi) (cubic meters of water) (cubic meters of air) 2387 0.331 " 8.27 x 1073 1.1 x 10° 234Th - 0.331 ' 1.65 x 1072 o 331 234 py, 3.31 x 10-3 a ' a 234Mpg 0.331 a a 2347 0.331 1.10 x 10~2 8.27 x 10% 230 Th 0.33L 0.165 .13 x 10° 226 Rg, 0.331 11.0 1.65 x 10° 222Rp 0.331 a 110 21l8pg 0.331 a a 214pp 0.331 a ) 2l4Bj 0.331 a, : a 2l4pg 0.331 a a 210 py 0.331 3.31 8.27 x 10% 210 B 0.331 8.27 x 1078 1.65 x 103 210 pg 0.331 0.473 4,73 x 104 2385y 0.0154 5.15 x 10-% 3.86 x 103 231Th 0.0154 7.72 x 10-% 0.39 231pg 0.0154 0.0172 3.86 x 108 227 pAc 0.0154 7.72 x 1073 1.93 x 10° 227Th 0.0152 7.61 x 1074 2.54 x 108 2237y 2.2 x 1074 a a 223Rg, 0.0L54 0.0221 1.93 x 103 219Rn 0.0154 a a 218pg 0.0L54 a | a 211pp 0.015k4 a a 211Bj 0.0154 a a 21llpg h. 6 x 10-5 a a 2077 0.0L54 a _ a Total 4.8 15.1 ‘ 5.21 x 108 ®RCG is not specified; isotope is ignored in calculation of hazard. -3 " 23 INTERNAL DISTRIBUTION 1. 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