ORNL-NUREG-TM-12.txt

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emiphoyees, noi anv of thelr cortiaciors, subcontractors, or their employges, makes
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accuracy, coripletenass or usefulness of any information, apparatus, prodoct or
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ORNL/NUREG/TM-12
Dist. Category UC-11

Contract No. W-7405-eng-26

CHEMICAL TECHNOLOGY DIVISION

CARBON-14 PRODUCTION IN NUCLEAR REACTORS

Wallace Davis, Jr.

Manuscript Completed: January 1977
Date Published: February 1977

Prepared for the
U.S. Nuclear Regulatory Commission
Office of Nuclear Material Safely & Safeguards
Under Interagency Agreement ERDA 40-549-75

Prepared by the
OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
operated by
UNION CARBIDE CORPORATION
for the
ENERGY RESEARCH AND DEVELOPMENT ADMINISTRATION
CARBON-14 PRODUCTION IN NUCLEAR REACTORS
W, Davis, Jr.

ABSTRACT

Quantities of "*C that may be formed in the fuel and core structural materials of
light-water~cooled reactors {(L.WRs), in high-temperature gas-cooled reactors (HTGRs),
and in liquid-metal-cooled fast breeder reactors (LMEFBRs) have been calculated by use
of the ORIGEN code.' Information supplied by five L. WR-fuel manufacturers pertaining
to nitride nitrogen and’ gaseous nitrogen in their fuels and fuel-rod void spaces was used
in these calculations. Average nitride nitrogen values range from 3 to 50 ppm (by
weight) in LWR fuels, whereas gaseous nitrogen in one case is equivalent to an
additional 10 to 16 ppm. Nitride nitrogen concentrations in fast-flux test facility
(FFTF) fuels are 10 to 20 ppm. The principal reactions that produce “C involve N,
Y0, and (in the HTGR) "C. Reference reactor burnups are 27,500 MWd per metric ton
of uranium (MTU) for boiling water reactors (BWRs), 33,000 MWd for pressunized
water reactors (PWRs), about 95,000 MWd per metric ton of heavy metal (MTHM) for
HTGRs, and 24,800 MWd/MTHM for an LMFBR with nuclear parameters that pertain
to the Clinch River Breeder Reactor. Nitride nitrogen, at 3 median concentration of
25 ppm, contributes 4, 15, and 6 Ci of "C/GW{c)yr to BWR, PWR, and LMFBR
fuels. respectively. The contribution of 'O in BWR and PWR fuels is 3.3 and 3.5 Ci of
"CIGW(e)-yr, respectively, but it is Jess than 0.2 Ci/ GW(e)-yr, in blended LMFBR fuel.
In the HTGR fuel particles (UC: or ThO.), 10 Ci of "C) GW(e)-yr will be formed from
25 ppm of nitrogen, whereas 'O in the Th{: will contribute an additional
2 Ci/GW(e)yr. Ali 'C contained in the fuels may be released in a gas mixture (CO,,
CO, CHa, ete)) during fuel dissolution at the fuel reprocessing plants. However, some
small fraction may remain in aqueous raffinates and will not be released until these are
converted to solids. The gases would be released from the plant unless special equipment
is installed to retain the "“C-bearing gases.

Cladding metals and other core hardware will contain significant quantities of ",
Very little of this will be released from BWR, PWR, and LMFBR hardware at fuel
reprocessing plants; instead, the contained B30 to 60 Ci/GW(e)-yr for LWRs and
about 13 Cif GW(e}-yr for a CRBR, will remain within the metal, which will be retained
on site or in a Federal repository. The only core structural material of HTGRs will be
graphite, which will contain 37 to 190 Ci of "/ GW{e)-yr, exclusive of that in the fuel
particles, if the graphite (fuel block and reflector block) initially contains 0 to 30 ppm of
nitrogen. All of this is available for release at a fuel reprocessing plant if the graphite is
burned to release the fuel particles for further processing. Special equipment could be
installed to retain the "“C-bearing gases.
1.0 INTRODUCTION

The radioactive nuclide '“C is, and will be, formed in all nuclear reactors due to absorption of
neutrons by carbon, nitrogen, or oxygen. These may be present as components of the fuel,
moderator, or structural hardware, or they may be present as impuirities. Most of the “C formed in
the fuels or in the graphite of HTGRs will be converted to a gaseous form at the fuel reprocessing
plant, primarily as carbon dioxide; this will be released to the environment unless special equipment
is installed to collect it and convert it to a solid for essentially permanent storage. If the "“C is
released as carbon dioxide or in any other chemical form, it will enter the biosphere, be inhaled or
ingested as food by nearly all living organisms including man, and will thus contribute to the
radiation burden of these organisms. Carbon-14 is formed naturally by reaction of neutrons of
cosmic ray origin in the upper atmosphere with nitrogen and, to a lesser extent, with oxygen and
carbon. Large amounts of '“C have also been formed in the atmosphere as a result of nuclcar
weapons explosions.

For the last two decades, the quantities of '*C in the environment, and the mechanisms of
transfer of this nuclide between the atmosphere, land biota, and the shallow and deep seas have been
the subject of many research studies.”” These studies have shown that most of the "*C is actually
contained in the deep oceans, at depths greater than 100 m. The nuclear weapons tests increased the
total '*C inventory of the earth by only a few percent,’ but the atmospheric content was
approximately doubled. Since atmospheric weapons tests are no longer being conducted, the
atmospheric concentration of '*C is now decreasing as it enters the oceans as CO; and is
approaching the pretest value,

Some estimates of the amounts of "“C released from or formed in LWRs, " HTGR,"""* and
LMFBR" have been made previously on the basis of calculations or measurements. The purpose of
this report is to present detailed estimates of the production of '“C with emphasis on those pathways
that are likely to lead to the release of this nuclide, cither at the reactor site or at the fuel
reprocessing plant.

2.0 MECHANISMS OF CARBON-14 FORMATION IN NUCLEAR REACTORS

Carbon-i4 1s formed from five reactions of neutrons with isotopes of elements that are normal
or impurity components of fuel, structural materials, and the cooling water of LWRs, The
cutron-induced reactions are as follows:

I

(1) "C(n,y"C;
(2} "N@,p)"C;
3) "N@.d)"C;
(4) "O(n,'He)"'C;

(5) "Om,a} C.
In these reactions, standard notation has been used in which n refers to a neutron, p to a proton, d
to a deuteron ("H), and v to a gamma ray. Reactions 4 and 5 will occur in any reactor containing
heavy-metal oxide fuels and/or water as the coolant. Reaction | will be important only in the
HTGRs, while reactions 2 and 3 will occur in all reactors containing nitrogen as an impurity in the
fuel, coolant, or structural materials.

To facilitate calculations, the energy-dependent cross sections of nuclear reactions are typically
collapsed into a single, effective cross section that applies 1o the neutron spectrum of the reactor in
guestion. Such collapsed values are known with fairly good accuracies for reactions 1, 2, and 5 for
the thermal-neutron spectra of LWRs and HTGRs. Values listed in Table | for the BWR, PWR,
and HTGR are taken from the ORIGEN library' and its update'® according to the latest version of
the “Barn Book.”!” Because reactions 3 and 4 are highly endothermic, their cross sections are
assumed to be 0.0 in thermal reactors, as shown in Table I. Unfortunately, some of these cross
sections for the LMFBR are very uncertain. The following discussion concerning cross sections of
reactions 1-5, as they apply to the Clinch River Breeder Reactor (CRBR), has been provided by
A. G. Croff.™

Reaction | " Cn,v)"C

The cross section for this reaction is not well known for nonthermal neutron energies. The
assumed values were taken from ref. 19, in which the *C(n,y) cross section was calculated on the
bases of a few experimental data and nuclear systematics. The cross section obtained when the data
are collapsed to a single value using the CRBR neutron spectrum 15 0.5 ub (I ub = (0™ barns). The
fact that the thermal "C(n,¥) cross section is only about 1 mb (Table 1) couplied with the fact that
cross sections in the nonthermal energy regions are considerably smaller than thermal cross sections
tends to confirm that the 0.5 ub value is realistic.

Reaction 2 “N(n,p)"'C

Of the five ""C-producing reactions listed, this is the only one for which the experimental data
may be considered adequate. Energy dependent cross-section data for the *N(n,p)"*C reaction are
available from the ENDF/B"® compilation. Collapsing these data with the CRBR spectrum gives a
cross section of 12.6 mb, with an estimated error of +30%.

Heaction 3 lS!\/’(n,af}MC

The only cross-section data available for this reaction are some sketchy information on the
angular distribution of the deuterons when the neutrons have energies of 14 to 15 MeV., This
information, coupled with the fact that the reaction is endothermic (Q = -7.99 MeV), would
probably lead to a value of the reaction rate in the 0.00 to 0.1 mb range. However, for
calculational purposes, a value of 1.0 mb was used.

Reaction 4 'O, He)"' C

Of the five reactions considered, the data for this reaction are by far the least well-known. It is
highly endothermic (Q = -14.6 MeV), indicating that greater neutron energies are required for the
teble 1. Cross sections for formation and yields of *¢ in BWR, PWR, HTGR, and LMFER®

 

140 Tormation

 

 

Reaction Cross section for formation of 14¢ in (curiesAper gram of parent element)
No. Reaction BWR PWR HTGR IMFBR BWR PAR HTGR IMFER
1 Y20(n,v) e 1.00 mp 1.00 mb 0.416 mb 0.5 ub 1.51E-7 1.618-7 3.388-7 4. 81E-9
(3.69E+0)

2 N (n,p)*4C .48 v 1.48 b 1.02 12.6 mb 1.718-2 1.83E-2 3.84E-2 9.668-3

3 *Bn(n,a)t%c 0 0 0 1.0 mb 0 0 0 2.855.6

4 180(n,%He )2 4 0 0 0 0.05 b 0 0 0 3.82E-8
(4.53E-2)°

5 Y70(n,*Re ) % 0.183 0.183 v 0.110 b 0.12 mb 7-31E-T7 0 T.75E-7 . 1.79E-6 3. 4oE-8
(1.01E-1)" (0.878-2)" (2.25p-1)¢ (4.03E-2)C

 

 

aAll of tne valuves in this table woere obtained by collapsing available neutron cross-section data to a
single value, using neurvon spectra of the individual reactors, as discussed by 8el1.1 These values
are mnot ecual to 2200-m/sec cross sections, such asg 0.9 mb, 1.81 b, and 0.235 b for reactions 1, 2,

and 5, respectiveiy.

b . . . , .
Based on 10.93 MT of carbon/MTEM where HM = thorium pius uranium,

CBased on 8383 g-at. of oxygen/MTHM where EM = uranium or uranium plus plutonium, present as UO2 and
Pu0
& 2 -

Based on 2.9094 MT of thorium/MTHM with +thorium oresent as ThO2 and uranium as UC.
reaction to proceed. Information supplied by the Physics Division of Lawrence Livermore
Laboratory indicates that the cross section at 15 MeV should be less than | mb, and at 20 MeV it
should be less than 10 mb. By combining these “guesstimates™ with the CRBR spectrum and a
theoretical expression for the availability of high-energy fission neutrons, the reaction cross section
is estimated to be about 0.05 ub. The lack of information on both the high-energy cross sections and
the high-energy neutron spectrum makes this value very uncertain,

Reaction 5 " Ofn,a)"*C

As with reaction 1, the cross-section data for this reaction are not well known. The data, which
again are based on only a few experiments and nuclear systematics, were taken from ref. 19. The
cross section, which is calculated and based on the CRBR spectrum, is 0.12 mb.

The assumed LMFBR fuel model was the Atomics International Follow-On Design. Initial
concentrations of the isotopes of importance in this case (in g-atoms/ MTHM) are:

12

’C 33.33
"¢ 0.374
"N 1.42
"N 0.00528
"*Q 8383.

"0 3.27
0 17.2

The ORIGEN code' is not capable of explicitly accounting for (n,d) or (n,'He) reactions. This
difficulty may be circumvented by combining reaction 4 with reaction 5 and reaction 3 with
reaction 2, since the naturally occurring isctopes are present in a fixed ratio for each element.
Alternatively, since the depletion of the carbon, nitrogen, and oxygen is relatively small (<<2%j). the
calculation is easily performed by hand.

3.0 CARBON-14 FORMATION IN LIGHT-WATER REACTORS

Carbon-14 is formed in the fuel (UQ2), in core structural materials, and in the cooling water of
L.WRs.

3.1 Formation in the Fuel

Carbon-14 will be formed primarily by two reactions in the fuel: "O(n,a)"'C and “N(n,p)"'C.
The quantity of "*C formed from the first of these reactions can be calculated accurately on the basis
of the stoichiometry of UQ: (134.5 kg O/MTU) and an abundance of 0.039 at. 9% 'O in normal
oxygen, which corresponds with 55.6 g of "O/MTU or 3.27 g-atoms of '70/‘MTU. As listed 1in
Table 2, burnup of BWR and PWR fuels to 27.500 and 33,000 MW(t)d/MTU, respectively, leads to
the formation of 0.098 and 0.104 Ci of *C/MTU, which corresponds with 3.3 and 3.5 Cif GW{e)-yr,
respectively.
Table 2. Production of '*C in core hardware and Tuel at light-water reactors (BWR and PWR)

 

14
C existing 150 days after
Total '*C production

 

 

 

 

 

 

Qua:;;ity Quantity of element in core discharge of fuel (Ci/MTU)
core {g/M70) From From From Calcuiated Observed
Material {kg/MTU) Carbon Nitrogen Oxygen carbon nitrogen oXygen Ci/MTU cifcvi(e)-yr® Ci/GW{e)-yr
Soiling-Water Reactorb
Zircaloy-2 {Grade RA=1} 316 £85.3 $25.3 1.29E-5 4.338-1 0.k33 14,5
30k stainless steel 50 <40.0 50-80 0.60E-5 (0.86-1.37)E+0 0.86-1.137 28.7-45.9
Inconel-X 3.4 <34 0.058-5 0., 000 2.0
Uranium dioxide 1135 low 10 134,500 1.718-1 9.83E-2 0.269 9.0
Med 25 L.28E-1 0.526 17.6
High 79 1.288+0 1.38 6.3
ater 216 192,000 1.408-1 0,140 L.7 &°
Totals, Low 1.70 57
Med a.21 7h
High 3.32 111
Pressurized-Water Reactox‘ti
Zircaloy-i {Grade RA-2) 2135 <53%.5 £18.8 1.02E-5 2.74E-1 0.274 9.5
302 stainiess steel L.2 <3, L.2-6.7 0.05E-5 (0.61-0.98)8-1 0.061-0,098 2.1-3.4
304 stainless steel 7.1 £29.7 37.1-59.4 0. %8E-5 (5.42-8.671E-1 0.5k2-0, B6T 18.8-30.0
Inconel 716 12.8 1.3 C.02E-5 0. 000 0.0
Microbraze 50 2.6 0.3 c.2 1.1 0.00E~5 3,66E-3 0.858-6 0.004 0.12
Uranium dioxide 1135 tow 10 134,500 1,938-1 1,041 0.287 9.6
Med 25 4, 57E-1 0.561 i8.8
High 79 1. 37E+0 1.48 k9.5
Water 216 192,000 1.402-2 0. 1%g 5.0 ¢
Totals, Low 1.32 44
Med .77 59
High 2.87 9%

 

Bhased on 33.5 MTU/GWie)-yr.
Y)RIGEN caiculations assume 18.823 MWt)/MTU, & years in reactor, to 27,500 MHd/MTU; 2.6 wt % 2%*U. Quantities of metal in core from ref. 2%i.

. 2 . : . ; . . . - .
“The measured vaiuel at the Nine Mile Point reactor [625 MWie)}] was & Ci/yr; see text for comments on power density and steam/liguid water volume.

dORIGEN calculations assume 30,0 MW{t)/MTU, 3 vears in reactor, to 33,000 MWA/MTU; 3.3 wt % 2%, Quantities of meta: in core from re?. 22.
There is considerable variation in production of "“C from the '“N{n,p) reaction because of
variations in the nitrogen content of LWR fuels. Crow’’ presented the following brief summary of a
survey of five fuel fabrication plants:

Maximum nitrogen allowed by specification, ppm 75-100
Maximum nitrogen reported, ppm 100
Minimum nitrogen reported, p;fim 1
Average nitrogen in reactor fuel, ppm 255

He has indicated that the 25 +35 ppm average is not a true arithmetic average but a consensus
derived from discussions with representatives of fuel manufacturers.

Table 3 contains the results of 2 much more extensive survey of the nitrogen content of fuels
made at these same five plants. The current average nitrogen content varies from 3 to 50 ppm and
the standard deviation of each average is in the range of 40 to 70% of the average. The data shown
in Table 3 suggest that the median value of fuel from all plants is about 25 ppm,‘

The differences in the nitride-nitrogen concentrations in LWR fuels from the five manufacturers
listed in Table 3 are due to many variables. Some of these have been described qualitatively and are
discussed by Pechin et al.”* without reference to reaction times, temperatures, and concentrations.
Uranium hexafluoride from gaseous diffusion plants, enriched to 2 to 4 wt % in *°U, is the starting
material in the manufacture of LWR fuels. Four of the manufacturers use the ammonium diuranate
(ADU) process, and one uses the direct (dry) conversion (DC) process. Powdered UQ, is obtained
from both processes, cracked NH; being the preferred source of hydrogen reductant. Pellets are
obtained by pressing the powder into pellet form and sintering these in hydrogen, as in the
uranium-valence reduction step. Pellet pressing is performed as a dry operation (except for a little
lubricant). Sintering is performed at temperatures ranging from < 1600°C to = 1750°C. After
cooling, the pellets are loaded into Zircaloy fuel tubes (closed at one end), usually without any
additional treatment. Before the fuel tube is Welded closed in a helium atmosphere at all plants, air is
removed in a vacuum degassing step at four plants, but is left in place at one of the plants. During
the degassing operation, pellets in the fuel rods are unheated in some plants and heated in others. All
vaccum degassing operations are followed by filling the fuel rod with high-purity helium and closing
the second end by welding in 2 helium atmosphere. Helium is added under pressure to fuel tubes at
the plant at which the the vacuum degassing step is not employed. The gaseous nitrogen from 18 to
30 ¢c of air in a single fuel tube containing about 1.75 kg of UO; corresponds to an additional 10 to
16 ppm-of N, that is not included in Table 3.

Because of the wide range of nitrogen concentrations, three values of '*C production from the
“N(n,p) reaction are listed in Table 2. These correspond to 10, 25, and 75 ppm of nitrogen. At these
three levels, "*C production for the listed burnup conditions are 0.171, 0.428, and 1.28 Ci/MTU,
respectively, which corresponds to 5.7, 14.3, and 42.9 Ci/ GW({e)-yr for the BWR. Similar values for
the PWR are 0.183, 0.457, and 1.37 Ci/ MTU, respectively, and 6.1, 15.3, and 459 Ci/GW(e)-yr.

It may be noted that the same quantity of "C will be produced from "O(n,o) and ""N{n,p)
reactions when the nitrogen content of the fuel is about 5.7 ppm for both PWRs and BWRs.

The chemical form of "“C in the fuel is not known. When formed from any of the five nuclear
reactions presented in Sect. 2, this nuclide might become bound to uranium as carbide, remain as
impurity atoms, or be converted to carbon monoxide or carbon dioxide. A nitrogen impurity of
75 ppm corresponds to 1.28 Ci of "C/MTU in the case of the reference BWR and to 1.37 Ci of
“C/MTU in the case of the reference PWR (Table 2). These maximum expected activities
Table 3., Ni

trogen content of U0, fuels for LWRs and of FFTF fuelg?

 

 

 

-_—

 

 

 

 

 

 

 

 

 

 

 

FFTF fuels®[ (U,Pu)0. ]
Current production of LWR fuels (U0;) Compan

pany A fuel ComEanX B fyuol

 

 

 

 

 

 

 

 

 

 

 

Company Analyzed by Analyzed by
1 2 3 4 5 Company A  HEDL Companv B HEDL
No. of measurements 358 408 38 206 70 8G 10 80 10
Percent of measurements with nitrogen, ppm
<10 i¢o 75 42 14 10 68 100 78 9¢
it - 20 12 53 39 1 4 17
20 - 35 9 36 16 12 5 10
>35 4 5
35 - 50 10 27 2
>50 1 486 14
Mass-weighted av nitrogen, pom 2.8 13.3 13.7  21.6 47.8 <21.6%  <1p© <11.1¢ <9 2¢
Std deviation, ppmS 1.4 8.3 9.8 11.1  21.2 N.A N.A N.A N.A.
aPrimarin nitride nitrogern.
bFrom ref, 52.
CNumerical values are based on using the many values <10 ppm as 1C0.0 ppm.
It is emphasized that the distribution of nitrogen analyses is not normal. N.A. (not available) 1is

used because a meaningful sta

 

 

 

 

 

 

ndard deviation cannot be calculated,
correspond to a ratio of about 1 "*C atom/ 200,000 uranium atoms. Ferris and Bradley™ studied the

reactions of uranium carbides with nitric acid and found that 50 to 80% of the carbide carbon was
converted to carbon dioxide;, the remaining carbide carbon was converted to nitric acid-soluble
chemicals such as oxalic acid, mellitic acid, and other species, probably aromatics highly substituted
with -COOH and -OH groups. Formation of such compounds can be reconciled with the existence
of the polymeric -C-C- bonds of uranium carbides. However, at a ratio of 1 “C atom/ 200,000
uranium atoms, or even at a ratio | C atom/500 uranium atoms, which would correspond to an
impurity of 100 ppm of carbon in the UQ;, there will be a very low concentration of -C-C- bonds in
the UO- fuels. This suggests that a larger quantity of any carbide carbon, including that formed from
nuclear reactions, will be converted to €O, in dissolving operations at the {uel reprocessing plant
than the 50 to 80% reported by Ferris and Bradley®® for pure uranium carbides. An experimental
program to measure C liberated during fuel dissolution is now in progress.”

3.2 Formation in Core Hardware

Core structural materials include stainless steel support hardware, Zircaloy cladding, and nickel
alloys used as springs and fuel tube separators. According to specifications,”” "' the primary source
of “C in these materials is the nitrogen that is present in quantities listed in Table 4. The quantities
of each of the types of metal (i.e., stainless steel, Zircaloy, Inconel-X) are somewhat dependent on
the reactor type (BWR™ or PWR™ ) and on the year and size of the design within a reactor type.
For example, Fuller ei al.”” have presented data on the fifth and sixth generation BWRs (BWR/5
and BWR/6) from which the weight ratios are calculated to be 247 and 265 kg of Zircaloy-2/MTU,
respectively. Other estimates of quantities of structural hardware have been given by Griggs™ and by
Levitz et al.”” However, the quantities of these metals, the contained nitrogen, and the 'C produced
(as listed in Table 2} are based on information pertaining to present reactor designs provided by
Marlowe’ and Kiip.™* Carbon-14 values are based on calculations with the ORIGEN code' for a
BWR operated to a burnup of 27,500 MW(1)d/MTU in 4 yr and a PWR to a burnup of 33,000
MW(t)d/ MTU in 3 yr. The revised light-element library'® was used in these calculations. Most of the
“C formed in these structural components will be retained within the metal when the latter is
encapsulated for long-term disposal, although a very small fraction in the Zircaloy might be
dissolved in fuel leaching solutions at the fuel reprocessing plant. Experiments have never been
performed to evaluate this possibility.

3.3 Formation in Cooling Water

Oxygen of the cooling water and nitrogen-containing chemicals in this water are sources of HC,
An accurate calculation of the quantity of “C that will be formed would require integrating the flux
over the volume of water in and surrounding the core. Data to perform such an integration do not
appear to be readily available, but reasonable approximations can be made. Reference 34 gives
values for the atomic ratio H/U of 3.74 and 4.23 for BWRs and PWRs, respectively; these
correspond to 7860 and 8890 g-atoms of O (as H:0)/MTU. tuller et al.” give values of the
water; fuel volume ratio of 2.52 for BWR -5 and 2.50 for BWR ;6. A water density of 0.805 g/cm’
and a UQ, density of 10 g/cm’, both at 556"F, indicate a ratio of about 13,000 g-atoms of O/ MTU
for the BWR cores. Reference 36 gives a hot, {first care H,0/ UO; volume ratio (for a PWR) of 2.08,
Tavlie L. Specifications for carbon and nitrogen in

reactor structural and claddinrg metals

 

Specifications {(wt %)

 

 

Reactor
type Carbon Nitrogen Reflerences for specifications
. 27 25
Steinless steel 204 BWR <0.08 0.10-0.16 ASME SA213-73  and ASME SA-2L ‘
27 2
304 PWR <0.08 0.10-0.16 ASME SAZ213-73 and ASME SA-2LO
29
316 IMFBR 0.040-0. 060 <0.010 RDT M73-287
5
Zircaloy=2 BWR <0, 027 <0, 008 ASTM B253-71 (ANSI N12M-1973>3
. O
Zircaloy-L PWR <0,027 <0.008 ASTM B353-7% (ANST N124-1973}3
Tnconel-X RWR <0.10 Trternational Nickel Co. o-
) . ) o 31
Inconel 718 PWR =0.10 nternational Nickel Co.
Nicrobraze 50 PWR 0.01 0. 0066

 

01
11

which corresponds to about 10,500 g-atoms of O/MTU. For the purpose of this report, it is thus
assumed that the rate of reaction ''O(n,a)"'C is specified by a ratio 12,000 g-atoms of O/MTU and
a natural 'O abundance of 0.039 at. % in oxygen for both BWRs and PWRs. This corresponds
(Table 2) to about 4.7 and 5.0 Ci of "C/GW(e)-yr for BWRs and PWRs, respectively, from the
YO(n,a)"'C reaction: it also corresponds to an initial atomic ratio H/ U of about 220 for BWRs
and 175 for PWRs using fuels containing 2.6% and 3.3% °*°U, respectively.

The quantity of "C formed from impurity nitrogen cannot be estimated since there do not
appear to be any analyses Ifiertaining to the concentration of this element in reactor cooling water.
Although its concentration may be no more than a [ew parts per million, Cohen® mentions a value
as high as 50 ppm NH: in the primary cooling water of PWRs.

Quantities of '*C actually released from a BWR and three PWRs, as measured by Kunz and his
coworkers,'''* are listed in Table 2. From the BWR at Nine Mile Point (625 MW(e)] they
observed'” a release rate of 8 Ci of "*C/yr. These authors also reported 6 Ci of *C/GW(e)-yr on the
basis of their analyses of gaseous effiuents from the Ginna, Indian Point 1, and Indian Point 2
PWRs. At the PWR stations,'' over 80% of the "“C activity was chemically bound as CH, and C,H,;
only small quantities were bound as CO». At the Nine Mile Point BWR station'” the chemical form
of "“C was greatly different, with 95% as CO,, 2.5% as CO, and 2.5% as hydrocarbons.

On the bases of the fuel isotopic compositions and burnups shown in the footnotes of Table 2
and for the assumed ratio of 12,000 g-atoms of O/ MTU, an impurity of 1 ppm.of nitrogen in the
cooling water {corresponding to 0.216 g of N/MTU) would lead to the formation of 0.124 and 0.132
Ci of '4C/GW(e)~yr in BWRs and PWRs, respectively. The difference between a calculated 5 Ci of
M/ GW(e)yr from the "O(n,e) reaction and the observed 6 Ci/yr at the PWR stations'' (Table 2)
is probab!y well within limits of analytical uncertainty. The extrapolation to 16 Ci of ""C/GW(e)-yr
from the measured 8 Ci/yr at the Nine Mile Point BWR is based on maintenance of a constant
power density and a constant volume ratio H,0/ UO,. Values of this ratio tabulated for the Nine
Mile Point reactor’’ and for newer, larger reactors, such as those at Brown’s Ferry,42 do not differ
significantly (2.38 vs 2.43); the average power densities for the two reactors are 41 and 50.732
kW/liter, respectively. When these ratios are combined with data on the average void fractions
within a fuel assembly (a measure of steam/liquid water, and having values of 0.3 for the Nine Mile
Point core and 0.4 for the Brown's Ferry core), it is apparent that “C formation in a new 1100
MW(e) BWR (such as BWR/5"™) would be larger than 8 Cij GW(e)-yr, but significantly less than
16 Ci/ GW(e)-yr.

4.0 CARBON-14 FORMATION IN HIGH-TEMPERATURE GAS-COOLED REACTORS

The only structural materials in HTGRs in which "C will be formed to any significant extent
are the fuel containing and reflector blocks of graphite. There will be some nitrogen and oxygen in
the helium coolant.*’ However, the rate of "*C formation from coolant impurities will be very small
in comparison with similar rates in the fuel blocks; in addition, the helium cleanup system is
expected to remove CO:, a probable form of part of the "*C in the coolant.

4.1 F"ormation in the Fuel

The compositions of fertile and fissile fuel for HTGRs have not been positively established since
commercial reactors are not yet being roade. However, it is highly probable* that the initial and
12

makeup (the IM stream) fuel will be in the form of about 93 wt 9% of *U as UC,, that *"’U bred
from the fertile thorium will be recycled as UC; (the 23R stream), and that uranium recovered from
the IM stream after reprocessing, if it is recycled as the 25R stream, will also be in the form of UC,.
Similarly, the fertile thorium is expected to be in the form of ThO;. Uranium in the IM stream will
have a chemical history different than that of uranium in the 23R and 25R streams. In particular,
uranium for the IM stream will be received at a fresh-fuel fabrication pla,m45 as UF,, which will be
hydrolyzed with steam to UOQOsF:; this, in turn, will be reduced at about 650°C with H, ( from
cracked ammonia) to UO.. Subsequently, the UO: will be mixed with carbon flour, ethyl cellulose
and methylene chloride. 1t will then be dried, ground, separated into appropriate sizes, and heated in
a vacuum to cause the formation of UC,. Finally, it will be cooled in an inert atmosphere, which
may cither be nitrogen or argon. {n these successive processes, the uranium-bearing matenal never
exists as a nitrogen-containing compound, although it is exposed to N; from cracked ammonia at a
high temperaiure and may be exposed to nitrogen after formation of UC;.

On the other hand,' recycle uranium, both 23R and 25R streams, will pass through the uranyl
nitrate [UO2(NO;):] state i a fuel reprocessing plant. These materials will be denitrated and
converted to UQO, before subsequent carbonizing steps that are similar to those described {or the [IM
material. The significance of the differences in histories is that recycle uranium may contain more
nitrogen (from undecomposed nitrate) than does the initial or makeup 93% **U.

There are limited data concerning the quantities of nitrogen in potential HTGR fuel since this
fuel 1s not made on a routine basis. It is therefore assumed that all forms of UC; and ThO: contain
the same quantity of nitrogen (i.e., 25 ppm) used in this report as an industry concensus for LWR
fuels. On this basis, about 0.96 Ci of “C/MTHM, or about 9.7 Ci; GW(e)-yr will be formed from
the '4N(n,p) reaction.

Carbon-14 will also be formed to the extent of 0.225 CiyMTHM, or 2.3 Ci/GW(e)-yr, from the
reaction ' 'O(n,a)"C of oxygen present as ThO: (Table 35).

4.2 Formation in Graphite Blocks

Independently of the "N(n,p)'C reaction, significant quantities of "“C will be formed in
graphite of fuel and reflector blocks due to the reaction "‘C(n,y)*C. Based on a lifetime average
ratio of 10.93 MTC in fuel blocks; MTHM, about 3.7 Ci of "*C/MTHM, or 37 Ci/GW(e)-yr. will
be formed from this (n,v) reaction (Table 5). Additional *C will be formed in reflector blocks,
which are present to the extent of 16.29% of fuel blocks on a lifetime average basis. The neutron flux
in reflector blocks will be about 70 to 80% of the corc-average flux, although the "“C production
listed in Table 5 is based on a flux in these reflector blocks equal to the core average. The total *C
formed from the ''C(n,y) reaction in fuel blocks and reflector blocks is less than 4.3 Ci/ MTHM, or
less than 43 Ci/ GW(e)-yr.

The amount of nitrogen present in fuel-block or reflector-block graphite is uncertain. Four
samples of graphite were irradiated in the Oak Ridge Rescarch Reactor (ORR) and were
subsequently analyzed for “C.* The quantity of this nuclide in excess of that calculated to be
formed from the '"C(n,y)"*C reaction was ascribed to the reaction "*N(n,p)*C. On the basis of this
assumption. the equivalent nitrogen impurity was calculated to be 3.2 to 84 ppm on a
graphite-weight basis. The only other estimate of nitrogen content in an in-use graphite is 26 ppm."*
and is used here as the basis for the value of 30 ppm of nitrogen in fuel blocks and reflector blocks
listed in Table 5. Carbon-14 formed in graphite containing 30 ppm of nitrogen corresponds to
126 Ciy MTHM or 127 Ci/GW(e)-yr.
_Table 5. Production of 1% in graphite and fuel al High-Temperature Gas-Cooled Reactors

 

14: existing 160 days after

discharge of fuel

 

 

 

 

Impurity content Material Quantity of element in core {C1/MIHM )
. {g/MriM) ; F From Total '*c
Nitrogen Oxygen in core From rom ro <
Material {ppm} {wt. %3 _{MT /MTHM ) Carbon Ritrogen Oxygen carbon nitrogen OXygen Ci/MIHM Ci/GWie)-yr
Graphite in fuel blocks 107 10.5%° 1.0G3E+7 3.28E+2 3,69 12.58 i6.27 164
Graphite in reflector b . 4
tlocks 30 .77 1.77E6 3. 54E+E . <0.60 <2,0h . <2.6% <=6.6
IM uraniom (1K, } 25 0. oi;5h81 2.50E+1 G.95G G, Olli Q. L
Recycle uranium (s ) 25® o.chsize’ 2. 50E+1 0.559 0. ot G, s =
Thorium dioxide 25° i2.1g 0.9091'&1f 2,50B+1 1.25E+5 C.959 0,255 1.08 10.5

Total

 

®Rased on 10.11 MEM/CWlel=yT {eguivalent to 38.9% efficiency in converting hest to electricity).

bThi: ig an estimate based on the sssumplion that no great efforts will be made tc minimize the nitrogen content,

“See ref. 13.

dBased cn & neuirsn flux in reflecior blocks eguai te the coresaverage flux. HNowever, the fiux in the reflector blocks will be about 70 to 8% of the core-averusge value.
€Assumed to be the same as in IWR fuels.

Y¥rem rer. 13 the following values are obtained: 405,08 kg {034 3°®U) TH material, 294,07 kg 23R material, 107.83 kg 25R material, and 8394.7¢ kg thorium in the lifetime average annual
reload. values listed sre MY thoriwm or uranium/MTHM.

€211 of this is potentislly svailable for release at the [usl reprocessing plant except asbout 0.012 Ci/MTHM {0.12 Ci/GWie}-yr) in the initislly fissile particles cof the 25R stream
£ E% P P P r
which are designated 25W efter digcharge.
14
5.0 CARBON-14 FORMATION IN LIQUID-METAL FAST BREEDER REACTORS

T'he primary structural material of the core of an LMFBR will be 316 or A-286 stainless steel.
Carbon-14 will be formed from impurities in this metal as well as in the fuel. Since no LMFBR has
yet been buili, discussion presented here is based on the proposed reference design' of the Clinch
River Breeder Reactor (CRBR) and on recent updating of fuel composition.*™ A core element for
this reactor is shown in Fig. 1.

5.1 Formation in the fuel

In common with LWR fuels, ""C will be formed by the "O(n.a) and "“N(n.p) reactions in
LMFBR fuels; in both types of reactor very small quantities of "*C will be formed by the ''C(n.y)
reaction. Two other reactions produce "C in the LMFBR (Sect. 2): "N(n,d) and "O(n,'He).
Croff's'* estimates of cross sections and formation rates are listed in Table 1. Production of *C
from reactions involving oxygen are listed in Table 6; these values are based on 8383 g-atoms of
O/MTHM (in this case, MTHM is uranium plus plutonium) and 0.039 at. % of "0 in natural
oxygen (corresponding to 3.27 g-atoms of O/ MTHM).

The specification limit on the nitride nitrogen impurity in plutonium dioxide™ and driver fuel™
for the Fast Flux Test Facility (FFTF) 1s 200 ppm. Air in fuel rods is evacuated and replaced by
high-purity helium'' before the rods are closed by welding in a helium atmosphere. The maximum
fuel-pellet gas content of 0.09 cc (STP) per gram of fuel,™ exclusive of water, would correspond to
120 g of N/MTU 1if all the gas were nitrogen. Measured nitride citrogen concentrations in FFTF
fuels have been significantly less than specifications, gencrally in the 10 to 20 ppm range,” as shown
in Table 3. Therefore, it is assumed in this report that the concentration of nitrogen in CRBR fuel
will be about 25 ppm, with a range of 10 to 75 ppin. These values were used to estimate an average
and range (Table 7) of "*C formation due to neutron absorption by "N and ""N. The average value
is 0.166 Ci of "“"C/MTHM. or 6.1 Ci of "C/GW(e)-yr; the values range from 0.0665 Ci; MTHM
[2.45 Ci; GW(e)-yr] to 0.499 Ci/MTHM [18.4 Ci; GW(e)-yr]. Formation of "C from oxygen in the
fuel, 0.00364 Ci) MTHM, and from nitrogen would be equal if the nitrogen concentration in the fuel
were about 0.55 ppm.

5.2 Formation in Core Hardware

As noted above, 316 stainless steel (with specifications listed in ref. 29) or A-218, is essentially
the only metal in the CRBR core and may be the only metal in future commercial LMFBRs,
Specification RDT M3-28T, Table 4, requires that the oxygen and nitrogen concentrations be lower
than corresponding values for 304 stainless steel used in LWRs. In particular, the specification of
<0.010 wt % of nitrogen in 316 stainless steel is more than a factor of 10 below the specification of
0.10 to 0.16 wt 9% of nitrogen in 304 stainless steel for LWR applications.

Calculated quantities of "*C to be formed in CRBR cladding are listed in Table 7. These are
based on 100 ppm (0.01 wt %) of nitrogen and on the “mass ratios” shown in Table 6. These ratios
refer only to cladding plus shroud plus wire between bottom and top fuel elevations, The neutron
flux decreases very rapidly with elevation away from fuel levels. For this recason, "C formation in
regions above the fuel level in the upper axial blanket and below the fuel level in the lower axial
blanket 1s neglected.
OR&L DWG 78 - 14882

 

    

5 833-mmn (0 2307.) DIAM
WIRE WRAP .84 n {C. 2D
S f422-mm (0.056n) DiaM / 217 REQD

N\ H.78-cm (Tia) PITCH  /

 

1.62-cm (4.57%4n.)
HEX DUCT TUBE
3.048-mm (0.120-x) WALL

o

 

Fig, 1. Reference CRBR core fuel assembly.

ST
 

 

 

 

 

Table 6. Data pertaining to **C production in the CRBR
' ORIGEN - o , 14
Specif%c Mass sfgiilgzs Mas calculated Specific production of ~ C from
pPoWer of HM a.b ratio burnup _ Carbon Nitrogen Oxygen
MA (L) charged®’ steel®’ (MESS) [ngt)-d (Ci \ Ci ci )
CRBR region MTHM (M) (MT) MIHM { MTHM g c) g N 100 kg 0
Inner core 113.22 1.4361 10.63 0.66 93,066 9.98E-9 1.88e-2 §.398-3
Outer core 104.63 1.2006 9.11 0.66 86,005 6.92E-9 1.328-2 5.48E-3
Upper axisl blanket 3.482 1.0361 8.L0 0.66 2,862 1.47E-9 2.85E-3 1.03E-3
Lower axial blanket 7.276 1.0361 7.77 0.66 5,981 2.66E-9 5.13E-3 1.92E-3
Radial blanket 4.302 3.0373 20,0k 0.185 3,536 1.75E=9 3.39E-3 1.24E-3
Total in reactor 32.3505 56.25 0.393
d
Mass-average 30.154 24,811

 

83ee Rer. L8.

bThe heavy metal (HM) charge is the annual charge; annually, one-third-of the core and axial blankets and one-sixth of the
radial blankets are replaced, The stainless-steel mass is the total in the specified region, not Jjust the {fresh steei, The
mass ratio of stainless steel to heavy metal [{MTSS/MTHM), column 5)] is the sum {cladding mass + shroud mass + wire mass)

Caiculations are based on the following

betwern the bottom and top fuel elevations, Fig.
data for core and axisl blanket tubes (fuel pins, see Fig, 1):

1, per unit mass of heavy metal.
0D = 0.230 in.; ID = 0.200 in,; wire-rod spacer (running
4,575 in.; hex metal thickness = (.120 in.;

nearly coaxially with fuel pin) = 0.055 in. diam; hex face-to-face distance
fuel diameter = 0.20C in.; density of stainless steel = 8,02 g/em’; density of fuel {U0;) = 9.316 (85% of theoretical 10.96

g/cm’®).  The radial blanket fuel rod dimensions are: OD = 0.520 in.; ID = 0.490 in.:

are as given above.

-~

From the stoichiometry of {U,Pu)0,, therc are about 134 kg O/MTHM.

drnis corresponds to 36.80 MTHM/GW(ec)-yr, as used in Table 7.

fuel diam

= 0.485 in.; all other parameters

91
Teble 7. Production of *C in the CRBR™

 

Production of 3¢ in fuel from
Production of 144

 

 

 

 

 

Nitrogen : .
from nitrogen in
Oxygen Low { 10 ppm ) Average {25 ppa} High { 75 pom) stainless steel
CRBR region - - C1/MDHM Ci/GW(elwyT €1 /MTHM Ci/GwWle )-yr Ci/MTHM Ci/GwW{e}-yr C1i/MTHM 2i/Gw{el-yT Ci/MIHM Ci/GWie)-yr

inner core 1.138.2 1.11E-1 1.88E-1 1.84E+0 4, 7TOE-1 Y 61E+C 1.h2E+0 1.33E+1 1.2LE+O 1.22E+1
Duter core 7.35E-3 7.80E-2 1.39E-1 1. LOE+0D 3.30B-1 3.50E+0 3.00E-1 1.056+1 §.738-1 Q. 27E+0
Upper axial blanket 1,39E-3 4, k31 2,85E-2 9. 09E+0 7.12E~2 2.27E+1 2. 1hE-1 6.82F+1 1.868E-1 6.01E+1
Iower axial blanket 2.58E-3 3.9kE-1 5,.13E-2 7.838+0 1.28E~1 1,96E+1 3.85E-1 5. 87E+1 3.398-1 5. 18841
Radial blanket 1.6TE=3 b, 31F-3 3.398-2 8. T6E+0 8. 4B~z 2.19E+1 2. 5hf-1 £.57E+1 £.27E-2 1.65k+%
Mass-average 3, 6hE-3 1.34E-1 6,65E-2 2. 45E+0 1.66E-1 6,128+ 0 4,99E-1 1.8LE+1 3.L49E-1 1. 28E+1

 

®caleulations do not include formetion of '*C in stainless steel above the top or below the bottom of the fuel.

L1
18
6.0 COMPARISONS AND DISCUSSIONS

Calculated quantities of “C that arc or will be produced in the four types of reactors (BWR,
PWR, HTGR, and LMFBR) considered in this report are summarized in Table 8 in units of
Ci/ GW(e)-yr. Ranges are given for all calculated values of "*C from all reactors except the HTGR.
The ranges are due to variations in the nitrogen content of the fuel. Values spanning the full range of
10 to 75 ppm (by weight) are shown in Table 3, which is a suromary of manufacturing data.

The Barnwell plant of Allied General Nuclear Sesvices is designed to process about 5
MTHM/day, or 1500 MTHM/yr, of LWR fuel. Heavy metal (HM) is uranium or uranium plus
plutonium charged tc BWR, PWR, and LMFBR; HM is also uranium plus thorium charged to the
HTGRs, The Barnwell design corresponds to about 45 GW(e)-yr. Similarly, reference HTGR- and
LMFBR-fuel reprocessing plants are designed to process annually fuel that produced about 45
GW(e)-yr of energy. Using this factor as a multiplier for values listed in Table 8, it is appropriate to
examine the total quantities of '*C that would be released from the various fuel reprocessing planis if
cquipmeni is not instalied to collect and retain the gases containing this nuchide; it 15 also
appropriate to examine how much will be contained withun the hardware that becomes part of the
high-level waste that may be shipped to a Federal repository. Light-water reactor fuel processed in
| year in a Barnwell-sized plant will contain 400 to 2200 Ci of "*C: the hardware will contain 1400 to
2700 Ci of ""C. The calculated values for "'C in the hardware are conservatively high since they are
based on the assumption that all core hardware ~ not just the cladding — is in as intense a flux field
as 1s the cladding.

Lesser quantities of *C will be produced in LMFBR fuel. The fuel entering a reprocessing plant
of 45 GW(e)-yr capacity will contain 100 to 800 Ci of “C per year while the cladding will contain
about 600 Ci of "“C per year. Quantities of this nuclide in other hardware are not included in
Table 8.

The "'C content of HTGR fuel entering a 450 MTHM/ yr [45 GW(e)-yr] fuel reprocessing plant
in | yr will be about 530 Ci if the nitrogen content of the fuel is 25 ppm. Only this “median” nitrogen
content is considered because the graphiie probably will be the dominant source of "“C. In
particular, if there is no nitrogen in the graphite, the "*C content [due solely to the ''C(n.y)"'C
reaction)] of graphite entering the fuel reprocessing plant 1 | yr will be about 1660 Ci; the
“'N(n,p)"'C reaction will add about 5660 Ci of “*C if the nitrogen content of the graphite is 30 ppm.
The value of <200 Ci of ""C/GW(e)-yr shown in Table 8 for the HTGR corresponds to <9000 Ci

entering the fuel reprocessing plant cach year. These maxima include C in reflector blocks as well
as in fuel blocks. There is no metallic hardware in an HT'GR corresponding to cladding and other

structural components of the LWRs and LMFBRs.

6.1 Comparisons of Reactor Produced and Naturally Produced "C

The natural rate of ""C formation in the atmosphere from cosmic-ray induced reactions and the
contribution of "*C to the total radiation dose to man are valid bases for evaluating the impact of
reactor-generated quantities of this nuclide. Lingenfelter’' reported a global average production rate
of 2.5040.50 "C atoms cm ~ sec ' over the ten solar cycles prior to 1963. Reference has been made to
this value by Lal and Suess’ and in the UNSCEAR 1972 report.™ Using 5. 1E18 cm’ as the carth’s
surface area.” Lingenfelter’s value corresponds to (4.220.8)E4 Ci of "C yr. More recently, Light et
al.™ have calculated the average production rate from 1964 to 1971 to be 2.217+0.10 "C atoms
Table 8, Comparison of Y40 production in different types of reactors in units of i fow(e J=yr®

 

 

 

Cladding
and core ) '
In structural 1n coolsnt Total
Reactor fuel materials Calculated Cbserved calculated
BWR 43.3-60.4 4.7 8 °
Low value 3.0 57
Median value 17.6 . , - 7h
High value 46,3 111
PWR 30.5=41.6 5.0 6
Iow value 9.6 iy
Median value 18.8 _ 29
High value k9.5 %
- - . . - , o o
HTGR <190 nil N.A,
Median value 12.0 <200
C
IMFBR 12.8 nil N.A,
Low value 2.6 15
Median value 6.3 19
High value 18.5 31

 

aReactor paramefers pertaining to these calculations based on the ORIGEN program are as follows: BWR,
18.823 MW(t)/MTU, L years in reactor, to 27,500 MWA/MTU:; 2.6 wt % 2°°U; 33% thermal efficiency. PWR,
30.0 MW(t)/MTU, 3 years in reactor, to 33,000 MWA/MTU; 3.3 wt % 23°U; 33% thermsl efficiency. HIGR,
L MW(t)/MIEM, L years in reactor, to 95,000 MW4/MTU; 38.5% thermal efficiency; see lable 5 Tor fuei
compositions. IMFBR, 30.18 MWw(t)/MTHM {mass average), 75% on-stream time for 3 years, to 24,800
MWA/MTU (mass average); 35% thermal efficiency; see Table 6 for fuel-region specifications.

bA velue of 9.1 Ci/GW(e)-yr is presented in the following report, issued as the present report was in
the final stage of preparation: R. L. Blanchard, W. L. Brinck, H. E. Xolde, H. L, Krieger, D. M.
Montgomery, S. Gold, A. Martin, and B. Kahn, Radiological Surveillance Studies at the Oyster Creek
BWR Nuclear Generating Station, USEPA, EPA-520/5-T76-003 (June 1976).

Ҥ.4. = not applicable.

61
20

e sec . Based on projections of sunspot numbers for the remainder of the solar cycle, they also
estimate that the 11-yr mean rate could be as large as 2.2840.10 *C atoms cm ° sec '. (The error
limits on the rates apply only to the statistics of the calculation.) This value corresponds to
(3.8+0.2)E4 Ci of "C/yr. Thus, to one significant figure, the 1l-yr average natural rate of
production is 4.E4 Ci of "'C/yr. On this basis, the quantity of "*C in fuel annually entering an LWR
fuel reprocessing plant with a capacity of 1500 MTHM/yr [equivalent to 45 GW(e)-yr and about
fifty 1000 MW (e) reactors] is | to 5.5% of the natural production rate; corresponding values for '*C
entering an LMFBR fuel reprocessing plant are 0.3 to 2.09 of the natural production rate. The 1660
(from graphite only) to 9000 (from graphite, oxygen, 25 ppm of nitrogen in fuel, and 30 ppm of
nitrogen in ali graphite) Ci of "*C annually entering the HTGR fuel reprocessing plant, of the same
45 GW(e)-yr equivalent capacity, corresponds to 4 to 22% of the natural rate of production of this
nuchde.

6.2 Worldwide and Local Radiation Doses from
Reactor-Produced "C

World population radiation doses from all forms of radiation and from naturally produced '*C
provide a second form of comparison of the effects of discharge of this nuclide from fuel
reprocessing plants. World-wide dose rates to gonads, bone-lining cells, and bone marrow due to
internal and external irradiation from all natural sources in “normal” areas are about 90 mrad/yr
(Table 20 of ref. 54, UNSCEAR 1972). Oakley’ reports a gonadal dose equivalent to the
population of the United States from all natural sources of 88 mrem/yr. The contribution of “C to
this total is about 0.7 to 0.8 mrad/yr.™ Other values of the contribution of '*C to the total have been
as high as 1.6 mrem/yr.'""”™ Thus, based on the percentages histed above and a nominal | mrem/yr
due to natural "“C, after this nuclide becomes uniformly distributed over the earth, additional
radiation doses due to "*C will be in the range 0.004 to 0.06 mrem/yr for discharges from an LWR
fuel reprocessing plant of capacity equivalent to 45 GW(e}-yr. corresponding incremental doses due
to "C discharges from equivalent LMFBR and HTGR fuel reprocessing plants will be in the range
0.0004 to 0.023 mrem/yr and 0.035 to 0.19 mrem/ yr, respectively.

Potential radiological impacts of annual releases of 5000 Ci of "*C on the population out to
S0 miles from a fucl reprocessing plant have been analyzed by Killough et al.”” Three techniques for
reducing these local population doses were: (1) use of a discharge stack up to 1000 ft tall; (2) heating
of the discharged gas to obtain a large effect of buoyancy to increase the cffective stack height: and
(3) use of nocturnal, rather than continuous, emissioni in order to minimize the availability of the
discharged "‘C for uptake by vegetation. Using metecorological data for the Oak Ridge, Tennessee,
area and a 300-ft stack, the total-body dose of a population of 10" people within the 50-mile radius
was 110 person-rem/ yr; the average individual dose was 0.107 mrem/yr, and the maximum dose to
“fence-post man” (who spends all his time at 1.5 miles from the stack and eats food grown only at
this location) was 240 mrem/ yr.

6.3 Other Predictions of ""C Formation Rates

Tahle 9 summarizes predictions of "'C formation rates in BWR and PWR fuels presented in this
*7 Calenlated formation rates in BWR fuels range from 13.6 to 22 Ci/ GW(c)-yr.
In the BWR coolant. from the ' O(n.a) reaction only. the range 15 4.7 to 9.9 C1, GW(e)-yr.

and other reports.
) Iy
Table 9. Comparisons of some estimates of C production rates? in LWRs
(values are in Ci of '*C/GW(e)-yr)

 

Source of information

 

 

 

Region
Reactor of %C Parent Bonka Kelly Fowler This
type formation nuclide et al. et al.© NUREGd et al.® report
BWR | Fuel RN 12.9 16,9 we® 18. . 11.5
70 8.4 2.7 NC 4. 3.3
Yo+ 0 21.3 13.6 NC 22. 14.8
Coolant AN 1.3 NC NC 0.26 NC
17 9.9 NC 9.5 8.9 B
PWR Fuel 14y 12.2 10.9 NC 18. 12.2
‘o 7.1 2.7 NC b 3.5
1'%y + 170 19.3 13.6 NC 22. 15.7
Coolant 1Ay 1.28 NC NC G.09 NC
179 9.8 NC 8 3.2 5.0

 

8Baged on 20 ppm nitrogen (by weight) in the UO; except for Bonka et al.,60 whose basis is not given.
bref. 60.

CRef. 61.

Parameters in ref. 62 for the BWR and in réf. 63 for the PWR correspond to about 0.9 GW(e)-yr. ~Thus,

values in this column, which are taken from these references, should be increased about 107%.
€Ref. b4.
Calculations pertaining to teg produced in the BWR cooling water are based omn the assumpltion that there

gis noe void volume in the core due to steam.
NC means not calculated.

1¢
22

Corresponding values in PWR fuels also range from 13.6 to 22 Ci/GW({e)-yr, and in PWR coolant
they range from 3.2 to 9.8 Ci/GW(e)-yr. Carbon-14 formsation rates in cooling water from the
“N(n,p) reaction are small and uncertain, since data on concentrations of nitrogen are nearly
nonexistent. When the uncertainties in cross-section data are coinbined with the varying choices of
other nuclear parameters used by these different authors, it 1s perhaps not unexpected that the
largest values are about twice the smallest.

Bonka et al.*’ give "*C production rates from nitrogen in the fuel and coolant of LWRs. These
authors list the 2200-m/sec cross sections for the "C(n,y)"*C, “N(n.p)"*C, and '"O(n,a)""C reactions
without stating whether they used these or cross sections collapsed according to reactor fluxes. They
also do not indicate the nitrogen content of the fuel or cooling water. Thus, 1t i1s not possible to
comment on the agreements and differences between the values of Bonka et al.* and those of other
authors listed in Table 9.

Kelly et al.”' give '*C production rates 5 to 239 lower than values in this report (Table 9). These
authors also present only the 2200-m/ sec cross sections for reactions 1, 2, and 5, they do not discuss
collapsing cross-section data in terms of the fluxes of specific reactors. Again, no comparison can be
made between their model reactors and those of this report.

The U.S. Nuclear Regulatory Commission (NRC) has presented an estimate of 9.2 Ci of "C/yr
formed in the cooling water of a BWR® and of 8 Ci/yr in the cooling water of a PWR.** Both
values are based only on the "O(n,a)"*C reaction; formation of "*C from the "*N(n,p) reaction is
considered to contribute only a siall fraction of 1 Ci/yr because of the low concentration of "N in
the reactor coolant (less than | ppm by weight). The calculational procedure of the NRC reports
includes use of an average flux of 3.0E+13 neutrons cm ’ sec ' and a thermal neutron cross section
for 'O of 0.24 b for both BWR and PWR; the masses of water in the reactor cores are 39 and 33
MT, respectively. The product of flux and cross section corresponds to 7.2E-12 atoms of "“C per
second per atom of '"O.

Fowler et al.”® wrote a technical note partly to elicit comments concerning EPA calculations of
“C source terms and the radiological impact of this nuclide. The EPA has already published®’
proposed standards pertaining to releases of “Kr, '“’1, and certain long-lived transuranic nuclides
from nuclear power operations; no standard pertaining to '*C was proposed, because the knowledge
base available (in 1975) was considered inadequate for such a proposal. Calculations in the technical
note are based on assumptions of a flux of 5.0E+13 neutrons cm * sec”', an effective cross section of
1.1 b for the l4N(n,p)"’C reaction, and an effective cross section of 0.14 b for the ”()(n,a)C”
reaction, for both the BWR and the PWR. This choice of flux and cross sections corresponds to
5.5E-11 atoms of "*C per second per atom of nitrogen, and 7.0E-12 atoms of "*C per second per
atom of 'O, respectively, for both the BWR and the PWR. These authors™ also calculated a source
term for "'C formation from | ppm of nitrogen dissolved in the cooling water. This use of | ppm is
arbitrary since essentially no data are available on this concentration at operating reactors, as
discussed in Sect. 3.3. The calculations with | ppm of nitrogen were made because similar sample
calculations had been made in draft regulatory guides.”"’
in refs. 62 and 63 which were developed from these drafts.

Calculations in this report are based on parameters listed in footnote a of Table 8 and in
Sect. 3.1. From the effective fission cross sections (p. 72, Table A-l, of ref. 1), the ORIGEN code
calculates average fluxes of 2.07FE+13 and 2.92E+13 neutrons cm ~ sec ' for BWR and PWR,
respectively. However, the initial and final fluxes for the BWR are 2.00E+13 and 2.26E+13. and
initial and final fluxes for the PWR are 2.58E+13 and 3.456+13 neutrons cm - sec . The average
formation rates for a BWR are, therefore, 3.06E-i1 atoms of ''C formed per second per atom of *N

However, such calcuiations are not made
23

present and 3.79E-12 atoms of “C formed per second per atom of 'O present; corresponding values
for a PWR are 4.32E-11 and 5.34E-i2. Thus, the "*C formation rates calculated in this report for the
“N(n,p) reaction are only 55% (for the BWRY and 799% (for the PWR) as large as values presented
by Fowler et al* Carbon-14 formation for the "O(n,a) reaction rates in this report are only 53%
(for the BWR) and 74% ({for the PWR) as large as values in refs. 62 and 63, they are only 54% (for
the BWR) and 76% (for the PWR) as large as values in ref. 64.

Cross sections listed in Table 1 are the current best estimates for application to the steady state
of reactor operations {after the first few reloads). The most recent (1974) revisions (soon to be
incorporated in the ORIGEN library) of '“N cross sections for use in the ENDF/B-1V library™ WL(C
prc-»ented by Young, Foster, and Hale,” largely from an earlier revxew by Young and Foster.”
Croff” has used this revision and the XSDRNPM c.omputcr program’' to obtain & one-group value
of 1.45 b for the effective thermal cross section for the “N(n,p)"*C reaction for LWRs. This is very
close to the value 1.48 b used in this report.

6.4 Comparison with Releases from Russian Reactors

Rublevskii et al.”* have presented data, listed in Table 10, on measured releases of '*C from five
Russian reactors. These authors combined their data with Spinrad’s’” projections concerning
world-wide installed nuclear power to estimate the magnitude of *C discharges to the year 2010.
Neglecting the small Obninsk and ARBUS reactors, the data in Table 10 show releases ar the
reactor stations of 200 to 800 Ci of “C/GW(e)-yr. These values are far in excess of the
6 Ci/ GW(e)yr reported by Kunz et al.’' for the Ginna, Indian Point 1, and Indian Point 2 PWRs,
and of the 8 Ci/GW(e)-yr for the BWR at Nine Mile Point.”” The reported releases of "C from
Russian reactors are thus seen to be about of 10 to 100 times greater than corresponding releases
from the four-mentioned American reactors. Such a discrepancy implies that Rublevskii et al.”* have
grossly overestimated the potential releases of “C from non-Russian nuclear reactors, and that a
need exists for an analysis of the origin of "“C formation in the Russian reactors. This
overestimation appears in their conclusions that the daily production rates of “C in water<ooled,
graphite moderated reactors and in water-cooled, water moderated reactors (LWRs) are 0.75 and
0.25 mCi/ MW(1), respectively. The latter value corresponds to about 300 Ci/GW(e)-yr, which 15 40
to 50 times greater than was observed by Kunz et al.''"" Apparently, a detailed description is not
now available. However, on visits to Russian nuclear stations, Lewin'* was advised that nitrogen gas
is used to blanket the graphite of the pressure-tube reactors, such as those at Beloyarsk and
Sosnovyi Bor (near Leningrad).””’® In addition, a pressurized water reactor VVER-210 at
Novovoronezh'™ (Tabie 10) has been reported’’ to use nitrogen gas for pressurization; finally,
hydrazine and ammonium hydroxide are used in the primary cooling water to minimize radiolytic
oxygen formation, and for corrosion and pH control. Later PWRs constructed at Novovoronezh do
not use nitrogen pressurization; instead, steam is heated electrically by a method similar to that used
in the PWRs in the United States. ™ *

6.5 Reduéing the Releases of *C

Releases of '*C can be reduced by reducing the amount that is formed in nuclear reactors, by
collecting it at the reactor station and at the fuel reprocessing plant and converting most of it to
solid form for permanent retention, or by a combination of these methods. Snider and Kaye™ have
: ( , . . i a
Table 10. Carbon-1Y4 entering the atmosphere with gaseous wastes from some Russian reactors

 

 

Power 14 140

Rated rating ) 7 .

thermal during discharged dlscharged_b

power studies mCi [ ci 1

Reactor type M (4)] (MW (%] day GWiej-yqj

Water—-cooled, graphite moderated APéi
USSR Academy of Science, Obninskd 30 12 9 *+ 3 900 * 300
Water—-cooled, grapbite'moderated {AMB),
Beloyarsk APS € 285 210 140 + 50 800 t 300
Water-cooled, water moderated (VVER-210),
Novovoronezh APSC {FWR)E 760 740 120 = 30 200 + 50
Water-cooled, water moderated (VK-50j, o
(Boiling water test reactor) Ulyanovsk APS 150 g0 30 £ 10 400 = 130
Organic moderated and cooled test
reactor {ARBUS) 5 5 0.6 £ 0.2 150 + 59

 

aSee ref, 72.
b

Based on an assumed thermal-to-electrical efficiency of 30%, as used in ref. 72.

C,. . C
APS = atomic power station.

é . L. . . . R . .
A pressure-tube reactor of which the two 1000 MW(e) units at Sosnovyi Bor {near Leningrad) are the most

modern counterparts.

eEquivalent to a United States pressurized water reactor.

vT
25

recently analyzed many process options and the effects on the environmental impact of “C releases.
Reducing the quantity of "*C formed requires that the nitride nitrogen impurity content of the fue)
be reduced, and that air be removed from each fuel rod in a vacuum degassing step before the
second end of the rod is closed by welding. Such reduction to a maximum of [0 ppm of nitrogen by
weight is a goal that one fuel manufacturer (1 of Table 3) has already achieved and that two fuel
manufacturers (2 and 3 of Table 3) could achieve without much technical or economic impact, but
which the other two could not easily achieve. When the nitrogen content is reduced to 5.7 ppm
(Sect. 3.1}, the guantity of “C formed from the "O(n,a) reaction equals that formed from the
"'N(n,p) reaction in LWR fuels. | |

Retaining carbon dioxide in nuclear fuel reprocessing plants is another alternative now being
investigated for minimizing discharges of '*C to the environment. The fluorocarbon absorption
process,”” now in the pilot plant stage of development for the recovery of krypton from the off-gas of
LWR and LMFBR-fuel reprocessing plants, also collects CO: in the fluorocarbon solvent. The CO,
so collected could be discharged into a slurry of Ca(OH)." and converted to CaCO: for permanent
storage. Similarly, the KALC process™ ™' (Krypton Absorption in Liquid Carbon Dioxide) to
recover and retain krypton in the carbon dioxide gas stream of an HTGR fuel reprocessing plant is

also n the pilot plant stage of development. The "C-containing carbon dioxide of this process could
p p

also be converted'* to CaCQs.
1.

26

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76.
e
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79-

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Auerbach
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Blanco
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S. Bomar

D. Bond
Brooksbank
Brown
Burch

. Campbell

Jr.

. Culler
C. Dahlman
Davis, Jr.
K. Emanuel
G. Iee .

I.. Feldman
E. Ferguson
C. Finney
J. Frederick
Fulkerson
W. Godbee
Goeller
Grimes
S. Groenier
F. Harris
C. Haws
Hibbs

S. Hill
Hoffman
Houser
R. Irvine
R. Kasten
V. Kaye

L. Keller
G. Killough
Lewin

H. Lin:

L. Lotts

0

. C. Coutant
L
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33

INTERNAL DISTRIBUTION

80. A.
8l. J.
682, W.
3. L.
gL, 7.
85. L.
86. M.
87. XK.
88, A.
9, J.
90. H.
91. H.
92. D.
93. C,
ok, .
95. J.
96. P.
97. M.
98. A,
99. C.
100.  J.
101. E.
102. V.
103, J.
10k, D,
105, W.
106, P,
107. V.
108, B.
109,  J.
110. R.
111. W,
112. K.
113. ¢,
11k, R,
115-116.
117-118.
119-~120.
121.
122.

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ORNI/ NUREG/ TM~12
Dist. Category UC-11

Malinauskas

. McBride

MeClain
McNeese
Morrison, ORGDP
Morse
Myers
Notz
Olsen
Olson
Pfaderer
Traa
Reichle
Reliser
Richmond
Roddy
Rohwer
Rosenthnal

. Ryon
. Scott

Snider
Struxness
Tennery
TilL
Trauger
Unger
Vanstrum

. A. Vaughen

Vondra

. Witherspoon

Wyme r

Davis (consultant)
Gaden, Jr. (consultant)
Tece (consultant)
Richards (consultant)

Central Research Library

Document Reference Section
Iaboratory Records
laboratory Records
ORNL Patent Office

- RC
123.
12k,
125,
126.

127-151.
152,
153.
1.5k,
155,
156.
157,
158.
159,
160,
161.
162,
163.
164,
165.
166.
167.
168.
169,

170,

34

EXTERNAT, DISTRTBUTION

A. L. Ayers, Allied General Nuclear Services, P. 0. Box 847,
Barnwell, SC 29812

D. A. Baker, Radiological Health Research, Pacific Northwest
Laboratories, P.0. Box 999, Richland, WA 99352

N. ¥. Barr, Energy Research and Development Administration,
Washington, D. C. 20545

S. Beard, Exxon Nuclear Co., Field Box 3965, San Francisco,
CA 94119

R. M. Bernero, U.S. Nuclear Regulatory Commission,
Washington, D. C. 20555

M. B. Biles, knergy Research and Development Administration,
Washington, D.C. 20545

W. P. Bishop, U.S. Nuclear Regulatory Commission,

Washington, D. C. 20555

W. Broeker, Lamont-Doherty Geological Observatory, Palisades,
NY 1096k,

L. H. Brooks, General Atomic Co., P. 0. Box 81608, San

Diego, CA 92138

J. A. Buckham, Alljed Chemical Corporation, 550 Second
Street, Idaho Falls, ID 83hOl

L. L. Burger, Pacific Northwest Laboratory, Box 999, Richland,
WA 99352

L. Burris, Jr., Argonne National Laboratory, 9700 S. Cass Avenue,
Argonne, IL 60439

A. B. Carson, General Electric Co., 175 Curtner Ave., San Jose,
CA 95100

K. R. Chapman, U.S. Nuclear Regulatory Commission,
Washington, D. C. 20555

R. B. Chitwood, U. S. Nuclear Regulatory Comnission,
Washington, D. C. 20555

B. L. Cohen, Univ. of Pittsburgh, Dept. of Physics, Pittsburgh,
PA 15261

Jd. T. Collins, U.S. Nuclear Regulatory Commission, Washington
D. C. 20555

R. D. Cooper, Energy Research and Development Administration,
Washington, D. C. 20545

W. T. Crow, U.S. Nuclear Regulatory Commission, Washington,
DC 20555

R. E. Cunningham, U.S. Nuclear Regulatory Caommission,
Washington, D. C. 20555

J. H. Davis, Tennessee Valley Authority, River Oaks Bldg.,
Musele Shoals, AL 35660

Directorate of Health Protection, Commission of the European
Commanities, Luxemboury

G. G. Eichholz, Georgia Institute of Technology, School of
Nuclear Engineering, Atlanta, GA 30332

M. Eisenbud, New York University Medical Center, 501 First
Avenue, New York, NY 10016
171,
172,
173.
17k,
175.
176.
177,
178.
179~180.
181,
182.
183.

18k,
185,

186.
187,
188.
189.
190-191.
192.
193-194.
1G5,
196.
197.
198.

35

E. A. Evans, Hanford Engineering Development Iab., P. 0. Box 1970,
Richland, WA 99352 4

J. F. Fletcher, Hanford Engineering Development ILab., P. 0. Box
1970, Richland, WA 99352

R. F. Foster, Pacific Northwest Laboratory, Box 999, Richland,
WA 99352 |

T. W. Fowler, Envircnmental Protection Agency, 401 M 8t., S.W.,
Washington, D. C. 20460

R. E. Franklin, Energy Research and Development Administration,
Washington, D. C. 20545

F. Gera, C.N.E.N., Viale Regina, Margherita 125, 00198,

Rome, Italy

F. A. Gifford, Atmospheric Turbulence and Diff. Lab-NOAA, P. 0.
Box E, Oak Ridge, TN 37830

H. Gitterman, Burns and Roe, Inc., Industrial Division, P. O.
Box 663, Paramus, NJ 17652

S. D. Harkness, Combusflon kEngineering, Inec., Nuclear Power Dept.,
Windsor, .CT 06095

J. H. Harley, Health and Safety laboratory, Energy Research and
Development Administration, 376 Hudscon St., New York, NY 1001k
C. A. Heath, General Atomic Co., P. 0. Box 81608, San Diego,

CA 92138. -

R. L. Hirsh, Energy Research and Development Administration,
Washington, D. C. 20545

K. D. B. Johnson, A. E. R. E., Harwell, Didcot, Berks., England
J. Jordan, Climate Dynamics, National Science Foundationm,
Washington, D. C. 20550

B. Kahn, Georgia Institute of Technology, Environmental Resources
Center, Atlanta, GA 30332

C. D. Keeling, Scripps Institute of Oceanography, lLa Jolla,

CA 92037

W. W. Kellogg, National Center for Atmospheric Research,
Boulder, CO 80303

R. H. Kennedy, Energy Research and Development Admflnlbtxatlon,
Washington, D. C. 20545

G. R. Kilp, Westinghouse Electric Corp., Nuclear Fuel Div.,

Box 355, Pittsburgh, PA 15230

S. langer, General Atomic Co., P. 0. Box 81608, San Diego,

CA 92138

G. W. LaPier, Babcock and Wilcox, Nuclear M&terlals Div.,
Apollo, PA 156173

T. R. Lash, National Resources Defense Council, Inc., 66k
Hamilton Ave., CA 94301

H. Iawroski, Nuclear Services Corp., 1700 Dell Ave., Campbell,
CA 95008

W. H. Lewis, Nuclear Fuel Services, Inc., 6000 Executive

Blvd., Rockville, MD 20852

J. L. Liverman, Energy Research and Development Administration,
Washington, D.C. 20545
199.
200,
201,
202.
203.
20k,
205,

206.
207.

208,
209,

210,
211.

212,
213.
21k,

215.
216.

217,

218.

36

R. L. Loftness, Electric Power Research Tnstitute, 1750 New
York Ave., N. W., Washington, D. C. 20006

H. Lowenberg, U.S5. Nuclear Regulatory Commission,

Weshington, D. C. 20555

T.. Machta, Alr Research laboratory, NOAA, Silver Springs,

MD 20910

3. Manabe, Geophysical Fluid Dynamics Laboratory, NOAA,
Princeton, NJ 08540

B. J. Mann, Environmental Protection Agency, P. 0. Box 15027,
Jas Vegas, NV 8911k

M. 0. Marlowe, General Electric Co., Nuclear Energy Systems
Div., San Jose, CA 95125

J. B. Martin, U.S. Nuclear Regulatory Commission,

Washington, D. C. 20555

J. Matuszek, Department of Health, Albany, NY 12201

D. R. Miller, Energy Research and Development Administration,
Washington, D. C. 20545

W. A. Mills, Envirommental Protection Agency, Waterside Mall,
Washington, D. . 20L&60

F. McCormick, Ecology Program, University of Tennessee,
Knoxville, TN 30916

H. A. C. McKay, A.E.R.E., Harwell, Didcot, Berks., England
W. H. McVey, Energy Research and Development Administration,
Washington, D. C. 20545

W. 8. Nechodom, Exxon Nuclear Co., Inc., 2101 Horn Rapids Rd.,
Richland, WA 99352

R. I. Newman, Allied General Nuclear Services, P. 0. Box 847,
Barnwell, SC 29812

W. R. Ney, National Council on Radiation Protection and
Measurements, 7910 Woodmont Ave., Bethesda, MD 2001k

W. Niemuth, Exxon Nuclear Co., Inc., Richland, WA 99352

R. Nydal, Physics Dept. Technological Institute, Trondheim,
Norway

H. Oeschger, Physiecs Institute, Univ. of Bern, Bern,
Switzerland

I. U. Olsson, Institute of Physics, Uppsala Univ., Uppsals,
Sweden

W. S. Osburn, Energy Research and Development Administration,
Washington, D. C. 20545

C. L. Osterberg, Energy Research and Development Administration,
Washington, 0. C. 20545

F. L. Parker, Vanderbilt Univ., Dept. of Civil Engineering,
Nashville, TN 37235

G. I. Pearman, CSTRO, Division of Atmospheric Physics,
Aspendale, Victoria, Australia

A. M. Platt, Pacific Northwest Laboratory, Box 999, Richland,
WA 99352

G. B. Pleat, Energy Research and Development Administration,
Washington, D. C. 20545

A. J. Pressesky, Energy Research and Development Administration,
Washington, D. C. 20545
oo
0
™

T, Rafter, Institube of Nuclear Studies, Power Hutt, Hew
Zealand

7. Research and Technical Support Division, Cak Ridge Operationg
228, A. D. Riley, Allied Chemical Corporation, P. 0. Box 430,

Metropolis, IL 62960
229. A. P. Roeh, Allied Chemical Corp., 550 Second St., Idaho Falls,

D 83401 |
230-233. L. C. Rouse, U.5. Nuclear Regulatory Commission,
3

&2
™o

Waghington, D. C. 20555
k., W. D. Rowe, Environmental Protection Agency, MOL M St.,
S. W., Washington, D, C. 20460

235. J. L. Russell, Envirommental Protection Agency, 401 M
S$t., 3. W. , Washington, D. . 20460

236. E. J. Salmon, National Academy of Sciences, 2101 Constitution
Ave., Washington, D. D, 20418 |

23f. J. Schacter, General Offices, Union Carbide Nuclear Division,
Oak Ridge, TN 37830

238. K. G. Schiager, Umniv. of Pittshurgh, School of Public Health,
Pittsburgh, PA 15261

239. A. Schneider, Georgia Institute of Technology, School of
Nuclear Engineering, Atlanta, 0GA 30332 |

240, K. J. Schneider, Pacific Northwest Iaboratory, Box 999,
Richland, WA 99352

2hl. J. M. Selby, Pacific Northwest Iaboratory, Box 999, Richland,
WA 99352 | |

242, G. L. Simmons, Science Applications, Ine., 1200 Prospect
St., La Jolla, CA 92037

2k3. C. M. Slansky, Allied Chemical Corp., 550 Second St.,
Tdaho Falls, 1D 83401

2hl, B, Spinrad, Oregon State Tmiv., Radiation Center, Corvallis,
OR 97331

2h5. H. E. Stelling, Energy Research and Development Administration,
Washington, D. C. 20545

2h6., (. E. Stevenson Argonne National Laboratory, 9700 §. Cass Ave.,
Argonne, IL 60439

247. L. Stratton, Dept. of Public Healbh, 525 W, Jefferson,
Springfield, IL 62761

2kd. J. J. Swift, Environmental Protection Agency, 401 M St.,
8. W., Washington, D. C. 20460

249,  J. Swinebroad, Energy Research and Develomment Administration,
Washington, D. C. 20860

250. M. J. Seulinski, Atlentic Richfield Hanford Co., Federal Bldg.,
Richland, WA 99352

251. A. V. Seshadri, General Electric Co., Nuclear Energy Systems
Div., San Jose, CA 95125

252. J. J. Shefeik, General Atomic Co., P. 0. Box 81608, San Diego,
CA 92138

253, A. C. Stern, Dept. of Envirommental Sclences and Engineering,
Uhiv; of North Carolina, 602 Croom Court, Chapel Hill, NC
27514
38

254, K. G. Steyer, U.S. Nuclear Regulatory Commission,
Washington, D. C. 20555

255. H. Suess, Scripps Institute of Oceanography, La Jolla,
CA 92037

256, J. M. Taylor, Pacific Northwest Laboratory, Box 999,
Richland, WA 99352

257. L. S. Taylor, National Council on Radiation Protection and
Measurements, 7910 Woodmont Ave., Bethesda, MD 2001k

258, J. S. Theilacker, Westinghouse Electric Corporation,
P. 0. Box 158, Madison, PA 15663

259, M. T. Walling, Kerr-McGee Co., Oklahoma City, OK 73107

260. R, L. Watters, Bnergy Research and Development Administra-
tion, Washington, D. C. 20545

261. R. K. Weatherwax, Princeton Univ., Engineering Quadrangle,
Princeton, N. J. 08540

262, A, M. Weinberg, Osk Ridge Associated Universities, Oak
Ridge, TN 37830

263. W. Weinlander, GFK -~ Institute Heisse Chemie, Postbox 36L0,
D7500, Karlsruhe, West Germany

264, G.H. Whipple, Univ. of Michigan, School of Public Health,
Ann Arbor, MI 48104

265. A. K. Williams, Allied General Nuclear Services, P. 0. Box
87, Barnwell, SC 29812

266-531. Given distribution as shown in TID-4500 under Category

Uc-11 - (75 coples - NTIS)