ORNL-TM-0611.txt

 

OAK RIDGE NATIONAL LABORATORY

operated by
UNION CARBIDE CORPORATION %
for the

U.S. ATOMIC ENERGY COMMISSION
ORNL- TM- 611

COPY NO. - 9’5—

DATE - August 27, 1963

MASTER

 

INHERENT NEUTRON SOURCES IN CLEAN MSRE FUEL SALT

 

P. N. Haubenreich

 

 

 

ABSTRACT

 

Unirradiated MSRE fuel salt will contain an
appreciable neutron source due to spontaneous
fission of the uranium, and (@, n) reactions of
alpha particles from the uranium with the fluo-
rine and berylliium of the salt. The spontaneous
fission source in the core (25 f£t2 of salt) is
10® neutrons/sec. or less, mostly from U238, The
alpha-n source is much larger, giving sbout 4 x 10°
neutrons/sec. in the core. Nearly all of this

la’gzer source is caused by alpha particles from
yec4,

NOTICE

This document contains information of o preliminary nature and was prepared

primarily for internal use at the Oak Ridge Nationa! Laboratory. It is subject

to revision or correction and therefore does not represent o final report. The
information is not to bhe abstracted, reprinted or otherwise given public dis-

semination without the approval of the ORNL patent branch, Legal and infor-

mation Control Departmant,
 

 

 

LEGAL NOTICE

This report was prepared as an account of Government sponsored work. Neither the United States,

nor the Commission, nor eany person acting on behalf of the Commission:

A. Makes ony warranty or representation, expressed or implied, with respoct to the accuracy,
completensss, or usefulness of the information contained in this report, or that the use of
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privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of
any information, apparotus, method, or process disclosed in this report.

As used in the above, “person acting on behalf of the Commission' includes any employee or

contrector of the Commission, ot employee of such contractor, to the extent that such zmpioyee

or contractor of the Commission, or employee of such contractor prepares, disseminotes, or
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or his employment with such centractor,

 

 

qgg‘\.'
CONTENTS
Introduction==~wr-roeme e m e e e e e
Fuel Composition===-~===---rm-cccmmm e m e e
Spontaneous Fission Neutrons-----e----mccmcmcmcccc i e
Neutrons from Alpha-n Reacticng-=-----------w-e-cr-cmcmmemnnn--
Alpha Emission by Uraniume--=--cem-e-eereemera e e r e m e e e
Alpha-n Yields in Fuel Sglt----=-mcmmmmc e e e e
DiscuSSioN==m=mc- e e c o e e e e e e — -
Appendix — Calculation of Alpha-n Yields in MSRE Fuel Salt------
Dilution by Non-Productive Constituents of Fuel balt-------
Be® (@, n) Yield=-mmrmme e e oo e e
F12 (@, n) Yield=-==-=cmc e e e e e
117 (0, n) Yield=mm==m=mmmmmm e oo e e e

References

- v T wr S M T W e amm Sm e mw e e mw v G e e e e mm v el vl M SN T g M Gl e MDA AN EE ED G B NN B A NN N W RN e el WEoam wm m

 

\J

O ¢ 1 O\

10
12
12
1k
1h
14
16
»

Introduction

When a reactor is subcritical, the fission rate and the neutron flux
depend on the strength of the neutron source in the reactor due to various
reactions and the multiplication of these source neutrons by fissions in
the core. By supporting the fission rate in the subecritical reactor at
sufficiently high levels, a source performs several functions in reactor
operations. The source strength required for some functions is higher
than for others.

In all reactor fuels there is always a source of neutrons due to
spontaneous fission, but this source is relatively weak, particularly if
the fuel is highly enriched urasnium. In many reactor cores there is also
an inherent photoneutron source produced by interaction of gamma rays with
deuterium or beryllium in the cdre. This sourée is usually not significant,
however, until after fission product gamma sources have been built up by
power operation. . Therefore, in nearly all reactors an-extranebus neutron
source 1s inserted in or near the core. The MSRE is unusual in that the
fuel is a homogeneous fused salt in which alpha-émitting uranium is inti-
mately dispersed with large gquantities of fluorine and beryllium,both of
which readily undergo alpha-n reactions. Thus thefe is a strong alpha-n
source inherent in the MSRE fuel salt even before it has been irradiated.

It is conceivable that the source inherent in the MSRE fuel salt,
which is certain to be present whenever there is any chance of eriticality,
is strong enofigh-to satisfy some, if not all, of the requirements which
make an extraneous source necessary in most reactors. In determining
whether or not an extraneous source is required, it is necessary to pre-
dict the strength of the neutron source which 1s inherent in the clean
slat before the photoneutron source becomes important. The present report
describes this prediction. The question of source reQuirements will be

considered later, in a separate report.

 

Fuel Composition

The strength of the inherent neutron source depends on the composition

of the fuel salt and the isotopic composition of the uranium in it. Three
fuel salts, with compositions shown in Table 1, have been considered for
the MSRE. |

Uranium concentrations shown are for the initial critical experiment.
For power operation the uranium concentration will be higher by about 15%
(to compensate for control rods, xenon and other poisons).

The isotopic compositions shown for the uranium are the values used
in the criticality calculations., The U234 and U=3€ fractions are based on
typical analyses of uranium enriched in the diffusion plant to the indicated
U235 content.

The lithium composition is that of lithium actually on hand for fuel

salt manufacture.

Table 1. Tuel Salt Compositions

 

 

 

 

Fuel Type A B C
Sglt comp: LiF® 70 66.8 65
(mole %) BeFo 23.7 29 29.2
ZrF, 5 L 5
ThF, 1 0 0
UF, 0.313 0.189 0.831
U comp: =34 0.3
(atom %) =33
U235 0.3
UESB 5 5 & .L‘_
Density at 1200°F 14k .5 134.5 2.7
(1b/£t£3)

 

299.9926 % 1i7.

Spontaneous Fission Neutrons

The rate of neutron production by spontaneous fission is a specific
property of the each nuclide. In the clean MSRE fuel, U®®® has the shortest

half-life for spontaneous fission.* (See Table 2.)
Table 2. Neutron Production by
- Spoataneous Fission of '
Uranium Isotopes

 

Isotope Specific Emission Rate-

 

(n/kg — sec)
ges4 6.1
U=3s 0.51
[ 5.1
ye3s 15.2

 

The effective core of the MSRE (the’graphite-containing region plus
some of the fuel in the upper and lower heads ) contains o5 ££3 df fuel salt.
The amounts of each uranium isotope and the spontaneous fission neutron
source in this.Volume are given in Table 3 for each of the three fuels

described in Table 1.

Table 3. Spontaneous Fission Neutron Source in Core

 

 

 

 

| Fuel A Fuel B Fuel C
Isotope
M, (kg) S(n/sec) M (kg) S(n/sec) M, (kg) S(n/sec)

ye34 0.3 2 0.2 1 0.2 1

Uess 27.0 14 16.5 8 26.4 13

ysss 0.3 2 0.2 1 0.2 1

g=s38 1.5 o0 0.9 13 47.5 722
4o 23 737

 

- Neutrons From Alpha-n Reactions

Energetic alpha particles can produce neutrons by nuclear interactions
with several'different nuclides. Threshold energies vary widely, depending
upon the nuclide. Three nuclides, Li7, Be® and F12, have a-n thresholds
below the maximum energy of alphas from uranium. The neutron yield per
alpha particle}is a function of the initial energy of the alpha particle

and the composition of the medium in which it is slowing down.
Alpha Emission by Uranium

 

Among the uranium isotopes present in fresh MSRE fuel, UZ3% has by
far the highest specific alpha emission and also emits the highest-energy
alpha particles. The specific alpha sources are summarized in Table TR
Table 5 gives the total alpha scurce in the effective core of the MSRE
during the initial critical experiment (25 £t of salt, containing the

amounts of uranium shown in Table 3)}.

Table 4. Alpha BEmission by Uranium

 

 

Tsotope Half-Life for  Decay Rate Ea f Q Source
P a-decay (y) (ais/sec-kg) (Mev) (&/100 dis.) (a/sec-kg)
U234 2.48 x 103 2,28 x 101 L4.77 72 1.64 x 10%L
L.72 28 0.64 x 10%%
Uess 7.13 x 108 7.9 x 107 L. 58 10 0.79 x 107
L.y 3 0.24 x 107
4.40 83 6.56 x 107
.20 L 0.32 x 107
yess 2.39 x 107 2,35 x 10° L.50 73 1.72 x 10°
h.45 27 0.63 x 10°
=38 L.51 x 10° 1.23 x 107 L.19 T7 0.95 x 108
h.15 23 0.28 x 106

 

Note: Ea is the initial energy of the alpha particle and f is
the percentage yield of alphas of that energy in the natural alpha
decay of the nuclide.

Table 5. Alpha Source in MSRE Core

 

Source Strength (a/sec)

 

Isotope E, (Mev)

 

Fuel A Fuel B Fuel C
234 b7 L.7h x 10©  2.90 x 101° 3.71 x 10%°
L.72 1.85 x 10©  1.13 x 101 1.45 x 101°
=35 4 .58 2.13 x 108 1.30 x 108 2.09 x 10®
L4 0.65 x 108 0.40 x 108 0.63 x 10®
4.40o 17.7 x 108 10.8 x 108 17.3 x 10®8
4.20 0.86 x 108 0.53 x 10%8 0.85 x 108
Usse L .50 5.01 x 10° 3.06 x 108 3.92 x 108
L.Lh5 1.83 x 108 1.12 x 10%° 1.44 x 108
=38 4.19 1.40 x 108 0.86 x 10° 0.45 x 108
L.15 0.41 x 10° 0.25 x 108 0.13 x 10®8

 
Alpha-n Yields in Fuel Salt

 

The yields of neutrons from Be®, F®, and Ii” vary with the energy of
the alpha-particle, generally increasing with energy. Yields for 4.77-Mev
alpha-particles in thick targets of pure material are 40, 6 and 0.1 neutrons
per million alpha-particles in beryllium, fluorine, and lithium-7, respec-
tively. In the MSRE fuel salt, the productive nuclides comprise only a
fraction of the total, and the yield is affécted by the dilution with other
elements. Yields for'alpha-particles of each energy in Table 5, in each
of three fuel salts, were calculated by proéedures déécribed in the Appendix.
Table 6 illustrates how beryllium, fluorine, and lithium contribute to the
total yield for the most numerous and highest-energy group of alpha parti-

cles.

Table 6. Neutron Yields for L.77-Mev
Alpha Particles in MSRE Fuel Salt

 

Yield (n/10%/a)

 

 

Constituent
Fuel A Fuel B Fuel C
Be 2.65 3.32 3.20
F - L.36 by I, Lo
Li ~ 0.02 0.02 0.02

~ Total = 7.03 7.78 T.62

 

Table 7 gives the neutron source in the éffective core of thé MSRE

when the uranium concentration is at its Initial, clean, critical value.
10

Table 7. Alpha—n Neutron Scurce in Core

 

Alpha Neutron Spurce Strgngthr(n/sec)

 

 

 

 

 

E_ (Mev)
Source O Fuel A Fuel B Fuel C
ya34s b .77 3.33 x 105  2.26 x 105  2.83 x 105
h.72 1.21 x 10° 0.82 x 10° 1.03 x 10°
U235 4.58 1.14 x 10 0.78 x 10®  1.23 x 10°
b7 0.30 x 10° 0.21 x 10° 0.32 x 103
4. 40 7.54 x 10%- 5.20 x 10° 8.13 x 103
k.20 0.28 x 10° 0.20 x 103 0.31 x 103
U238 L.50 2.45 x 103 1.68 x 103 2.10 x 103
4 .45 0.83 x 103 0.57 x 103 0.72 x 103
y=38 4.19 L.36 3.07 157
©1.15 1.21 0.85 43
Total L.67 x 10° 3.17 x 107 3.99 x 10°
Discussion

The calculations indicate that the bulk of the neutron source inherent
in the clean MSRE fuel is due to alpha-n reactions, with spontaneous fis-
sion contributing relatively little. Furthermore, about 97 percent of the
neutron source is caused by alpha particles from a single isotope, U=3%,
which comprises a very small fraction of the total uranium. Therefore, the
neutron source will be very closely proportional to the UZ3% content of the
fuel sait.

In natural uranium, the abundance of U3* is only 0.0057%, or 0.0079
of the U235 gbundance. In a gaseous diffusion plant, however, the UZ34/y=35
ratio is increased, so that in uranium containing over 90% U=3% the
U234 /U235 ratio is 0.010 or above.

The U3* fractions which were used in the calculations are based on
typical analyses of enriched uranium, and thus are only estimates of what
will appear in uranium which will be used in making up the MSRE fuel salt.
The estimate is probably good to within +20% in the case of Fuels A and B,
which use highly enriched uranium. In the case of Fuel C it was assumed

that the uranium would be taken from the diffusion plant at about 35% U=3°,
11

and that the U234 content would be only 0.30%. It now appears that the

uranium may be added to the MSRE fuel salt in two batches: the first of
natural or depleted uranium; the second, highly enriched. If this course
is followed, the U23* content of Fuel C would probably be higher, perhaps
by as much as & factor of 1.4. The neutron source for Fuel C would then

be higher by the same factor.
12
APPENDIX
Calculation of Alpha-n Yields in MSRE Fuel Salt

Information on alpha-n yields from various nuclides usuvally appears
in one of two forms: 1) the microscopic cross section of the nuclide for
the a-n reaction as a function of alpha energy, or 2) the yield of neutrons
per million alpha particles of a given initial energy emitted in an infi-
nite medium of the pure nuclide. If the alpha particles are emitted in a
mixture, it is necessary to take into account the dilution of the produc-

tive nuclides by others which only slow down the alpha particles.

Dilution by Non-Productive Constituents of Fuel Salt

The correction for the dilution of a productive nuclide in a mixture
is essentially the fraction of the alpha energy loss which is attributable
to the productive constituent.

Let Doax be the yield of neutrons for alpha particles emitted in an
infinite medium consisting entirely of a productive nuclide. ILet n be the
yield for that nuclide in a mixture. It has been observed=’> that a fairly

good approximation is

 

n Nb Sp
T T o (1)
max j_Ni Si

where S is the 'relative atomic stopping power', N is the number density
of a nuclide, and p refers to the productive constituent.

The best information on relative stopping powers is still a 1937
article by Livingston and Bethe.* They give S relative to air for 16
elements for 6-Mev alpha-particles and for 6 elements at 7 other energies
from 2 to 52 Mev. Table 8 gives values of S for the constituents of the
MSRE fuel salt, obtained by interpolation in energy and atomic number of
the Livingston - Bethe data. The relative stopping powers in this table
are evaluated at 4.5 Mev, because this is approximately the energy of the

uranium alpha-particles.
13

Table 8. Relative Stopping Power of
Constituents of MSRE Fuel Salt
for 4.5-Mev Alpha Particles

 

 

 

. NS/),N; S3
- Constituent S -
' ' "Fuel A TFuel B  Fuel C
S Li 0.57 0.163 0.159 0.149
Be . 0.70 0.068 0.085 0.082
F 1.19 0.692. 0.705 0.699
Zr 2.8 0.057 0.0Lk7 0.056
Th 3.9 0.016 0 0
U ‘ .2 0.005 0.003 0.014

 

”?Atomic stopping power relative to air.

If the microscopic cross section of & nuclide for the alpha-n reaction
is known, then the number of neutrons produced by an alpha particle can be

found from®

EC)
.- N_o(E)
n = f(_Ei—:E_:T dbE (2)
O d‘X,

1 ag\~t :
5 a;)- as a function of alpha energy for several

different substances. For a mixture one may assume that

Harris! has presented

aE 1 4E
T & - Puix Z‘“i("'p'dx : | (3)
1

where w, is the weight fraction of constituent i in the mixture. Table 9
gives'values of % %g for the constituents of the MSRE taken from refer-

ence 1, and the products of this quantity and the weight fractions for
the three different fuel salts. The sum at the bottom of each column is

L i for each salt.
p_ .. Ox
mix

 
1k

Table 9. Slowing-Down Parameters for Alpha Particles

 

1 dE\  Mev wi (} 1ahy Mev
"5 ) e > &x ), 5/

 

 

 

 

. p d.X i
+ Fuel A Fuel B Fuel C
4 Mev 5 Mev
4 Mev 5 Mev 4 Mev 5 Mev 4 Mev 5 Mev

Li 885 781 103 91 108 95 96 85
Be 840 741 43 38 57 50 53 L7
F 730 €5 L7k 419 513 153 L89 432
Zr 385 351 Lo 38 37 33 L2 39
Th 228 208 13 12 0 0 0 0
U 222 202 h Y 3 2 10 9

679 602 718 633 690 612

 

Be® (o, n) Yield
For alpha particles with an initial energy EO, emitted in pure beryl-
2,3

 

Jium

n = 0.152 E 358 neutrons/10% « (4)
max o

For Fuels A, B, and C, n/nmax is 0.068, 0.085 and 0.082 respectively.

(See Table 8). The product is the yield in the fuel salt which is tabu-

lated in Table 10,

F¥° (o, n) Yield

Segre and Wiegand® measured neutron yields for alpha particles of
various energies in thick targets of F. The yield, LN begins to be
measurable at 3 Mev and rises to 10 neutrons/10® alphas at 5.3 Mev. From
Table 8, n/nmaX for Fuels A, B and C are 0.692, 0.705 and 0.699. The pro-
duct of this n/nmax and Noox from the data of Segre and Wiegand is given

in Table 10.

Li” (@, n) Yield

This writer knows of no direct measurements of nmax for Li7. The

 

cross-section for the Li” (&, n) B reaction as a function of alpha energy
15

was calculated and reported by Hess.® Above a threshold at 4.36 Mev, the
cross-section rises to 8 mb at 4.8 Mev, then decreases to about & mb at
higher energies. Hess' cross-section was used to compute yields from Li7
in the fuel salt from equations (2) and (3). The integral in Eq. (2) was
evaluated by fepresenting the cross section curve by straight-line segments
and‘by‘approximating - % ;“ %ED-l vs. B by linear relations fitted to
points at 4 and 5 Mev ginQXin Table 9. Results appear in Table 10.

Table 10. Neutron Yields for Alpha Particles in MSRE
Fuel Salt (n/108 )

 

 

 

By Fuel & =~  Fuel B Fuel C
(Mev) 5o 5 Ii Be F Ii Be F Li
h.77  2.65 L4.36 0.019 3.32 L.4h 0.019 3.20 L4.40 0.017
L2 2.60 3.9% 0.016 3.25 L4.02 0.016 3.13 3.98 0.01k
L.58 2.31 3.0k 0.008 2.89 3.10 0.008 2.79 3.08 0.007
4.50 2.18 2.70 0.005 2.72 2.75 0.004 2.62 2.73 0.004
L.y o 2.12 2.56 0.003 2.65 2.61 0.003 2.% 2.59 0.003
Lis 2,10 2.42 0.002 2.63 2.47 0.002 2.53 2.45 0.002
L.ho 2.01 2.25 0.00. 2.52 2.29 0.001 2.43 2.27 0.000
h.20 1.70 1.52 0 2.13 1.55 0 2.05 1.54 0
L.19 1.67 1.45 0 2.09 1.48 0 2.02 1.47 0
4,15  1.63 1.31 0 2.04 1.34 0 1.97 1.33 0

 
16

References

1. D. R. Harris, "Calculation of the Background Neutron Source in
New, Uranium-Fueled Reactors,” USAEC Report WAPD-TM-220, Bettis Atomic
Power Iaboratory, March 1960.

2. A. O. Hansen, "Radiocactive Neutron Sources," p. 3 in Fast Neutron
Physics, Part 1, ed. by J. B. Marion and J. L. Fowler, Interscience, New
York, 1960.

3. 0. J. C. Runnalls and R. R. Boucher, "Neutron Yields from
 Actinide-Beryllium Alloys," Can. J. Phys., 34: 949 (1956).

L. M. 8. Livingston and H. S. Bethe, "Nuclear Dynamics, Experimental,"
Revs. Mod. Phys., 9: 272 (1937).

5. W. N. Hess, "Neutrons from (@, n) Sources,' Annals of Phys., 2:
115233 (1959).

6. E. Segre and C. Wiegand, "Thick-Target Excitation Functions for
Alpha Particles," USAEC Report LA-136, Los Alamos Scientific Laboratory,
September 194L (also issued as USAEC Report MDDC-185).

 

 
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