JAERI-Conf 95-021

PB40
COMPARISON OF CALCULATED SHIELDING EFFECTS FOR 8 MATERIALS
N. Nakanishi, T. Shikata*, and S. Fujita
RIKEN ( The Institute of Physical and Chemical Research ) ,Hirosawa 2- 1, Wako- shi, Saitama 351 - 01 Japan
andT. Kosako
Research Center for Nuclear Science and Technology,
The University of Tokyo,
Yayoi 2-11-16, Bunkyo-ku, Tokyo 113 Japan
2. Calculation Procedure

Abstract
The depth dependence of neutron and secondary
photon dose equivalents in the inside of 4 meter thickmaterials of iron, lead, ordinary concrete, heavy
concrete, graphite, marble, water and paraffin were
calculated for monoenergetic source neutrons ofenergies less than 400 MeV. Their shielding
characteristics arecompared and discussed
phenomenologically. Calculations were carried out by
using the one-dimensional discrete ordinates code
ANISN-JR and the cross section library DLC87 /HILO. Systematic knowledge concerning theshielding materials was successfully obtained.
1. Introduction
In recent years, the radiation shielding of highenergy accelerators has become one of the important
problems as the particle energy and/or the beam
intensity get higher. In order to achieve effective
shielding, it is necessary to know about the shielding
characteristics of various materials for radiation,
especially neutrons.
So far, iron, lead, ordinary concrete and water are someof the most widely used shielding materials, and theirshielding characteristics have been studied extensively
for source neutrons in the wide energy range. Recently,
comparison of shielding characteristics between 8materials was given for dose equivalents at the exit
surfaces of shields of varying thickness I). (Hereafter theordinary concrete is abbreviated simply as concrete.) Inthis paper, iron, lead, concrete with 5.5 % water byweight, heavy concrete with 5.5 % water by weight,graphite, marble, water and paraffin are taken as typicalshielding materials. The depth dependence of dose
equivalents in the inside of these materials is estimatedfor monoenergetic neutron sources of energies less than
400 MeV. Calculations were carried out by using theone-dimensional discrete ordinates code ANISN-JR 2)
and the cross section library DLC-87/HILO 3). The
calculation procedure and tl1e results are given in section
2 and 3, respectively. Details for them are descTibedelsewhere 1) .

The ANISN-JR is a refined version of theANISN4) which was developed to solve the onedimensional Boltzmann transport equation for neutronsand photons in materials by using multigroup crosssection data, and is used to solve penetration problems
under the condition of anisotropic scattering.
The DLC-87 /HILO is one of several neutronphoton multigroup coupled cross section libraries
which have been offered from the Radiation Shielding
Information Center. In this library the cross section
data are given in the structure of the 66 neutron energy
groups from the tl1ermal energy to 400 MeV and the 21
photon energy groups from 10 keV to 14 MeV. The
transport calculations were made for neutrons andsecondray photons with scatterings of an S16 angularquadrature and a P3 Legendre cross section expansion.
The primary neutron in a group of a specified energy
range is normally incident on a shielding material witha slab geometry, and its transitions to elastic andnonelastic channels are evaluated numerically. The
attenuation of the secondary photons produced byneutron nonelastic reactions in materials is also
obtained. Energy-to-dose conversion factors are quoted
from ICRP Publication 21, and then the unit of thedose equivalent was converted from rem to S v using
the relationship 1 Sv =100 rem .
3. Results AND Discussion
3.1. Attenuation of Neutron Dose Equivalent
The depth dependence of the neutron andsecondary photon dose equivalents is shown in Fig. Ifor iron, lead, concrete with 5.5 % water by weight,
heavy concrete with 5.5 % water by weight, graphite,marble, water and paraffin. Here, marble is regarded as
pure calcium carbonate, CaC03. In Fig. 1, solid andbroken lines show the variation of neutron andsecondary photon dose equivalents, respectively.
Numbers on the right of each curve refer to the group
number of source neutron energies, and those underlined
represent the group numbers for the source neutrons
producing the secondary photons. (See Table 1)*deceased-362-ww0)

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Table 1 Adopted Energy Groups
Grouu Nos.
1915
24
35
4652
55

Neutron Ener~y (MeV)
400
375
200
180

10090

40
1035
8.19

1.11

9.07(-1)

1.50(-1) -1.50(-2) --8.65(-2)
7.10{-3)

These materials can be divided into three groups
by the feature of attenuation curves. The first groupwhich is composed of iron, heavy concrete and concrete,is efficient for the shielding of neutrons and secondary
photons in the almost all energy regions, whereas thesecond group consisting of water and paraffin isefficient for shielding low energy neutrons but not so
efficient for shielding secondary photons. The lastgroup consisting of lead, marble and graphite, however,. is not efficient for shielding neutrons and secondary
photons.
Iron is a very effective shielding material forneutrons at energies above 30 MeV and below 100
keV. For the neutrons of around 1 MeV, however, ironis rather inferior to concrete and heavy concrete. Heavy
concrete is also one of the most effective shielding
materials, i.e. it is superior to iron for neutrons below
10 MeV. Although heavy concrete is more efficientthan concrete in the whole energy region, concrete is
widely used because of its lower cost. Water andparaffin are excellent shielding materials for neutrons
below a few tens of MeV. Lead and graphite do notseem to be effective shielding materials in the wholeenergy region. Marble, as a whole, does not havespecial features. It has a similar attenuation pattern tothat of concrete without water as can be supposed
easily.
The dose equivalents are reduced steeply at theend of the slabs, which is due to the lack of backscattering from the outside of shielding materials.
3.2. Attenuation of Dose Equivalent by
Secondary Photons
Source neutrons incident on shielding materials
induce secondary photons by way of a radiative capture
process and de-excitation of reaction products. The
photon dose equivalents caused by monoenergetic
incident source neutrons are shown as a function ofmaterial depth in Fig. I together with neutron dose
equivalents. As the secondary photons are inevitably
accompanied by neutron penetration in shielding
materials, both neutrons andl secondary photons should
be considered inclusively in shielding problems, i.e. asummed dose equivalent will be important practically.
Although photon fluxes are of nearly the same order as
neutron ones for all materials and in whole sourceenergy regions, the photon dose equivalents in shielding

materials are less than the corresponding neutron onesby 2 orders of magnitude. This is due to the fact that thesecondary photon energy is almost independent of theincident source neutron energy and that the flux-to-dose
conversion factor of photons is smaller than that ofneutrons.
The attenuation characteristics of photon doseequivalents should be considered by combining theattenuation of primary and secondary photons.
Secondary photon dose equivalents of source
neutrons above 90 MeV also increase near the vicinityof the surface of shielding materials. However, thesehave little contribution to summed dose equivalents inthis energy region.
4. Conclusions
Quantitative depth dependence of dose equivalentsinside each shielding material is almost same as that atits exit surface of varying thickness. Iron, heavy
concrete and concrete have excellent shielding effects forneutrons of energies more than a few tens of MeV, andwater, paraffin, iron and heavy concrete are efficientshielding materials for neutrons of lower energies. For
neutrons in the thermal energy region, iron and heavy
concrete are especially efficient.
The shielding materials studied can be classified
into the three groups, i.e. heavy elements, hydrogenous
materials and others, according to the difference ofdegradation mechanism of neutron energies.
The attenuation pattern of photon dose
equivaients in shielding materials will be considered toresult in competing processes between the attenuationof photons due to the density and the supply of
secondary photons due to inelastic photon emission
processes of neutrons.
In application to practical shielding, we should
understand characteristics of each dose equivalentattenuation of incident neutrons, primary and secondary
photons in shielding materials.
References1) N. Nakanishi, T.Shikata, S.Fujita, and T.
Kosako, to be published in Nucl. Tech. in Oct.
2) K. Koyama, Y. Taji, K. Minami, T. Tsutsui,
T. Ideta and S. Miyasaka, "ANISN - JR, a One Dimensional Discrete Ordinates Code for Neutron andGamma-Ray Transport Calculations," Report JAERI-M
6954, Japan Atomic Energy Research Institute (1977).
3) "HILO, 66 Neutron, 21 Gamma-Ray Group
Cross Sections for Radiation Transport for Neutron
Energies up to 400 MeV," Radiation Shielding
Information Center, Oak Ridge National Laboratory,
DLC- 87(1981).
4) W. W. Engle, Jr., "A User's Manual for
ANISN, A One-Dimensional
Discrete Ordinates
Transport Code with Anisotropic Scattering," K-1693,
Oak Ridge Gaseous Diffusion Plant (1967).

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