Fujita Technical Research
ReportNo.53 2017

Discussion on radiation shielding materials for slit space
between accelerator buildings

Ken-ichi Kimura
Abstract
For the new construction of an accelerator building adjacent to an existing facility, it is necessary togive careful consideration with reference to two regulations; the building and radiation codes. For theslit space between the accelerator buildings in case of using of the two facilities as one, two kinds ofinvestigation were conducted on the development of shielding flexible materials and on theevaluation of neutron dose attenuation through elemental data on several types of concrete in theradiation shielding calculations. A new flexible shielding material was tested for several aspects, andthe radiation shielding calculation for application of the material was carried out to confirm theperformance. As a result, a urethane-based material was selected and applied as designed. Anothercalculation of neutron dose attenuation for the slit space between the buildings was conducted forthirteen different types of concrete for shielding. This investigation emphasized that proper concrete
data should be used for the radiation shielding calculation, while the hydrogen content and thedensity of the concrete were dominant for neutron attenuation.

Keywords: accelerator radiation shield, concrete for shield,
Monte Carlo, dose attenuation−49−㻙㻌㻝㻌㻙㻌

Fujita Technical Research Report No.53§1. Introduction
aspect. Flexibility was necessary to absorb theConcrete is one of the main materials used for

displacement of the buildings in case of an

constructing buildings, infrastructure, and facilities,

earthquake, in order to avoid damage caused by thewhich are a major part of human public activities.

two buildings crashing together when their shaking

These materials are also typically used in reactor

cycles differed. For the radiation shielding aspect,

and accelerator facilities for shielding.

neutron shielding performance was assumed as the

The construction of an accelerator facility requires atmain issue for the material, while the cranked

least two aspects of regulation; building and radiation.

shaping concrete wall would be good for shielding

The building code provides the shape and soundness asagainst gamma ray. The narrow-slit space for thethe container of the building, including structural andmaterial fitted between buildings was outside of theaseismic excellence. On the other hand, radiationbuilding

protection is, of course, very important for the facility. In

performance should be maintained throughout aparticular, the above two aspects should be consideredcertain period despite several problems posed by thevery carefully in the case of using the two facilities as one.

weather, reasonable durability was necessary.

There are two accelerators located at Kyushu

Another important factor in selecting the material

University; one Fixed-Field Alternating Gradient

was construction-related performance, such as

accelerator (FFAG) and one tandem accelerator. The

workability, ease of manufacture, and reasonablebuilding for the FFAG was constructed in 2008 [1],

cost.

and

mostlyshaded.

Becausetheand that for the tandem accelerator in 2014. Since
the newer building was constructed as a separate

2.2 Durability test for the material
Fifteenstructure away from the FFAG building with a joint

candidatematerial

samples(siliconradiation controlled area, special consideration was

rubber, water sealant, special epoxy, urethane

required to construct the newer building withpolymer, polyethylene and mortal slate) were

sufficient space between the two buildings.

gathered at the location so as to be reasonably
available once supply was confirmed within the§2. Flexible radiation shielding material for

construction schedule. Upon receiving the above
samples, we started to test certain aspects of theaccelerator buildings

durability
A new type of flexible shielding material was

performance,

assumingasettledcondition for the actual construction.
The outlines for the results [2] of the tests were asapplied to the narrow space of 8 cm in widthfollows;

between the two accelerator buildings (the existingbuilding and the newly constructed one) for neutron1) A comparison of the before and after tests

shield. The main requirements of this shielding

showed no shrinking and no mass change formaterial were flexibility, shielding performance,

any samples.
2) Three samples had changed in size (a reduction

and durability.

of more than 15 % due to the curing) after anaccelerated

2.1 Requirements of the material
Four kinds of performance for the material fittedasbuilding

structural

aspect,

test

and

heat

resistance test.

to the slit space between buildings were required,such

weather-proof3) Besides the above, a few samples changed inradiation

colour after tests, but the changes did notshielding aspect, material aspect and construction

affect the performance.

−50−
2Discussion on radiation shielding materials for slit space between accelerator buildings

2.96 [3]. Figure 1 shows the calculation geometry for this25cm

investigation. Because the border space between the twobuildings was a very complicated shape and around 10 m
in height, we used R-Z cylindrical geometry with the

Estimation area
for calculation

vertical direction of the building for the radial (R)Vertical direction (cm)direction and with the horizontal direction for the Z axis,
to achieve a reasonably small statistical error. The green

Existing wall

portion on the left indicates the existing building with the
FFAG (typically a 150 MeV proton beam) and the right

New wall8cm

orange portion indicates the new building. The model forthe calculation was focused on the vertical attenuation of

Flexible shielding
material

the dose generated by the proton beam to the virtual ironset around the border of the two buildings.

Proton beam Collision Point 2

The proton beam was transported from the existing

Collision Point 1

building through the beam hole in the existing wall,
which was on the centre line of the calculation geometry

Horizontal (beam) direction (cm)(vertical direction = 0 cm in Figure 1) with 150 MeV. The

Figure 1 Calculation model for the two buildings.

virtual iron target was set on the line at Z = 200 cm,
which was assumed right after the existing wall and right
below the narrow slit fitted by the flexible shielding
material (Collision Point 1, referred a CP1 in figures). In

Further investigation was conducted for other

another case, the calculation was conducted with the ironconditions such as flexibilty, stability, workablity,

target set at Z = 327.5 cm on the centre line under theshield performance to select the material to apply toborder space of the two buildings (Collision Point 2,

the space.

referred to a CP2 in the figures). Several estimations were

We had developed normal urethane into a new

conducted for the outdoor border space, for which there is

type of flexible shielding material in workablity anda 25-cm width to the celling of the first floor of the new

manufacturability.

building, which is 8 m high.

The shape of the material was set as a block (230 mm x

Table 1 shows the material elemental data for the

350 mm x 80 mm) from the point of view of construction.

calculation. For the new building, chemical analyses wereMore than 1000 blocks were manufactured at the

conducted for 18 samples from constructed concrete,construction site, and applied to spaces of 80 mm and 250

which are recorded as FH02 [4] in the table. Another piecemm space.

of concrete data in the table, shown as AV03 [4] assumed
the typical concrete shielding data discussed in the

2.3 Calculation for the adaptation of the material

Atomic Energy Society of Japan. The elemental data for

After the material durability tests, estimation of thethe flexible shielding material were calculated

performance of the urethane-based flexible shielding

according to the chemical equation listed in thematerial was conducted using Monte Carlo code, PHITS

material safety data sheet.

Table 1 Elemental data of materials for the calculation
Material

H

C

O

Na

Mg

ElemetElement
Al
Si

Av03
FH02
SFM

0.6
0.5
9.8

0.6 48.6
5.1 48.8
72.5 16.0

1.7
0.7

1.3
0.5

5.7
2.7

26.2
16.0
0.1

S

K

Ca

Ti

Fe

0.1
0.0

1.9
1.1

10.0
23.7

0.3
0.1

3.0
0.6

㻙㻌㻟㻌㻙㻌
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N

1.2

w%

DensityClg/cm-3

0.4

2.26
2.27
1.00

Fujita Technical Research Report No.53

the slit to 50 cm upon the surface of the concrete.
Figure 2 indicates the calculated dose attenuation ofphoton and neutron for the border space between thebuildings, from proton-iron collision at CP1 and CP2,
respectively. The dose for neutron from CP1 (indicatedwith a solid black line) is the highest among the four
curves, and the orders describe the doses for neutron from
CP2, for photon from CP1, and photon from CP2. Since
the doses for the neutron are 100 times higher than thosefor photon for all calculated heights, further estimation
was conducted for neutron attenuation for CP1 only.
The case with the materials installed in the 8-cm slit
and the case without the materials were compared. Four
Figure 2. Calculated dose attenuation in the borderarea

different calculations with/without the installation of thematerial were conducted. The indexes in the figure for the
calculation conditions refer to the installation of theflexible shielding material as “wM” in the last two
characters and to the case not using the material (set air
in this case) as “wO” in following figures.
In the calculation, the paths of the neutrontransportation were also estimated using the counterfeature in PHITS code. The curves with a “ThS” index in
the first three characters show the calculated dose for theneutron which were transported through the 8-cm slit,
while the other curves with an “OuS” index show thosefor the neutron which were transported out of the slit forwhole the life in the calculation. The curve with a“ThS_wM” index refers to the calculation in the case with
the installation of the flexible shielding material for theneutron transported through the slit, for example.
Figures 3 shows a comparison of neutron dose

Figure 3. Comparison of neutron dose attenuation for the
installation of the flexible shielding material originatedproton-iron reaction at CP1.

attenuation between cases with and without the flexible
shielding material installed, with neutron occurring due
to proton-iron collision at CP1. This figure indicates thatthe neutron dose decreased to less than half, and that the

The area for the calculation was the 25-cm border spaceneutron dose through the slit significantly decreased to

between the two buildings which is outlined in red inless than 1/20 with the material installed at a high of 390

Figure 1. T-Track tallies were set in the area for the

cm from the beam line with a proton-iron reaction

neutron and photon spectrum and the dose with dose

originating at CP1. As the quantity of the neutron doseconversion factors [5].

with “All_wO” was less than that with “All_wM”, theThe new urethane-based flexible shielding material

installation was clearly effective in reducing the radiation

was installed at the slit area between the two buildings,

dose in this case.

which was 8 cm in width and 150 cm in length from thebottom of the concrete slab for the new building through㻙㻌㻠㻌㻙㻌
−52−

Discussion on radiation shielding materials for slit space between accelerator buildings

F02SD was chemically analyzed data with existing§3. Comparison of the dose attenuation for theconcrete samples in this building. STD17 was defined
through a discussion on the standardization of concrete

slit space on different concreteelements for radiation shield in Japan.
In shielding design for new buildings, existing

These elemental values for each density were used as

published concrete data are used for calculation.

concrete in the calculation for both the existing and new

These data, however, might not have been chosen

walls in Figure 1. The dose attenuation curves for thereasonably.

the

25-cm slit (referred to as “estimation area” in the figures)

differences between these concrete data is estimatedwere calculated using the proton reaction at CP1. The

in this session.

flexible shielding material described in Section 2 was not

Therefore,

theinfluenceonapplied to the narrow, 8-cm slit in this calculation.

3.1 Concrete data for the calculation
The calculation geometry and the Monte Carlo code for3.2 Result of the calculation

the investigation were the same as in Session 2 (Figure 1

Figure 4 shows the calculated neutron doseand PHITS2.96). Elemental data were gathered on

attenuation curves for the 25-cm slit space with thethirteen types of concrete to be used in calculating for theelemental data for the different types of concretedose attenuation.

listed in Table 2, as well as originating proton iron

Table 2 summarizes the elemental data for thirteen

reaction at CP1 in Figure 1. The thirteen curves

types of concrete, namely ANL5800 type 1 (referring

were divided into three groups by attenuation rates

ANL501), ANL5800 type 2a (ANL502a) [6], ANL6443 type(“High”, “Medium” and “Low”). The “High” group

2b (ANL602b) [7], composition of CONCRTE PORTLAND

contains ANL502a, NIST98, NBS03, and F02SD,

in NIST (NIST98) [8], NBS type03 (NBS03), NBS type 04

while the “Low” group contains ANL501, HFDd and(NBS04), Los Alamos National Laboratory concrete

STD17, and the “Medium” group contains the others.

(LANL), Oak Ridge National Laboratory concrete

These categorizes also come from the weight(ORNL) Hanford Dry concrete (HFDd) [9], JAERI-M-6928

percentage of hydrogen given in Table 2. The figures

type 1 (JAERI1), JAERI-M-6928 type 2 (JAERI2) [10],and the table indicate that the neutron dose

F02SD, and Standard concrete 2017 (STD17) [11]. Data forattenuation for the slit space is influenced by thethe first eleven were provided by national organizationshydrogen content of the surrounding concrete.

in the USA, and the next two by JAERI. The data for

Figure 5 shows the other calculations of neutron

Table 2. Elemental data of materials for the calculationName
ANL501
ANL502a
ANL602b
NIST 98
NBS03
NBS04
LANL
ORNL
HFDd
JAERI1JAERI2
F02SD
STD17

H

C

w%

Element

O

Na

Mg

Al

Si

P

S

K

Ti

Mn

Fe

0.21 5.58 49.3
1.00 0.10 53.0
0.52 0.10 51.3
1.00 0.10 52.9
0.85 5.01 47.3
0.56 0.00 49.8
0.45 0.00 51.3
0.62 17.52 41.0
0.40 0.00 48.2
0.42 0.00 50.7
0.48 0.27 49.6
0.88 5.10 50.1

0.00
1.60
1.67
1.60
0.00
1.71
1.53
0.03
0.22
0.00
1.10
0.63

0.21
0.22
0.23
0.20
2.42
0.26
0.00
3.26
1.41
0.12
0.95
0.51

0.51
3.39
3.55
3.39
3.61
4.57
3.56
1.08
6.94
0.45
5.61
2.63

18.8
33.7
35.2
33.7
14.5
31.5
36.0
3.4
27.8
38.6
29.9
14.4

0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.03

0.08
0.00
0.00
0.00
0.30
0.13
0.00
0.00
0.00
0.07
0.16
0.00

0.00 24.9 0.00
1.30 4.34 0.00
1.36 4.55 0.00
1.30 4.40 0.00
0.17 24.69 0.00
1.92 8.29 0.00
0.00 5.79 0.00
0.11 32.1 0.00
1.30 8.02 0.00
0.00 6.87 0.00
0.80 8.25 0.17
1.07 24.0 0.08

0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.00
0.25
0.02

0.31
1.39
1.46
1.40
1.10
1.24
1.38
0.78
5.75
2.74
2.55
0.57

0.34

1.64

1.12

5.42

24.5

0.04

0.13

1.86

0.04

2.91

1.23

47.4

㻙㻌㻡㻌㻙㻌
−53−

Ca

13.1

0.26

Density
g/cm

-3

2.33
2.30
2.20
2.30
2.35
2.35
2.25
2.30
2.18
2.30
2.43
2.32
2.10

Fujita Technical Research Report No.53

Figure 4. Comparison of neutron dose attenuation for the slit space on different types of concrete with the proton-ironreaction originating at CP1.
dose attenuation for the proton iron reactionhydrogen content and density as the key indexes oforiginating in the slit at CP2 with those originating

the attenuation. It is also important to choose

at CP1 for “High” group and “Low” group. The doseproper data from among the variety of publishedattenuation for CP2 are almost all smaller thanconcrete elemental data in order to estimate thethose for CP1, and it is indicated that the “Low”

dose attenuation accurately.

group (ANL501a and STD17) would have a different
attenuation due to its density because of theattenuation paths.
Therefore, the investigation of neutron dose
attenuation for the slit space with the elemental
data of different concrete was conducted with the㻙㻌㻢㻌㻙㻌
−54−

Discussion on radiation shielding materials for slit space between accelerator buildings

Figure 5. Comparison of neutron dose attenuation for the slit space with different types of concrete at CP1 and CP2.

§4. Conclusion

Acknowledgements

In the case of new construction as a separate,

The author wishes to acknowledge Prof. Nobuhiro

adjacent

Shigyo and Prof. Nobuo Ikeda of Kyushu University.

building,

two

investigations

wereconducted. For the narrow-slit space (8 cm in width
between buildings), a new type of flexible shielding
material was developed with consideration for the
specifications of the required performance for thefour

aspects(structural,radiation

shielding,material, and construction). After several durability
tests

and

radiation

shieldingcalculations,

Referencesa[1] Y. Yonemura, A. Takagi, M. Yoshii, Y. Mori, M. Aiba,

urethane based material was selected and applied

K. Okabe, and N. Ikeda, “Development of RF

as designed.

acceleration system for 150MeV FFAG accelerator”,

Elemental composite data on thirteen types of

NIM, A 576, 294-300 (2007)

concrete for radiation shield calculation were[2] K. Kimura, N. Shigyo, S. Tahkahashi, H. Hirasawa, S.

gathered to compare the dose attenuation for the slit

Utsumi, N. Ikeda, and K. Ishibashi, “Development of

space between the two buildings. The calculation

radiation shielding flexible material for accelerator

results for the neutron dose indicate that thebuilding”, ICRS-13 (2016), to be published.

hydrogen content and density of concrete were

[3] T. Sato, K. Niita, N. Matsuda, S. Hashimoto, Y.

important for the attenuation. It was also important

Iwamoto, S. Noda, T. Ogawa, H. Iwase, H. Nakashima,

to choose reasonable concrete data to properly

T. Fukahori, K. Okumura, T. Kai, S. Chiba, T. Furuta

estimate the radiation shield because of the wideand L. Sihver, Particle and Heavy Ion Transport Codevariation in hydrogen contents and density in data

System PHITS, Version 2.52, J. Nucl. Sci. Technol. 50:9,

published in the past.

913-923 (2013).
[4] K. Kimura, T. Ogata, M. Nakata, K. Okuno, Y.
Sakamoto, T. Hirouchi, M. Taniguchi, K. Tanaka, K.
㻙㻌㻣㻌㻙㻌
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Fujita Technical Research Report No.53

Oishi, T. Tsukiyama, S. Ishikawa, H. Sakamoto, N.

[9] RJ McConn, et. al, Compendium of material

Kawano, H. Kawano, M. Yoshida, T. Amano, K. Kosako,

composition data for radiation transport modelling,

Y. Hirao, N. Shigyo, N. Ikeda, and K. Ishibashi,

PNNL-15870, March 2011.

“Standardization of concrete for radiation shielding –[10] K. Koyama et al., Multi-group cross section sets for

Discussion on the concept of standard concrete-”,shield materials”, JAERI-M 6928, 1977.

SATIF-13 (2016), to be published.

[11] K. Kimura, et al., Discussion on the standardization[5] Atomic Energy Society of Japan, Radiation Doseof concrete composite for radiation shielding design I

Conversion-concept of standard concrete for radiation shielding -,Coefficientsfor

Radiation

ShieldingCalculations: 2010, AESJ-SC-R002, Tokyo, Japan,

Proceeding of ICAPP 2017, April 2017.

(2010).
Short comment
Accelerator facilities are one of the targets
of application for the Low-Activation
technology. In addition, investigation ofelemental composite for concrete is also
important in estimating dose attenuation
properly.

[6] Argonne National Laboratory, Reactor Physics
Constants ANL-5800, July 1963.
[7] Argonne National Laboratory, A Summary ofshielding constants for concrete ANL-6443, November
1961.
[8] National Institute of Standard and Technology web

Ken-ichiKimurahttps://physics.nist.gov/cgi-bin/Star/compos.pl?
matno=144.

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