The 12th Symposium on Accelerator Science and Technology, Wako, Japan 1999

Shielding Design of the Cancer Therapy Facility using Compact Proton Synchrotron
Yo MAKITA, Takashi NAKAMURA*, Hiroyuki TSUCHIDATE, and Masaharu YAMAZAKI
Mitsubishi Electric Corporation
8-1-1 Tsukaguchi-honmachi, Amagasaki-shi, Hyogo, 661-8661, Japan*Cyclotron and Radioisotope Center, Tohoku University
Aramaki, Aoba-ku, Sendai-shi, Miyagi, 980-8578, Japan
Abstract
2 Bulk Shielding Design
In these days, due to the increase of proton or heavy ion
cancer therapy facilities, the radiation shielding ofintermediate energy accelerators is becoming an importantproblem.
Radiation shielding design of a cancer therapy facility
using compact proton synchrotron was carried out. For
shielding analysis, the one-dimensional discrete ordinate
code ANISN and the skyshine calculation code SHINE-II
were used. As an emission of neutron by proton beam losses,
measured neutron yield spectra of 256MeV incident proton
at LANL, and the high-energy cross section library DLC119/HIL086 were used for the ANISN calculations.
Results from the calculation satisfy the limits from thedomestic law of the radiation protection.
1 Introduction
The radiation shielding designs of recent accelerator
facilities are becoming important for constructing thefacility because either the particle energy or the beam
intensity becomes higher. The shielding analysis plays animportant role in the conventional facility design andreducing the total facility costs. To reduce the shielding
design costs, the precise estimate of the neutron leakage,
which corresponds to the significant part of the dose
equivalent, is one of the key issues for the shielding designof those accelerators. This is due to the drastic change ofcross sections of the materials around this energy region.
Mitsubishi Electric has been developing a cancer
treatment facility based on the proton synchrotron. Low
cost and compactness are important features for widespreaduse in the hospital environment. The layout of the facility isgiven in figure 1. Table 1 shows the specifications of theaccelerator and the operating condition.
The radiation shielding design of the facility was carriedout. The one-dimensional discrete ordinate code ANISN [1]with the multigroup cross section library DLC-119/HIL086[2] was used for the bulk shielding calculation. NakamuraUwamino's formula [3] was used for the maze calculation.
Also the skyshine calculation code SHINE-II [4] was used
for determining the skyshine.
This paper describes the results of the calculation.

Bulk shielding design of the facility was assessed at thepoints where dose equivalent limits should be determined.
The entire facility was divided into the areas defined by thelaw, such as the controlled area, the area where workers
have full time access, the hospital, and so on. The
calculating points were defined at borders between the areas
as shown in figure 1 as A1- A5 and B1- B7.
The expected radiation sources due to beam losses in theaccelerator, the beam transport, and the delivery systems aregiven in figure 1 as S1 - S6. Radiation occurring from beamlosses in RFQ and losses in synchrotron in processes suchas injection, RF capture, and acceleration were considered
as negligible.
Majority of radiation from the accelerator is caused bylosses occurring in the course of extraction process at thedeflector of synchrotron, given in S1, which was expected
5% of the overall beam current. This is caused bybombardment of protons on the septum made of iron.
Losses occurring in the irradiation process during thetreatments, given at S6, S7, and S8 in figure 1, wereexpected 100% of the beam current. The beam dump is
designed to absorb the full beam. Therefore, losses at thebeam stopper S2, and the beam shutters S3, S4, and S5 wereexpected to 100% of the beam current.
The shielding calculations were carried out assuming theoperation is always at the maximum proton energy of 250
MeV.
Dose equivalent behind concrete shield can be simply
estimated from the equation [5];d_)H= Hn(O) ·J ·exj __rz

~ Aamc(O)A4Table 1 Specifications of the accelerator. cond'.
an d operatmg
thon
Accelerating Particle
ProtonParticle Accelerating Energy
250 (MeV)
Operating current and time per week max. 4388 (nA·min/week)
Beam tuning time per week
max. 335 (nA-min/week)
Fig.l The layout of the cancer therapy facility using
compact proton synchrotron-346-(1)The 12th Symposium on Accelerator Science and Technology, Wako, Japan 1999

Table 2 Source term H0 and A in concrete for neutrons by·256Me:V proton bom bardment on Ironan d carbon

where:
H is the dose equivalent per week at the calculating

cpoint.(~-tSv/week);

Angle

Ho(O) is the source term in the direction 0. (Sv •cm2/nNmin);
J is the proton beam current per week. (nA ·min/week);r is the distance between the source and the calculating
point. (r>d) (em);AconlO) is the attenuation length in concrete in thedirection 0. (em);d is the thickness of the concrete. (em);(de _g.)7.53060120150---------------------------------------------------------------·'·-------- ~ s,'-120 .

·-----~----sv-150-----------.-------- .......

----- ---------------ya_~~~~-~~:t.':l!. _____________________:-----·····-----------------------------------------------------------~----········

1o·• ----·---.. ··-------·-·---· .. ·----------·------------------------·-----.......
--------------------------------·······················---------~---------------------·--·

100

200300400500600

radius [mm]

A

21.2
18.5
8.3
3.2
2.5(em)
52.2
46.5
33.4
20.2
18.5

HoA

35.0
26.9
15.0
12.7
15.9(em)
42.6
37.4
27.0
18.3
17.1

H0 IS g1ven by (Sv·cm 2/nNmin).
Concrete sphere w/ Fe target [7.5 deg.]to• .................................................................... ..

To calculate the transmission curve, the one-dimensionaldiscrete ordinate neutron-photon transport code ANISN wasadopted. The calculations were carried out with onedimensional spherical geometry. The geometry consists of avacuum layer with a radius of 4 m and a concrete layer of 2m thick. Total neutron yield produced by 256 MeV protons
bombarding stopping-length targets of iron and carbon,
found in literature [6] was assumed to be the neutron losses
from beam dumps points and isocenter of gantry. Neutronsand secondary photons are concerned in the calculation.
Photons emitted from beam loss points were neglected. The
neutron-photon multigroup, consisting of 66 neutron energy
groups and 22 photon energy groups, coupled cross section
library DLC-119/HIL086 was used for the calculation.
The calculated energy spectrum is converted to dose
equivalent by multiplying the flux-to-dose equivalent
conversion factor given by the recommendations of ICRP
Publication 51 [7]. Finally, the factor 2, requested by thegovernment, is multiplied to the conversion factor in orderto estimate safely.
Figure 2 shows dose equivalent for each emission angle
calculated by ANISN. The attenuation length of theshielding material is obtained by fitting the transmission
curve of dose equivalent through the shielding with an
exponential function. The Ho and A for each emission angle
are give in table 2.
Fig. 3 and 4 are comparisons of dose equivalents andCONCTERE sphere (ri=400, d=200)~sv 7.5
neutron emission from Fe target ---- =~~=~gFe

Ho....................................... ~sv by ANISN ----------··~sv by l/r'2

10·• k----;-k;,.------,.kn----,~---;;-~;;,.------.rlvc---zJ,
0100200300
400500600r[mm]Fig. 3 Dose equivalents inside of the concrete spherecalculated by ANISN and al/r2 attenuation curve
normalized at r= 1
109__.__source neutron----before wall (AN!SN) .......... .. ............~_..,,.
---+-before wall(alt. with 1/r'2).................................... -----

10 -1 L_L~::tiG'---'-,--.L----'--,---'---L---L_l_j

10"710'3
10'1
Energy [MeV]

101

103

Fig. 4 neutron spectra calculated by ANISN and all~attenuation curve normalized at r=l
neutron spectra inside of the concrete sphere calculated by
ANISN and all~ attenuation curve normalized at r=l,respectively. As shown in Figure 3, the dose equivalentcalculated by ANISN does not attenuate in inverse
proportion to square of the distance r. Also, as seen in thelow energy region of the neutron spectrum calculated by
ANISN (Fig. 4), the component of thermal neutrons does
not attenuate inverse proportionally but keeps almost
constant value. Therefore, the dose equivalent calculated by
ANISN remains a factor 2 - 2.5 larger than the one
attenuates in inverse proportion to square of r.
We have considered this as effect of the back scatteringfrom the concrete wall and applied a correction factor of 3
to the result of eq.(l). Thus, the dose equivalent estimate isgiven by eq. (2).

Fig. 2 Dose equivalent plot of one-dimensional concrete
sphere model for ANISN.

H'=3·H

These results are given in Table 3.

-347-(2)The 12th Symposium on Accelerator Science and Technology, Wako, Japan 1999

. pomtT abl e 3D ose eg mva ent of eac h caIcu Iatmg
pt. H' (!-~Sv/week) I pt. H' (!-~Sv/week) pt. H' (!-ISv/week)
AlA2

A3
A4AS

69.6
0.6
30.9
4.0e-3232.6

B6Bl0.2
B796.2
B2Cl
B3
530.7
C2289.5
B4
11.0
B5* given in ~-tSv/3months

5 Summary

274.5
0.2
6.5 *
206.7 *

3 Attenuation of Radiation from Access Mazes
The neutron and secondary photon dose equivalent at
the exits of the maze were calculated. There are a number offormulas for estimating the attenuation of the radiation inaccess maze. Here, Nakamura-Uwamino 's formula, derived
from simple equations, was chosen.
The dose equivalent at the maze entrance was estimatedfirst and then the attenuation in each leg of the maze wasestimated by the formula. Finally at the maze exit theattenuation of the slow neutron component by a door madeof 10-cm-thick polyethylene with 2-wt. % boron was taken
into account.
Contribution of the secondary photon can be assumed to
be 1/10 of the neutron dose equivalent according to theDOT3.5 calculation [3]. The factor, 2, requested by thegovernment was applied as well as in the case of bulk
shielding. Table 4 shows the results of the attenuation in themaze.
Table 4 Dose equivalent at the entrance
and the exit of the mazes
Calc.
Hcxit
Hcntrancc(!-~Sv/week)(!-ISv/week)
Point
12.5
8.3xl04B4
21.2
8.0x104B5
0.4
4.lxl04B6
0.3
2.4x104
B7

Overall dose equivalents for all the calculating points
are given by table 6. Contributions by the radiation via themaze and the skyshine are around 10-50 % and lower than
10 %of the directed transmission, respectively.
The calculating points, denoted by Al - AS, Bl - B7,
and Cl - C2 are located at the border of the controlled area,at the area of the area where workers have full time access,and the border of the facility, respectively. Dose equivalent
limits for these borders by the law are 300 f!Sv /week,
lmSv /week, and 1.3mSv/3months, respectively.
Since all results satisfy the limit, our analysis supportsthat the shielding design of the facility is appropriate.
. Ient of ca Icu Iat1on. pomts.
Tabl e 60vera lldose eqmva
pt. H' (!-~Sv/week) pt. H' (!-~Sv/week) I pt. H' (!-~Sv/week)

AlA2

A3
A4AS

69.6
0.6
30.9
4.0e-3
232.6

0.2
B6Bl
B7
96.2
B2Cl
530.7
B3C2
B4
310.7
23.5
B5
* given in ~-tSv/3months

274.9
0.5
6.6 *
207.0 *

Acknowledgement
The authors deeply appreciate Messrs. T. Kurosawa, M.
Sasaki, H. Iwase, and T. Nunomiya for their continuous
support and assistance in the computer code calculations.

References

4 Skyshine
The neutron and secondary photon dose equivalentcaused by skyshine effect was calculated by the skyshine
calculation code SHINE-II. First, in order to estimate theleakage neutron and secondary photon from the roof of thefacility, the angular flux of radiation from the concrete
sphere of 2 m thick was calculated by ANISN.
Second, the neutron and secondary photon dose
equivalent by skyshine was calculated by SHINE-II. Two
calculating points were defined;• Cl is the point at the border of the facility. The distancefrom the nearest beam loss point is 75 m.
• C2 is the point at the border of the hospital. The distancefrom the nearest beam loss point is 45 m.
Table 5 shows the results of the calculation.

[1] RSICC COMPUTER CODE COLLECTION ANISNORNL, CCC-254, Oak Ridge National Laboratory, OakRidge, Tennessee.
[2] RSICC DATA LIBRARY COLLECTION HIL086,
DLC-119 HIL086, Oak Ridge National Laboratory,
Oak Ridge, Tennessee.
[3] Y. Uwamino, T. Nakamura, "Shielding Design andDetection of Neutrons from Medical and Industrial
Electron Accelerators", Radioisotopes, Vol. 35, 2(1986)[4] K. Hayashi eta!. "DUCT-II and SHINE-II: Simple
Design Code for Duct-Streaming and Skyshine",
JAERI-M 91-013 (1991)[5] H. W. Patterson and R. H. Thomas, "Accelerator
Physics", Academic Press, 1973
[6] M.Meire, eta!., "Neutron Yields from Stopping- andNear-Stopping-Length Targets for 256-MeV Protons",
Nucl. Sci. and Eng., Vol. 104, 1990
[7] International Commission on Radiation Protection(ICRP) Publication 51. bJY SHINE-II
Table 5 Resu ts of the sk sh'me ca Icu Iation
Calculating pt. H,kv,hin" (!-1Sv/3months)
0.07
Cl
0.31
C2-348-