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Neutron Streaming Evaluation for the DREAM FusionPower ReactorYasushi Sekia, Seiji Morib, Satoshi Nishioa, Shuzo Uedaa & Ryoichi Kuriharaa

a Japan Atomic Energy Research Institute Naka-machi, Naka-gun, Ibaraki-ken 311-0193b Kawasaki Heavy Industries LTD. Minamisuna, Koto-ku, Tokyo 136-8588Published online: 27 Aug 2014.

To cite this article: Yasushi Seki, Seiji Mori, Satoshi Nishio, Shuzo Ueda & Ryoichi Kurihara (2000) Neutron StreamingEvaluation for the DREAM Fusion Power Reactor, Journal of Nuclear Science and Technology, 37:sup1, 268-275, DOI:10.1080/00223131.2000.10874888

To link to this article: http://dx.doi.org/10.1080/00223131.2000.10874888

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Journal of NUCLEAR SCIENCE and TECHNOLOGY, Supplement 1, p. 268-275 (March 2000)

Neutron Streaming Evaluation for the DREAM Fusion Power Reactor

Yasushi SEKI*,t, Seiji MORl**, Satoshi NISHIO* , Shuzo UEDA* and Ryoichi KURlHARA*

* Japan Atomic Energy Research Institute ** Kawasaki Heavy Industries LTD.

Aiming at high degree of safety and benign environmental effect, we have proposed a tokamak fusion reactor concept called DREAM, which. stands for DRastically EAsy Maintenance Reactor. The blanket structure of the reactor is made from very low activation SiC/SiC composites and cooled by non-reactive helium gas. High net thermal efficiency of about 50% is realized by 900 C helium gas and high plant availability is possible with simple maintenance scheme. In the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Neutron streaming through the cooling pipes could cause hot spots in the superconducting magnets adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement. Neutron streaming could also activate components such as gas turbine further away from the fusion plasma. The effect of neutron streaming through the helium cooling pipes was evaluated for the two types of cooling pipe extraction scheme. The result of a preliminary calculation indicates the gas turbine activation prohibits personnel access in the case of inboard pipe extraction while with additional shielding measures, limited contact maintenance is possible in the case of outboard extraction.

KEYWORDS: Neutron streaming, Dose, Radiation shielding, DREAM reactor, SiC/SiC composite, helium gas cooling, tokamak,/usion reactor, superconducting magnet

I. Introduction

There are a number of shielding analyses for fusion power reactors(1,2). The DREAM fusion reactor concept(3) with excellent environmental and safety characteristics using very low activation SiC/SiC composite (hereafter shortened as SiC) structures and high temperature helium gas cooling for high thermal conversion efficiency has been proposed. In the design of the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Furthermore, a low activation material with a low atomic number is generally not a good shielding material so that more space is needed for shielding using only the low activation materials. Neutron streaming through the cooling pipes could cause hot spots in the TFCs adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement.

Neutron streaming through the pipes could also activate components such as gas turbine further away from the fusion plasma to increase the dose rate level during the maintenance operations. The effect of neutron streaming on the TFC and gas turbine was evaluated for the two types of cooling pipe extraction scheme, i.e. extraction in the torus inboard direction or in the outboard direction. In the case of the extraction in the inboard direction, there is little space for additional shield or bending the cooling pipes to reduce the streaming. On the

* Naka-machi, Naka-gun, Ibaraki-ken 311-0193 ** Minamisuna, Koto-ku, Tokyo 136-8588

Corresponding author, Tel. +81-29-270-7502 Fax. +81-29-270-7519, E-mail:sekiy@naka.jaeri.go.jp

other hand, there are spaces for such measures in the outboard direction.

In the next section, a brief description of the DREAM reactor is given. In the third section, the calculational models used for the Monte Carlo evaluation of fast neutron fluxes are introduced. In the fourth section, the calculational methods are described. The neutron streaming through the helium cooling pipes has been evaluated using the MCNP-4B code(4) and JENDL-3.2(5) based nuclear cross section library. The derivation of the conversion factors for the dose rate during the maintenace from the fast neutron flux is also given. The calculated results are presented in the fifth section followed by the conclusions in the end.

II. DREAM Reactor

The DREAM reactor concept(3) with excellent environmental and safety characteristics using very low activation SiC structures and high temperature helium gas cooling for high thermal conversion efficiency is shown in Fig. 1. The figure shows the inboard extraction of cooling pipes. The reactor also aims at achieving a high plant availability through a simplified blanket replacement scheme of removing an entire sector of I /16 torus by a single horizontal motion between the toroidal field coils. The reactor is large with 16-m plasma major radius and 2-m minor radius. Its magnetic field is high with the maximum toroidal field strength at 20 T. A large fusion power of 5.5 GW and a thermal power of 6.4 GW combined with the high net thermal efficiency of 45% lead to a large net electric power of 2.9 GW. This could lead to a relatively low cost of electricity depending on the

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Crane

GasTurbine

Cryostat

NBI Duct

Toroid al Coil

Divertor

ooling Pipe Assemblage

(Cen:ral Pipe) Vacuum Pum

Fig. 1 Bird's eye view of the DREAM Reactor with the inboard extraction of cooling pipes

--..c: 10 -> en '-'"

S as II: m fI) o C ti 10-as -c::: o 010-

2FPYat Neutron Wall Load 5MW/m2

FW Surface

impurities

10-2 100 102 104

Cooling time after shutdown (y)

Fig. 2 Contact dose rate of the first wall of various structural materials as a function of cooling time after shutdown, irradiated at 5MW 1m2 for 2 full power years

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cost of SiC and 20 T superconductor. In the case of SiC, the environmental and safety aspect of

the reactor improves greatly by the rapid decrease of radiation dose level, decay heat and radioactivity after reactor shutdown .

Figure 2 shows the rapid decrease of contact dose rate of irradiated SiC first wall in comparison with the other candidate materials for fusion reactor first wall. This could lead to lower dose level during maintenance operation and lower external dose in case of activated SiC release. Reduction of gamma­ray dose during the maintenance of a fusion reactor eases the requirement of radiation hardening of the remote maintenance components. Low contact dose could reduce the worker dose. Furthermore increased personnel access could result in higher reliability and availability of the fusion plant.

The decay heat of the first wall decreases to less than 1 W/cc and 10-3 W/cc, 1 minute and 1 day after shutdown, respectively(6), so that no emergency cooling system nor active cooling during maintenance operation are needed. As for the long-term decay heat relevant to radioactive waste management, low decay heat will enable the disposal without an interim storage the cost of which is non-trivial.

It has been evaluated(7) that all the radioactive SiC structure can be disposed of by the shallow land disposal if only the nitrogen content of the existing SiC sample(8) can be reduced to 5xlO-4 Wt.% or 1120 of the present amount.

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Fig. 3 Horizontal cross section of DREAM for the inboard extraction ofthe cooling pipe

..... 8

Turbine room ® t ®

I

Con eret e

I

@ I

..... Ul 0

Ul 0 UJ ..... 00 Z ~

~x UJ

00 ~

Fig. 4 Streaming calculation model for the inboard extraction ofthe cooling pipe (not to scale, unit in cm)

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Toroidal field coil

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As described above, the SiC offer excellent environmental and safety characteristics. However, it should be noted that to achieve such excellent characteristics, some further developments are needed.

The in-vessel components must be produced with the minimum amount of high activation materials. The very low contact dose cannot be accomplished if significant amount of high activation material is present. This requirement is difficult to satisfy considering for example, the divertor surface material with low sputtering erosion or a material with good neutron shielding capability, which can endure high temperature condition. Since low activation material are generally not a good neutron shielding material, the thickness becomes more than 2 m when only the low activation materials such as SiC, Li20 and Be are used. Furthermore, neutron streaming is also a problem for helium cooling with large diameter coolant header tubes. At present, the use of SiC has resulted in all waste qualifying for the shallow land disposal but the volume of the waste has increased. It may be necessary to introduce some amount of medium activation, good shielding material at a certain depth of the shield to compromise effective shielding capability and low activation.

III. Helium Cooling Pipe Extraction Models

Fig. 5 DREAM reactor vertical cross section for the outboard The horizontal cross section of the DREAM reactor core extraction of the cooling pipe at the mid-plane with cooling pipes extracted in the inboard

direction is shown in Fig. 3. In the figure, the additional shield to protect the TFCs from the neutron streaming through the cooling pipes can be seen. The space around the cooling pipes may be filled with shielding materials but for the efficient

Tu rbine room

Concret e

36

.~--------­

SiC

z

~x Fig. 6 Streaming calculation model for the outboard extraction of the cooling pipe (not to scale, unit in cm)

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Table 1 Nuclide densities of materials assembling and disassembling operations, it is desirable to leave the space open. The sixteen 80 cm diameter pipes are connected to one large pipe with the diameter of 600 cm in the torus center and brought up to the gas turbines above the reactor core as can be seen in Fig. l. The X-Z cross section of the cooling pipe model for the inboard extraction is shown schematically (i.e. not to scale) in Fig. 4. The structural material for the blanket, shield and pipes are SiC which are in gray color. The main cooling pipe diameter is 80 cm and its thickness is 4 cm. It extends in the radial direction from the outer surface of the blanket penetrating through the 150-cm thick shield to the torus center where it is connected perpendicularly to the central pipe with the diameter of 600 cm. The central pipe then goes up through I-m thick biological shield made of concrete to the turbine room. Fast neutron flux detectors are placed at the surface of pipe inlet (F2), inside the pipes (F4I, F42) and pipe exit (F43) and around the pipe (F44 and F45). The boundary conditions for the neutron transport calculation is reflective condition at the left in the X-coordinate direction on the plasma center side, vacuum condition at the right and vacuum condition at the top and bottom side in the Z-coordinate direction. In the page depth (Y-coordinate) direction, the thickness is assumed to be 10m and reflective conditions to apply on both directions. As the result, the plasma region becomes a rectangular cube with 10m in the depth direction.

[Unit in 10'4 atoms/em']

The vertical cross section of the DREAM reactor concept with the cooling pipes extracted in the outboard direction is shown in Fig. 5. In this case, the pipe is bent several times before reaching the ring header at the top. From the ring header, pipes pass through the I-m thick biological shield concrete to the turbine room. The vertical cross section of the cooling pipe model for the outboard extraction is shown in Fig. 6. In this model, the 80-cm diameter cooling pipe is bent perpendicularly for three times after passing between the TFCs located behind the 150-cm shield. After reaching the ring header at the top, a 150-cm diameter pipe goes up from 5-m offset location in the depth direction of the ring header (which is a straight pipe in this model) through the I-m thick biological shield concrete to the turbine room. Fast neutron flux detectors are placed inside and around the cooling pipe. The same set of boundary conditions as the ones used in Fig. 4 is used.

The nuclide compositions of the materials shown in Fig. 4 and Fig. 6 are shown in Table 1.

IV. Calculation Methods

Three dimensional Monte Carlo neutron transport calculation code MCNP-4B(4) is used to evaluate the fast neutron flux at the detecter regions with energy greater than 0.1 MeV. Weighted window method is used for variance reduction. The nuclear data are obtained from the evaluated nuclear data library JENDL-3.2(5).

As for the activation dose rate evaluation in the turbine room, the flux to dose conversion factors obtained from a

Material

Plasma

Beryllium

Natural

Lithium

85%TD

Concrete

SS316LN

Element Number

Density

D 1.00E-ll

Be 1.24E-OI

6Li S.12E-03

7Li 6.39E-02

a 3.45E-02

H 6.36E-03

C 3.00E-04

a 4.12E-02

Na 6.94E-04

Si 1.68E-02

K 2.73E-04

Ca 3.30E-03

Mn 6.08E-05

Fe 6.08E-04

lOB 1.75E-06

lIB 7.09E-06

C 8.95E-05

Na 2.39E-04

a S.97E-06

Al 8.85E-OS

Si 8.50E-04

31p 3.86E-05

S 1.12E-05

K 6.1IE-07

Ti 1.50E-04

V 3.75E-06

Cr 1.61E-02

Mn 1.57E-03

Fe 5.56E-02

Co 4.05E-OS

Ni 9.97E-03

Cu 7.52E-05

Zr 1.05E-06

Nb 2.57E-06

Mo 1.25E-03

Sn 8.05E-07

Ta 2.64E-07

Material Element

SiC for Si

transport C

calculation

Blanket* SiC

Be

Li2a SiC C

composition N

for a activation Na

calculation Al

[Wt.%] Si

CI

K

Ti

V

Cr

Mn

Fe

Ni

Cu

Zn

Zr

Mo

Sc

Ba

Hf

W

Au

Pt

one dimensional activation calculation ofthe DREAM reactor *Blanket region consists of34.8% Li,O, 22.2% Be, 3.5% SiC.

as shown in Fig. 7, was used. The induced activation calculation code system THIDA-2(9) was used for the

Number

Density

4.67E-02

4.67E-02

3.S%

22.2%

34.8%

2.99 X 10+1

1.00 X 10-2

3.80 x 10-1

1.70 X 10-4

S.60 x 10-5

6.97 X 10+1

2.00 x 10-2

8.20 x 10-3

1.90 x 10-4

3.00 x 10-5

2.10 X 10-4

8.00 X 10-5

4.S0 X 10-4

8.00 x 10-5

2.10 x 10-4

3.70 X 10-5

1.00 x 10-4

2.20 x 10-5

2.10xl0-7

4.00 X 10-5

1.60 x 10-5

1.00 x 10-5

1.00 x 10-8

2.00 x 10-4

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calculation. The neutron flux in the one dimensional model of the DREAM reactor was calculated using the ANISN code(lO) and the group cross section set FUSION-40(l1). The nuclide composition of each regions for the transport calculation using the one-dimensional model of Fig. 7 is shown in Table 1. The SiC composition based on Ref.(8) for the activation calculation is also shown in Table 1. The activation due to continuous irradiation of 20 years at the neutron wall load of 3 MW 1m2

was calculated. The dose rate was calculated for the cooling times of one day and one month after shutdown. The calculated dose rate values at each location were divided by the corresponding values of fast neutron flux and the ratio (flux to dose conversion factor) is plotted in Fig. 8.

V. Results

1. TFC Streaming Effect The fast (E>O.1 Me V) neutron flux along the outside surface

of the cooling pipe is shown in Fig. 9. The fast neutron flux near the TFC at the distance around 200 cm from the end of the blanket is 1 ~2xl 011 nlcm2/s. It is about 300 times the value of the fast neutron flux behind the bulk shield which is known to satisfy the shielding criteria. In order to reduce the fast neutron flux by 11300, about 90 cm thick SiC at 80% packing fraction is needed as the additional shield for which there is sufficient space around the TFC.

2. Streaming Effect on Biological Dose for Inboard Pipe Extraction

The result of Monte Carlo neutron transport calculation for the inboard cooling pipe extraction is shown in Table 2. Fast neutron flux with energy larger than 0.1 Me Vat the detectors are shown together with the fractional standard deviation (FSD). The non-streaming neutrons passing through cells of the outer surface of the 150 cm bulk shield are also shown. It can be seen that the fraction of penetrating flux is about 15-

273

24% in the cooling pipe further away from the pipe inlet. It should be noted that these flux value is for the case with just one cooling pipe whereas there are sixteen pipes altogether. Therefore, the flux values in the central cooling tube at F41, F42, F43, F45, F46 should be multiplied by 16 to obtain the actual flux value. At the pipe exit to the turbine room F43, the flux value becomes 3.5x106 nlcm2/s. The flux to dose conver-

Thickness Region o (em)

Plasma

Scrape-off 215 1.5 First wall

11.5 Be pebble

33.0 Breeder pebble

Material

void

void SiC (Be) 73.3%

Be 60%

Li2060%

19.0 High temperature shield SiC 80%

280

53.0 Duct space SiC 50 %

333 2.5 Shield box wall SiC 100%

66.0 Low temperature shield SiC 80%

2.5 Shield box wall SiC 100% 404 10.0 Thermal insulation void

10.0 Cryostat SS316100%

5.0 Vacuum void

10.0 Helium can SS316100%

5.0 Vacuum void

Superconducting magnet

229

Fig. 7 One dimensional model of DREAM reactor

Ratio of Shutdown Dose Rate to Fast Neutron Flux

'iii'

LE-02

LE-03

""Ei LE-04 CJ

S LE-05 ;::;. ~ LE-06

~ LE-07 o ~ I.E-OS

.--

~

~ LE-09

LE-lO 200

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~ir8tWall .. ~ / Blanket I '-' / ~ J .1DAY I W elMONTH

......... ~ ~ SCM

/' ..",./ .... V~

..-'" "-- Cryostat - I I I

250 300 350 400 450 500 Distance from Plasma Center (cm)

Fig. 8 Flux to dose conversion factor

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1.E+12

~ E ~ 1.E+11 >< ::I

:;::: t:: E "5 ~.

1.E+10

o 50 100 150 200 Distance from blanket end (cm)

Fig. 9 Fast neutron flux along the pipe outer surface

Table 2 Fast neutron flux for the inboard extraction case

Detector Fast neutron FSD Penetrating FSD Penetrating flux flux Itotal

(n/cm2/s) (%) (n/crr?/s) (%) (%)

F2 5.29E+ll 2.07% - - -

F41 1.01E+08 26.3% 3.12E+06 5.31% 3.09%

F42 2.37E+06 10.8% 4.46E+05 5.75% 18.82%

F43 2.27E+05 12.1% 3.97E+04 6.95% 17.49%

F44 8. llE+05 6.4% 1.90E+05 6.54% 23.37%

F45 4.25E+03 19.5% 6.20E+02 38.1% 14.57%

sion factor for stainless steel structure can be read as 3x10·3

from the value at the cryostat surface in Fig. 8. The flux to dose conversion factor for SiC structure can be read as 5xlO·9

from the value at the first wall in Fig. 8. Applying these fac­tors, the dose rate one day after shutdown in the turbine room becomes 10 mSv/h if the turbine structure is made of stainless steel and 0.02 mSv/h for SiC structure. If the turbine structure is made of stailess steel, there is no possibility of personnel access but with SiC structure personnel access will be permit­ted. Some measures to decrease neutron streaming into the turbine were considered but with the limited space, only 1110 reduction seemed possible.

3. Streaming Effect on Biological Dose for Outboard Pipe Extraction The result of Monte Carlo neutron transport calculation for

the outboard cooling pipe extraction is shown in Table 3. It should again be noted that these flux value is for the case with just one cooling pipe whereas sixteen pipes are connected to the ring header. Therefore, the flux values in the ring header at F44, F46, F46, F47 should be multiplied by 16 for conser-

Table 3 Fast neutron flux for the outboard extraction case

Detector Fast neutron FSD Penetrating FSD Penetrating flux flux Itotal

(n/cm2/s) (%) (n/cm2/s) (%) (%)

F2 8.75E+ll 2.31% - - -

F41 2.54E+09 15.5% 2.38E+07 7.92% 0.94%

F42 1.73E+08 9.07% 2.99E+07 4.53% 17.3%

F43 2.70E+07 6.73% 7.82E+06 7.47% 29.0%

F44 1.67E+05 6.49% 4. 12E+04 4.89% 24.7%

F45 1.64E+05 6.40% 4.lOE+04 4.90% 25.0%

F46 1.12E+04 7.53% 2.82E+03 6.23% 25.1%

F47 4.16E+05 5.88% 1.14E+05 4.61% 27.3%

vative estimate, or 3 taking only the effect of the adjacent pipes to obtain the actual flux value. At the pipe exit F46, the fast neutron flux value is 1.1x104 nlcm2/s for one pipe which corresponds to the dose rate of 33 f-LSv/h one day after shut­down for the steel structure case. This is the level at which personnel access may be permitted. If the streaming from other pipes are considered and a factor of 3-16 are multiplied, the access time will need to be restricted. For the case with SiC turbine structures, personnel access will be allowed. Here, only the activation of turbine structures are considered but in real­ity, the activation of cocrete wall, especially the gamma-rays from 24Na should be considered. With such consideration, the personnel access is permitted only after one week. It is of in­terest to note that the fraction of penetrating flux is rather large with more than 25 % further away from the inlet of the cool­ing pipe. This result suggests that bulk shield needs to be in­creased in addition to reducing the neutron streaming through the pipe. Various additional shielding measures have been considered. The most cost effective seems to be to increase

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the thickness ofthe 36 cm SiC shield above the pipe bending region in Fig. 6, to 90 cm. In this case, the dose rate at F46 becomes about 2.2 mSv/h for one pipe with steel turbine struc­ture.

VI. Conclusions

The results of the evaluation of the neutron streaming through the 80-cm diameter helium-cooling pipe of DREAM reactor are summarized as follows:

1) Additional shielding for which there is sufficient space can solve the deleterious effect on the toroidal field coils adja­cent to the pipes.

2) For inboard extraction of the cooling pipe, personnel ac­cess to the turbine room will not be permitted if steel tur­bine structures are used. If turbine structure is made of SiC personnel access will be permitted.

3) For outboard extraction of the pipe, the dose rate can be lowered so that personnel access to the turbine room will be permitted even in the case with steel turbine structure.

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Fusion Energy, Vol. 3, International Atomic Energy Agency, Vienna, 693 (1997).

(4) Briesmeister, J. F. Ed. : LA-12625-M(1997). (5) Kosako, K., et. al. : Proc. 1996 Topical Meeting on Radiation

Protection and Shielding, VoU, p. 1088-1095 (1996). (6) Seki, Y., et al. : J. Nucl. Materials, 258-263, 1791-1797 (1998). (7) Seki, Y., et al. : Fusion Technology, 34, 353-357 (1998). (8) Noda, T. : J. Nucl. Materials, 233-237, 1475-1480 (1996). (9) Seki, Y., et al. : JAERI-1301 (1986). (10) Engle, W. W. Jr. : K-1693, Oak Ridge Gaseous Diffusion Plant

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