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59 Neutronics and Shielding - Preliminary Neutronics and

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59 Neutronics and Shielding - Preliminary Neutronics and Powered By Docstoc
					                               FIRE Engineering Report
                                    August 2003
5.9    Neutronics and Shielding                    to have an operation schedule of four
                                                   pulses per day with 3 hours between
5.9.1 Introduction                                 pulses.

The design is in the preconceptual                 It is assumed that 2% of the tritium
design phase where many different                  generated in the DD pulses will burn in
design options and operating scenarios             the plasma producing some 14 MeV
are being considered. DT pulses with               neutrons. These amount to only 2% of
widths up to 20 s and fusion powers up             the 2.45 MeV neutrons produced in the
to 150 MW producing a total of 5 TJ of             DD pulses and are only 0.03% of the 14
fusion energy are planned. In addition,            MeV neutrons produced in the 150 MW
DD pulses with different widths and                DT pulse. For the total 5 TJ DT fusion
fusion powers up to 1 MW yield total               energy 1.77x1024 14 MeV neutrons are
fusion energy of 0.5 TJ. The baseline              produced. On the other hand, for the
design has a major radius of 2.14 m and            total 0.5 TJ DD pulses 4.07x1023 2.45
an aspect ratio of 3.6. The average                MeV and 8.15x1021 14 MeV neutrons
neutron wall loading during the 150 MW             are produced.
DT pulses is 1.85 MW/m2. The
corresponding values at the outboard               The impact of different FW/tile design
(OB) midplane, inboard (IB) midplane,              options on the nuclear parameters was
and divertor are 2.22 MW/m2, 1.67                  assessed [3]. The FW/tile design chosen
MW/m2, and 1.11 MW/m2, respectively.               consists of 0.5 cm plasma facing
It utilizes 16 wedged Cu TF magnets. A             component (PFC) (90% Be), 1.8 cm Cu
double walled vacuum vessel (VV) with              tiles (80% CuCrZr) and 0.2 cm gasket
integral shielding has been adopted. The           (50% Cu). A 2.5 cm water-cooled
plasma facing components include Be                CuCrZr (15% water) VV cladding is
coated Cu first wall (FW) and divertor             employed. A 1.3 cm space is provided
plates made of tungsten rods mounted on            between the gasket and VV cladding and
water-cooled Cu heat sink.                         is devoted for diagnostics. It consists of
                                                   wires, stainless steel tubes, and MgO
5.9.2 Calculation Models                           electrical insulation. A composition of
                                                   20% CuCrZr, 10% steel, 10% MgO, and
Nuclear analysis has been performed to             60% void was assumed for that zone.
evaluate the impact of design options
and assess if the major performance                Detailed neutronics calculations were
objectives of the project can be met               performed for the outer divertor that is
without jeopardizing performance of the            exposed to the most severe conditions in
radiation sensitive components. The                the divertor region. The front layer is a
neutronics and shielding calculations              0.5 cm W Brush (90% W) followed by a
were performed using the ONEDANT                   0.1 cm region (84% W, 14% CuCrZr, 2%
module of the DANTSYS 3.0 discrete                 void) where the W rods are joined to the
ordinates particle transport code system           Cu heat sink. The 1.9 cm heat sink is
[1] with the most recent FENDL-2                   made of Cu finger plates (78% CuCrZr,
nuclear evaluated data [2]. Both the IB            20% water, 2% void). A 3 cm region
and OB regions were modeled                        (47% CuCrZr, 48% SS316, 5% void)
simultaneously to account for the                  represents the mechanical attachment
toroidal effects. The machine is assumed           between the Cu finger plates and the 7 cm

                                           5.9-1
                                FIRE Engineering Report
                                     August 2003
thick backing plate (84% SS316, 16%                   OB FW/tiles at midplane for the baseline
water).                                               design. Nuclear heating was calculated
                                                      also for the components in the 1.3 cm
The VV consists of 1.5 cm thick inner                 diagnostics space behind the tiles. The
and outer facesheets made of 316SS. The               nuclear heating values in these
space between the VV facesheets (VV                   components at midplane are 25.5 W/cm3
shielding zone) includes 60% 304SS and                in MgO, 23.7 W/cm3 in SS and 24.9
40% water except in the IB region where               W/cm3 in Cu. Fig. 5.9.2 gives the radial
11% 316SS and 89% water is used                       variation of nuclear heating in the VV at
because of the small thickness. The                   the OB midplane. Nuclear heating in the
thicknesses of the VV at the IB midplane,             VV drops by an order of magnitude in
OB midplane, and divertor region are 5,               ~18 cm. Table 5.9.2 lists the peak
54, and 12 cm, respectively. A 1.5 cm                 nuclear heating values at the outer
thick layer of thermal insulation (10%                divertor. Relatively high nuclear heating
Microtherm insulation) is attached to the             is deposited in the W PFC. Fig. 5.9.3
back of the coil-side VV facesheet. The               shows the nuclear heating distribution in
Cu TF coils are included in the model                 the outer divertor plate.
with 90% packing factor. While
beryllium copper is used in the inner legs,              Table 5.9.1. Peak Nuclear Heating
OFHC copper in utilized in the rest of the                     (W/cm3) at Midplane
TF coils. A 316SS coil case is used in the
OB region with 4 cm front thickness and                                      IB        OB
6 cm back thickness.                                   Be PFC               20.65     22.07
                                                       Cu Tiles             29.08     28.71
5.9.3 Nuclear Heating                                  Gasket               25.17     25.17
                                                       Cooled Cu Vessel     24.92     24.86
Nuclear heating deposited in the                       Cladding
different components was determined                    H2O FWCoolant        17.11     19.16
and used in the thermal analysis. The                  SS Inner VV Wall     20.96     19.16
calculations were performed for the DT                 SS VV Filer          20.40     17.67
pulses with 150 MW of DT fusion                        H2O VV Coolant        9.24      9.61
power to determine the largest nuclear                 SS Outer VV Wall     18.79     0.043
heating generated. Nuclear heating
                                                       Microtherm            6.08     0.012
results can be modified for lower power                Insulation
pulses by scaling linearly with the fusion             SS Inner Coil Case    NA       0.024
power. For the DD pulses with the
                                                       Cu Magnet            12.09     0.012
largest fusion power (1 MW), nuclear
                                                       SS Outer Coil Case    NA     1.73x10-5
heating values are at least two orders of
magnitude lower than the values for the
150 MW DT pulses.                                        Table 5.9.2. Peak Nuclear Heating
                                                           (W/cm3) at the Outer Divertor
Table 5.9.1 gives the peak power density
values in the different components at the              W rods in divertor            30.4
chamber midplane. The largest power                    Cu heat sink in divertor      10.7
density values in the magnet occur in the              SS structure in divertor       9.2
IB region at midplane. Fig. 5.9.1 gives                SS VV                          4.2
the nuclear heating distribution in the                Cu Magnet                      1.1

                                              5.9-2
                                                        FIRE Engineering Report
                                                             August 2003
The amount of nuclear heating in the TF                                      breakdown of total magnet nuclear
magnets      strongly influences       the                                   heating. The total nuclear heating in the
achievable pulse length. The radial                                          16 TF coils during the DT shots is 15.5
variation of nuclear heating in the IB                                       MW with 91% of it deposited in the
magnet at midplane is shown in Fig.                                          lightly shielded IB legs.
5.9.4 for the DT and DD pulses. Nuclear
heating in the IB magnet drops by an                                               Table 5.9.3. Total Magnet Nuclear
order of magnitude in ~ 28 cm. The total                                                    Heating (MW)
nuclear heating in the 16 TF coils for
150 MW DT fusion power was                                                          IB region                          14.2
estimated based on the results of the 1-D                                           OB region                          0.03
calculations taking into account the                                                Divertor region                     1.3
poloidal variation of neutron wall                                                  Total                              15.5
loading, shielding thickness, and magnet
toroidal coverage. Table 5.9.3 gives the



                                      50
                                                    150 MW DT Fusion Power
                                                          2
                                                2.22 MW/m Neutron Wall Loading
                                                   Water Cooled Vessel Cladding
                                      40                                                                   3
                                                                                                   W/cm        of Be
              Power Density (W/cm3)




                                                                                                           3
                                                                                                   W/cm        of Cu
                                                                                                           3
                                                                                                   W/cm        of w ater

                                      30



                                      20



                                                                 Gasket (50% Cu)
                                      10       Be PFC (90% Be)


                                                        Cu Tiles (80% Cu)             Ves sel Cladding
                                                                                    (80% Cu, 15% w ater)
                                       0
                                           0             1              2               3             4                       5
                                                             Depth in OB FW/Tiles (cm)

                                      Fig. 5.9.1. Nuclear heating distribution in the OB FW/tiles.




                                                                     5.9-3
                                           FIRE Engineering Report
                                                August 2003


                                                                    150 MW DT Fusion Power
                         101                                              2
                                                                2.22 MW/m Neutron Wall Loading
                                                                   Water Cooled Vessel Cladding




 Power Density (W/cm3)
                                                                                               3
                                                                                        W/cm       of SS
                                                                                               3
                                                                                        W/cm       of Water

                         100




                         10-1                                                                         VV Outer Wall

                                    VV Inner Wall

                                                            VV S hielding Zone
                                                           60% SS, 40% Water
                         10-2
                                0             10           20          30          40                 50         60

                                               Depth in OB VV at Midplane (cm)

                         Fig. 5.9.2. Nuclear heating distribution in the OB VV.

                         40

                         35                              150 MW DT Fusion Power
Power Density (W/cm3)




                         30                          1.1 MW/m 2 Neutron Wall Loading

                         25
                                                                             W/cm3 of 316SS
                         20                                                  W/cm3 of Cu
                                                                             W/cm3 of W
                         15                                                  Total Power Density


                         10

                          5
                                      Heat         Mechanical           Backing Plate
                                W     Sink         Attachment
                          0
                                0         2            4           6        8           10              12       14
                                       Depth in Outer Divertor Plate (cm)
                                Fig 5.9.3. Nuclear heating in the outer divertor.




                                                           5.9-4
                                                              FIRE Engineering Report
                                                                   August 2003

                                                 2
                                            10




               of Cu)
                                                                                 FIRE Inboard Magnet at Midplane
                                                 1
                                            10
              3
               Magnet Power Density (W/cm
                                                 0
                                            10



                                                 -1
                                            10                                    1 MW DD Pulse
                                                                                  150 MW DT Pulse


                                                 -2
                                            10



                                                 -3
                                            10
                                                      0       10         20            30           40             50
                                                          Depth in IB Magnet at Midplane (cm)

                                             Fig 5.9.4. Radial variation of nuclear heating in IB magnet.


5.9.4 Vacuum Vessel Radiation                                                      Table 5.9.4. Peak End-of-life He
      Damage                                                                          Production (appm) in VV

Since the VV is protected from the                                                     IB midplane         0.091
fusion neutrons by the thin FW/tiles, the                                              OB midplane         0.125
issue of reweldability was addressed.                                                  Divertor            0.013
The end-of-life helium production in the
VV structure should be limited to 1                                            5.9.5   Copper Radiation Damage
appm to allow for rewelding. This is the
limit used in ITER [4]. Table 5.9.4 gives                                      Table 5.9.5 gives the end-of-life peak dpa
the results at different poloidal locations                                    values in the Cu tiles, VV cladding, Cu
for the FIRE goal of cumulative 5 TJ DT                                        heat sink in outer divertor, and Cu TF
and 0.5 TJ DD fusion energy. The                                               coils for the FIRE baseline design.
contribution from DD shots is very small                                       Although the damage levels are very low,
(< 0.15%). Lower VV He production                                              significant effects on physical and
occurs in the divertor region as a result                                      mechanical properties might occur. These
of shielding by the relatively thick                                           effects are strongly dependent on
divertor plate. The results imply that                                         irradiation temperature and have been the
reweldability of the VV should not be a                                        subject of numerous studies as part of the
concern.                                                                       ITER R&D program [5].


                                                                       5.9-5
                                FIRE Engineering Report
                                     August 2003
Radiation embrittlement at T < 150C is              The effect of irradiation on the creep of
a concern for Cu alloys with reductions              Cu alloys is uncertain due to limited
in tensile ductility (uniform elongation)            data. The extremely low doses expected
below 5% being observed for damage                   in FIRE reduce the importance of
levels on the order of 0.01 dpa.                     irradiation creep. The magnitude of the
However, the fracture toughness is                   irradiation creep can be estimated using
typically maintained at a sufficiently               the creep compliance coefficient B.
high level, at least in precipitation                Using a conservative B value of ~ 3x10-6
hardened alloys such as CuCrZr and                   MPa-1dpa-1 and the peak cumulative dpa
CuNiBe. It is possible to maintain the               value in FIRE (0.03 dpa), the irradiation
high tensile ductility by periodically               creep for an applied stress of 100 MPa
annealing the Cu at ~ 300C for ~ 50 hr.             would amount to a total deformation of
Irradiation to ITER doses of 1-10 dpa at             only 10-5 (0.001%) at end-of-life in
higher     temperatures      showed     a            FIRE. It was recommended [5] that the
pronounced increase in the uniform                   operation temperature of high strength
elongation of CuCrZr compared with                   Cu alloys should be limited to < 300C
irradiation at lower temperatures.                   for applied stresses of 100-200 MPa to
However, at T > 300C this is                        have tolerable irradiation and thermal
accompanied by significant softening.                creep at ITER conditions of 1-10 dpa.
This was demonstrated by about an order              The thermal creep strength begins to
of magnitude loss of yield strength at               decrease rapidly for temperatures >
about 300C for 2-10 dpa.                            300C. This might cause deformation in
                                                     the Cu during extended operation (> 100
 Table 5.9.5. Peak End-of-life Cu dpa                hr) at 300C. Due to the low doses in
                                                     FIRE significant deformation from
                       Total dpa in Cu at            irradiation creep is not anticipated. Some
                          end-of-life                thermal creep deformation in Cu alloys
IB Tiles                     0.0271                  might occur if operated at elevated
OB Tiles                     0.0298                  temperatures (>300C). There is a lack
Divertor                     0.0125                  of detailed studies on fatigue, fracture
IB VV Cladding               0.0178                  toughness and fatigue crack growth rate
OB VV Cladding               0.0204                  behavior     in    high-strength,     high-
Magnet at IB                 0.0055                  conductivity copper alloys [6,7].
Magnet at OB               6.26x10-6
Magnet at Divertor         3.78x10-4                 The Cu alloys operate at different
                                                     temperatures in the FIRE components.
Void swelling takes place in copper                  The tiles can get to temperatures over
alloys irradiated in the temperature range           400C. The tiles carry no primary
of 180 to 530C. While void swelling is              stresses and should be basically
pronounced in copper containing oxygen               unloaded except for thermal stresses and
impurities (~ 2.5%/dpa), it is only ~                disruptions. Therefore, problems with
0.5%/dpa in pure Cu and is generally                 high-temperature softening and creep
insignificant in Cu alloys up to doses of            should not be of concern. In addition, the
60 dpa. Therefore, for the low dose                  tiles can be easily replaced if needed.
levels in FIRE, void swelling is not a               The temperature of the VV Cu cladding
concern.                                             is lower than 250C. At this peak

                                             5.9-6
                               FIRE Engineering Report
                                    August 2003
temperature, occurring at midplane, the            5.9.6 Radiation Induced Resistivity in
low-temperature      embrittlement   for                 the Cu Conductors of the TF
CuCrZr is not an issue. That will be a                   Coils
concern only for the lower temperature
parts of the cladding at the top and               The 17510 BeCu alloy is used in the
bottom of the chamber. However, the                inner legs of the TF coils with 10200
dpa level will also be lower at these              OFHC copper being utilized in the rest
locations resulting in alleviating the             of the TF coils. A concern with Cu
embrittlement concern. We also have the            magnet conductors is the increased
option of annealing out the copper                 electrical resistivity that impacts the
damage if we bake the vessel to >                  performance of the TF coils through
300C. The Cu in the divertor will have            increasing the I2R heating and re-
peak temperatures close to 500C. The              distributing the current across the coil.
peak damage level is only 0.015 dpa.               The temperature at the end of the pulse
The issue here will be mainly thermal              will increase with possible impact on the
creep. The temperature of the TF coils             achievable pulse length. The increase in
rises from 80 to 373 K during each                 electrical resistivity results from solute
pulse. The main issue here is the low-             transmutation products and displacement
temperature embrittlement. The low                 damage. In a low fluence machine like
temperature embrittlement data on                  FIRE resistivity increase is dominated
CuNiBe and OFHC Cu are limited to                  by point defects and defect clusters
tensile tests between room temperature             produced by displacement damage.
and 100C [8,9]. The concern is
                                                   The solute transmutation component of
primarily at the IB midplane where the
                                                   resistivity    increase     is    directly
peak damage rate is ~ 0.007 dpa, which
                                                   proportional to the solute content. We
is at the lower range of damage for the
                                                   calculated     the    concentration     of
occurrence of radiation embrittlement.
                                                   transmutation products accumulating at
Much lower damage levels occur at
                                                   the end-of-life as a function of position
other locations of the TF coil.
                                                   in the TF coils. The BeCu alloy includes
Based on the irradiation levels and                1.8% Ni and 0.4% Be and the
operation conditions in FIRE and the               transmutation products are dominated by
available data on Cu alloys, we can                Ni, Zn, Co, Fe, and H. The worst
identify the R&D needs as follows:                 conditions for the BeCu occur in the IB
 Data on loss of ductility of BeCu (or            leg at midplane where the peak
    OFHC) at temperatures between 80               cumulative dpa is 0.0055. For the OFHC
    and 373 K with doses < 0.01 dpa.               copper, the worst conditions are behind
                                                   the divertor at the top/bottom of the
 Data on fatigue, fracture toughness
                                                   machine where the peak cumulative dpa
    and fatigue crack growth rate
                                                   is 3.78x10-4. Table 5.9.6 gives the peak
    behavior in high-strength, high-
                                                   concentration of the transmutation
    conductivity copper alloys.
                                                   products at these poloidal locations. The
 Thermal creep data for CuCrZr at
                                                   Ni included in the unirradiated BeCu
    temperatures up to 500C. There is             transmutes at a rate comparable to the
    no need to perform irradiation creep           production rate of Ni from Cu resulting
    measurements on Cu alloys for the              in negligible net production of Ni. Fe is
    low doses proposed in FIRE.                    produced from the transmutation of Ni.

                                           5.9-7
                                  FIRE Engineering Report
                                       August 2003
The resistivity increases associated with              Therefore, the largest resistivity increase
these concentrations were calculated                   from transmutation effects is 0.783 p-
using the solute resistivities of 1.12, 0.3,           m in the BeCu alloy and 0.124 p-m in
6.4, 9.3, and 1.5 p-m/appm for Ni, Zn,                the OFHC copper. These values are
Co, Fe, and H, respectively [8]. The                   much smaller than the unirradiated
results are given in Table 5.9.6. Due to               resistivities in the temperature range of
the high mobility of hydrogen in copper                the FIRE magnets. In addition, the
at room temperature, it diffuses out of                neutron flux drops as one moves deeper
the material as the magnets heat up at the             in the magnet resulting in decreasing
end of each pulse. Thermal annealing is                transmutations and lower resistivity
not expected to cause any changes in the               increase.
resistivity increase contributed by the
other solute transmutation products.

 Table 5.9.6. Peak Resistivity Increase from Solute Transmutation Products at Different
                          Locations of TF Coils at End-of-life

  Solute    Peak solute concentration (appm)             Peak resistivity increase (p-m)
              BeCu alloy      OFHC copper                BeCu alloy          OFHC copper
             IB midplane     Behind divertor            IB midplane         Behind divertor
  Ni              0               0.0867                     0                   0.0970
  Zn            1.602             0.0591                   0.481                 0.0178
  Co            0.034             0.0014                   0.217                 0.0090
  Fe            0.009                0                     0.085                    0
  H             0.261             0.0147                   0.393                 0.0221

An analytical formula given by Eq. 1                   which injects solute into the copper
[10] can be used to estimate the                       matrix. Based on available experimental
resistivity increase due to displacement               data, we used saturation values
damage for copper and Cu alloys. At                    (parameter A in Eq. 1) of 1.2 and 4.2
high doses the displacement damage                     nΩ-m      for   OFHC       and     BeCu,
component approaches a constant                        respectively. These include the effect of
saturation value due to displacement                   annealing during the pulse as the magnet
cascade       overlap     effects.     The             warms up.
recommended value of the parameter B
is 100 for both BeCu and OFHC copper                                    -B*dpa
                                                              ∆~ A[1-e       ]              (1)
under FIRE operating conditions [11].
The parameter A represents the                         Periodic annealing of the copper
saturation resistivity. Based on electrical            components at temperatures well above
resistivity measurements, the expected                 recovery Stage V (~ 425 K,
saturation resistivity increase for pure               corresponding to thermal dissociation of
Cu irradiated near room temperature is                 vacancy clusters [13]) would cause a
1.2 nΩ-m [12]. Higher resistivity                      significant additional reduction in the
increases have been measured for Hycon                 displacement damage component of the
CuNiBe [6] due to partial dissolution of               resistivity increase beyond that already
precipitates by displacement cascades,                 achieved at room temperature. It is

                                               5.9-8
                                                                     FIRE Engineering Report
                                                                          August 2003
speculated that up to ~ 90% of the                                                        resistivity increase is much lower. In
displacement damage component of the                                                      addition, the resistivity increase drops as
resistivity increase present in Cu                                                        one moves from the plasma side of the
specimens     irradiated   near   room                                                    coil deeper in the magnet.
temperature could be recovered by
annealing near 573 K [11]. Hence, if the                                                  Fig. 5.9.5 shows the resistivity increases
magnets of FIRE can be baked-out to                                                       at end-of-life from solute transmutation
temperatures above 200C (preferably                                                      products and displacement damage in
300C), we can significantly reduce the                                                   the BeCu alloy as a function of depth in
resistivity increase. We made the                                                         the inner leg of the TF magnet at
conservative assumption that such bake-                                                   midplane. It is clear that the total
out is not employed.                                                                      resistivity increase is dominated by
                                                                                          displacement damage with the resistivity
Using Eq. 1 along with the cumulative                                                     increase from solute transmutation
atomic      displacement    values,    we                                                 products contributing less than 0.05% of
determined the largest increase in BeCu                                                   the total resistivity increase. The
resistivity to be 1.79 nΩ-m in the inner                                                  resistivity increase drops by a factor of ~
leg at midplane. The largest increase in                                                  30 across the magnet thickness.
OFHC copper resistivity is only 0.044                                                     Resistivity     increase    from     solute
nΩ-m behind the divertor at the                                                           transmutation products contributes less
top/bottom of the machine. Notice that                                                    than 0.3% of the total resistivity increase
these resistivity increases occur during                                                  of the OFHC copper conductor. The
pulses near the end of the machine’s life.                                                spatial distribution of the resistivity
Early in the life of the machine, the                                                     results in re-distributing the current
accumulated dpa is very low and the                                                       across the coil.
                  101
          Increase in Electrical Resistivity (n -m)




                                                      100
                                              




                                                      10-1

                                                                 Total Fusion Energy of 5 TJ DT and 0.5 TJ DD
                                                      10-2              Inboard BeCu Magnet at Midplane



                                                      10-3


                                                      10-4
                                                                 From Transmutation Prod ucts
                                                                 From Displacemen t Damag e
                                                      10-5
                                10        20          30     0   40         50
                             Depth in IB Magnet at Midplane (cm)
Fig. 5.9.5. Resistivity change in BeCu as a function of depth in magnet at IB midplane.

                                                                                  5.9-9
                                 FIRE Engineering Report
                                      August 2003
The unirradiated resistivity of the 68%                5.9.7 Magnet Insulator Dose
IACS BeCu used in FIRE varies from ~
10 nΩ-m at 80 K to ~ 30 nΩ-m at room                   The insulator dose rate in the TF magnet
temperature [14,15]. This implies that                 was calculated at the front layer of the
the maximum increase in resistivity of                 magnet winding pack. For 5 TJ of DT
the BeCu at end-of-life varies from ~                  fusion energy and 0.5 TJ of DD fusion
18% at the start of the pulse to ~ 6% at               energy, Table 5.9.7 provides the peak
the end of the pulse. The unirradiated                 cumulative magnet insulator dose in the
resistivity of the 10200 OFHC copper                   IB, OB, and divertor regions. The peak
used in FIRE ranges from ~ 2 nΩ-m at                   cumulative magnet insulator dose is
80 K to ~ 16 nΩ-m at room temperature                  1.05x1010 in the lightly shielded IB leg
[14,16]. Hence, the maximum increase                   at midplane. At this location the fast
in resistivity of the OFHC copper at end-              neutron fluence (E >0.1 MeV) is
of-life varies from ~ 2% at the start of               8.1x1018 n/cm2 and the total neutron
the pulse to ~ 0.25% at the end of the                 fluence is 1.5x1019 n/cm2. About 55% of
pulse. The unirradiated resistivities                  the neutron fluence is above 0.1 MeV at
given above are without applied                        the front of the magnet and drops to 35%
magnetic       field.   Magnetoresistance              at the back of the magnet. The DD shots
effects can lead to significant resistivity            contribute 13% of this value. The dose
increase at cryogenic temperatures of up               rate decreases by three orders of
to a factor of 10 depending on the                     magnitude as one moves poloidally to
magnetic field and Cu purity [17,18].                  the OB midplane. The relative DD
However,              much         smaller             contribution decreases as one moves
magnetoresistivity increases occur in                  poloidally from the IB midplane to only
copper components operating at the                     1.6% at the OB midplane due to
higher temperatures of FIRE. The largest               increased attenuation of low energy DD
magnetoresistivity effects occur in the                neutrons. The insulator dose decreases as
inner leg. For the peak magnetic field of              one moves radially from the front to the
15 tesla the peak magnetoresistivity                   back of the winding pack as shown in
increase in the BeCu is ~ 7% at the start              Fig. 5.9.6. The gamma contribution to
of the pulse and drops to ~ 2% at the end              the total dose is 35-50%. The dose
of the pulse when the magnet heats up.                 decreases by an order of magnitude in ~
The magnetoresistivity effects also                    22 cm of the IB magnet. The radial
decrease as one moves deeper in the                    variation of he neutron and gamma
inner leg of the coil due to the                       fluence in the IB leg of the TF coils is
decreasing magnetic field. These                       shown in Fig. 5.9.7.
increases are the same during all pulses
over the life of the machine. The                       Table 5.9.7. Cumulative Peak Magnet
magnetoresistivity effects are lower than                      Insulator Dose (Rads)
those resulting from irradiation effects
during pulses at the end of the machine                                    Total      % from
life. However, during pulses early in the                                  Dose      DD Shots
life of the machine magnetoresistivity                                    (Rads)
effects yield resistivity increases larger             IB midplane      1.05x1010      13%
than those resulting from irradiation                  OB midplane      1.05x107       1.6%
effects.                                               Divertor         8.10x108       10%

                                              5.9-10
                                                         FIRE Engineering Report
                                                              August 2003

                                              1011


                                                         Total Fusion Energy of 5 TJ DT and 0.5 TJ DD




           End-of-life Insulator Dose (Rad)
                                              1010




                                               109




                                               108            Neutron Dose
                                                              Gamma Dose
                                                              Total Dose



                                               107
                                                     0   10            20           30             40             50

                                                           Depth in IB Magnet at Midplane (cm)


      Fig. 5.9.6. Radial variation of insulator dose in the IB magnet.

                                              1020


                                                         Total Fusion Energy of 5 TJ DT and 0.5 TJ DD
       End-of-life Fluence (particles/cm )
                                        2




                                                                                 Fast Neu tro ns (E >0 .1 Me V)
                                              1019                               Total Neutrons
                                                                                 Total Gamma Photons




                                              1018




                                              1017
                                                     0   10             20           30             40            50

                                                              Depth in IB Magnet at Midplane (cm)


Fig. 5.9.7. Radial variation of neutron and gamma fluence in the IB magnet.




                                                                       5.9-11
                                 FIRE Engineering Report
                                      August 2003
The mechanical strength, dielectric                    whole device lifetime with the proposed
strength, and electric resistivity are the             operation scenario and load conditions.
important properties that could be
affected by irradiation. The shear                     5.9.8 Source Geometrical Effects on
strength is the property most sensitive to                   Peak Nuclear Parameters in
irradiation. The commonly accepted                           FIRE TF Coils
dose limit for epoxies is 109 Rads. This
is the limit used in ITER [4]. Polyimides              Neutronics calculations performed using
and bismaleimides are more radiation                   a one-dimensional cylindrical model that
resistant with experimental data showing               includes both the IB and OB regions
only a small degradation in shear                      account for the toroidal effect and the
strength at dose levels in excess of 1010              impact of each region on the nuclear
Rads. However, they are difficult and                  parameters in the other. However,
expensive to process due to their high                 important three-dimensional effects are
viscosity and requirement for high                     not included. In the 1-D model the
temperatures to fully cure. In addition,               volumetric neutron source in the plasma
they have initial mechanical properties                zone is assumed to be uniform and
lower than those achievable with                       extended infinitely in the vertical
epoxies. Hybrids of epoxies, polyimides,               direction. In the real 3-D geometry, the
bismaleimides, and cyanate esters are                  neutron source has finite vertical
being developed to both improve the                    extension with a spatial profile that
insulation system’s ability to withstand               peaks near the plasma major radius at
high levels of radiation and to improve                the reactor midplane. As a result, the
the resin system’s overall processibility              angular distribution of source neutrons
[19].     That     includes     irradiation            incident on the first wall will be quite
measurements at dose levels close to                   different. In the 1-D model most source
those expected in FIRE [20]. The                       neutrons are incident at a glancing angle
information provided here gives                        compared to the mostly perpendicular
guidance to the insulator development                  distribution in the exact 3-D geometry.
R&D program regarding the dose levels                  Consequently, the 1-D model is expected
and the proper mix between neutron and                 to yield higher nuclear parameters near
gamma irradiation to insure relevance to               the first wall. The 3-D and 1-D results
the FIRE conditions.                                   for ARIES-ST and ARIES-AT were
                                                       compared and the 1-D approximation
In the FIRE design with wedged coils                   was found to overestimate the peak
and added compression ring, the TF                     damage parameters at the first wall by
inner leg insulation does not have to                  up to a factor of ~1.5 depending on the
have significant bond shear strength,                  aspect ratio [21]. The IB legs of the TF
which is most sensitive to radiation. The              coils in FIRE are lightly shielded with
peak torsional shear stresses occur at the             relatively large nuclear heating and
top and bottom of the IB leg behind the                insulator dose that impose restrictions on
divertor. The end-of-life insulator dose               the achievable pulse length and device
at these locations is reduced to ~ 109                 lifetime. We performed calculations to
Rads due to the additional shielding                   assess the impact of the source
provided by the divertor. Based on that,               geometrical effects on the peak nuclear
it is expected that insulation materials               parameters in the FIRE TF coils
will be identified that can last for the               occurring at the IB midplane.
                                              5.9-12
                                     FIRE Engineering Report
                                          August 2003
                                                                                     19
A     two-dimensional       model      was               is normalized to 2.66x10 14.1 MeV
developed for FIRE using the                             neutrons per second. The calculations
TWODANT module of the DANTSYS                            were performed also for the DD shots
discrete ordinates particle transport code               with 1 MW fusion power. In this case, in
system [1]. The 2-D model is adequate                    addition to the 2.45 MeV DD neutrons
because      of    the      axi-symmetric                produced some 14.1 MeV DT neutrons
configuration around the major axis. In                  will be produced due to the 2% burnup
addition, only the upper half of the                     of produced tritium. For the DD shots,
machine is modeled because of                            the neutron source in the upper half of
symmetry around the midplane. The                        the plasma is normalized to 4.08x10
                                                                                                  17
plasma boundary used is shown in Fig.                                                        15
                                                         2.45 MeV neutrons and 8.15x10 14.1
5.9.8. The neutron source density profile
                                                         MeV neutrons per second. Since, the
was approximated by the parabolic
                                                         primary interest is in the peak nuclear
formula given in Eq. 2.
                                                         parameters at the IB midplane, the
                      2 4.75                             detailed radial build in the IB region was
S(r) = S(0) [1-(r/a)   ]                  (2)            modeled. On the other hand, only an
                                                         averaged       homogenized         material
For the DT pulses with fusion power of                   composition was used in the OB and top
150 MW, the total neutron source                         regions.
generated in the upper half of the plasma
                    1.20

                    1.08

                    0.96

                    0.84

                    0.72
            Z (m)




                    0.60

                    0.48

                    0.36

                    0.24

                    0.12

                    0.00
                           1.56 1.68 1.80 1.92 2.04 2.16 2.28 2.40 2.52 2.64
                                                   R (m)


                         Fig. 5.9.8. Plasma boundary in FIRE.
Fig. 5.9.9 shows the radial variation of       midplane during a 150 MW DT shot.
nuclear heating in the IB Cu magnet at         The results are compared in Fig. 5.9.10

                                                5.9-13
                                                                 FIRE Engineering Report
                                                                      August 2003
to the results obtained from the 1-D                                                ~4 as one moves to the top and bottom
calculation with an IB neutron wall                                                 of the IB leg of the TF coil as illustrated
loading of 1.67 MW/m2 at midplane.                                                  in Fig. 5.9.14.
The peak value is 9.25 W/cm3 which is
~23% lower than that obtained from the                                              Notice that the peak insulator dose of
1-D calculations. Nuclear heating drops                                             8.71x109 Rads is based on a total of 5 TJ
also as one moves away from the                                                     DT and 0.5 TJ DD fusion energies. We
midplane as shown in Fig. 5.9.11. The                                               have 1.57x109 Rads per TJ of DT fusion
total end-of-life cumulative insulator                                              energy and 1.7x109 Rads per TJ of DD
dose was calculated for a total of 5 TJ                                             fusion energy. These values can be used
DT and 0.5 TJ DD fusion energies. The                                               to determine the peak insulator dose for
radial variation is given in Fig. 5.9.12 at                                         any combination of DT and DD pulses.
the IB midplane. The contributions from                                             For the current baseline experimental
the DT and DD shots are shown                                                       plan with 6.336 TJ DT and 0.049 TJ DD
separately. The results are compared to                                             the peak insulator dose is 1010 Rads. For
the 1-D results in Fig. 5.9.13. The peak                                            the proposed extended operation
insulator dose is 8.71x109 Rads which is                                            schedule with 12.771 TJ DT and 0.185
~17% lower than that predicted by the 1-                                            TJ DD the peak dose will be 2.04x1010
D model. This value drops by a factor of                                            Rads.

                                                  101

                                                                           150 MW DT Pulse
                                                                        Inboard Leg at Midplane
             Magnet Power Density (W/cm of Cu )
            3




                                                  100




                                                  10-1



                                                             2-D Model with Correct Plasma Boundary
                                                                   and Neutron Source Profile


                                                  10-2
                                                         0      10         20          30          40         50
                                                               Depth in IB Magnet at Midplane (cm)


         Fig. 5.9.9. Radial variation of nuclear heating in TF coil at IB midplane.



                                                                           5.9-14
                                                                                     FIRE Engineering Report
                                                                                          August 2003

                                                                    102



                                                                                                              2-D Model



                Magnet Power Density (MW/cm of Cu)
                                                                                                              1-D Model
                                                                    101
       3




                                                                    100




                                                                    10-1

                                                                                           150 MW DT Pulse
                                                                                        Inboard Leg at Midplane


                                                                    10-2
                                                                           0      10           20        30           40        50
                                                                                 Depth in IB Magnet at Midplane (cm)
Fig. 5.9.10. Comparison between 1-D and 2-D magnet nuclear heating results.
       Magnet Power Density at Front of Inboard Leg (W/cm of Cu )




                                                                    10

                                                                                              150 MW DT Pulse
      3




                                                                                           Inboard Leg at Midplane
                                                                     8




                                                                     6




                                                                     4




                                                                     2
                                                                               2-D Model with Correct Plasma Boundary
                                                                                     and Neutron Source Profile


                                                                     0
                                                                           0    20        40        60        80          100   120
                                                                                       Height above Midplane (cm)

    Fig. 5.9.11. Vertical variation of nuclear heating at front of IB TF leg.

                                                                                               5.9-15
                                                          FIRE Engineering Report
                                                               August 2003

                                          1010


                                                                                     5 TJ DT Shots
                                                                                     0.5 TJ DD Shots
                                                                                     Total




       End-of-life Insulator Dose (Rad)
                                          109




                                          108




                                          107

                                                       2-D Model with Correct Plasma Boundary
                                                             and Neutron Source Profile


                                          106
                                                 0        10         20         30            40       50
                                                         Depth in IB Magnet at Midplane (cm)

   Fig. 5.9.12. Radial variation of insulator dose at IB midplane.

                                          1011



                                                                                      2-D Model
                                                                                      1-D Model
     End-of-life Insulator Dose (Rad)




                                            10
                                          10




                                          109




                                          108

                                                     Total Fusion Energy of 5 TJ DT and 0.5 TJ DD
                                                                Inboard Leg at Midplane


                                          107
                                                 0        10         20          30            40       50
                                                         Depth in IB Magnet at Midplane (cm)

Fig. 5.9.13. Comparison between 1-D and 2-D insulator dose results.



                                                                     5.9-16
                                                                                     FIRE Engineering Report
                                                                                          August 2003
                                                                    1010




              End-of-life Insulator Dose at Front of IB Leg (Rad)
                                                                                        5 TJ DT Shots
                                                                                        0.5 TJ DD Shots
                                                                                        Total


                                                                    109




                                                                               2-D Model with Correct Plasma Boundary
                                                                                     and Neutron Source Profile


                                                                    108
                                                                           0    20        40        60         80      100      120
                                                                                       Height above Midplane (cm)

          Fig. 5.9.14. Vertical variation of insulator dose at front of IB TF leg.


5.9.9 Radiation Environment at FIRE                                                                     impurity pellet guide tubes require
      Midplane Diagnostic Ports                                                                         straight holes through the port shielding
                                                                                                        plugs. Other diagnostics, notably optical
Neutron and gamma fluxes can affect                                                                     systems such as Thomson scattering,
plasma diagnostic performance through                                                                   will make use of labyrinthine
enhanced conductivity of electrical                                                                     penetrations to curtail the streaming.
insulation    and     scintillation   and                                                               Such penetrations will include four
absorption in optical components close                                                                  bends with mirrors at the corners. A
to the tokamak. Determination of the                                                                    schematic of the diagnostics and
radiation environment is essential for                                                                  penetrations in the midplane diagnostics
estimating shielding requirements for                                                                   port plug J is shown in Fig. 5.9.15. This
diagnostic components such as insulated                                                                 port was identified as the most critical
cables, windows, fiberoptics and                                                                        diagnostic port that requires special
transducers, as well as detectors and                                                                   attention regarding radiation shielding.
their associated electronics. In addition,
streaming       through        diagnostics                                                              Two-dimensional neutronics calculations
penetrations could lead to excessive                                                                    have been performed to determine the
doses outside the machine.                                                                              nuclear radiation environment at selected
                                                                                                        locations in the diagnostics penetrations
In FIRE, a few diagnostics, such as the                                                                 and to assess the impact of streaming on
neutral particle analyzer (NPA) and                                                                     the average flux outside the port flange.

                                                                                               5.9-17
                                 FIRE Engineering Report
                                      August 2003
The neutronics calculations have been                  and flange. This is representative of the
performed using the two-dimensional                    NPA tube. The third case includes a
module of the DANTSYS neutral                          penetration with four bends and
particle transport code [1]. A simplified              represents the Thomson scattering laser
geometry was modeled in r-z cylindrical                well. Due to the limitation of two-
geometry. In the model both the inboard                dimensional modeling, all bends were
and outboard regions are modeled                       modeled in the same plane as shown in
simultaneously to properly account for                 Fig. 5.9.16. Notice that conservative
the toroidal geometry effects. The                     estimates are obtained using these two-
calculations were performed for the DT                 dimensional      models      since      the
pulses with 150 MW fusion power using                  penetrations are modeled as slots that
the midplane radial build for the FIRE                 extend toroidally and attenuation by
machine with 2.14 m major radius. The                  components in the penetrations is not
front of the 110 cm thick port plug                    included. The total neutron flux
facing the plasma is at a radius of 282.2              (integrated over all energies), the fast
cm. The radial distance between the                    neutron flux (E > 0.1 MeV) and the total
front of the port plug and the 2.5 cm                  gamma flux were calculated at the front
thick port flange is 339 cm. The port                  of the plug, along the penetrations, at the
plug and flange are assumed to consist                 back of the plug and at the back of the
of 80% steel and 20% water. Three                      port flange. In addition, the absorbed
different models were used in the                      dose rates in silica (SiO2) and alumina
calculations. The first one assumes no                 (Al2O3), were calculated. This includes
penetrations in the plug and flange. The               the contributions from both neutrons and
results from this calculation are used as a            gammas. The gamma contribution to the
reference to quantify the impact of                    dose varies from ~30% at the front of the
streaming. The second case considered is               port plug to ~80% at the back of the port
for the worst case streaming with a 10                 flange.
cm straight penetration through the plug




               Fig. 5.9.15. Schematic of diagnostics and penetrations in port plug J.

                                              5.9-18
                                 FIRE Engineering Report
                                      August 2003




      0.4 m




                                                                                             Port Flange
                    0.4 m



                                  0.3 m

    Port Shield Plug




                 1.1 m
        Fig. 5.9.16. Two-dimensional model used for the Thmpson scattering penetration.

Table 5.9.8 gives the neutron and                      port plug increases the radiation
gamma fluxes for the case without                      environment in the test cell area behind
penetrations at the front of the port plug,            the flange by about four orders of
at the back of the plug and at the back of             magnitude. This implies also that the
the port flange. The fluxes attenuate by               biological dose rate will increase
about seven orders of magnitudes in the                accordingly.
shield plug. For the case with a straight
penetration the neutron and gamma                      For the Thompson scattering penetration
fluxes along the penetration are given in              with four bends, the fluxes and absorbed
table 5.9.9. In addition, the average                  dose rates were calculated at the mirrors
values at the three radial locations are               located at each of the bends. In addition,
provided. At the back of the port plug, a              results are given at the front of the plug,
flux peaking factor of ~7 occurs due to                back of the plug, and back of the flange.
the penetration. At the back of the flange             Table 5.9.10 lists the neutron and
the peaking factor is only ~4 following                gamma fluxes along the penetration. The
neutron and gamma transport in the large               average values at the back of the plug
void space between the plug and flange.                and at the back of the flange are given in
Notice that the average flux values                    table 5.9.11. At the back of the plug, the
behind the plug and flange are about                   peak values occur at the exit of the
four orders of magnitude higher than                   penetration with a peaking factor of ~4.
those in the case without the penetration.             Since the penetration does not go
Hence, using a straight penetration in the             through the port flange, the peaking

                                              5.9-19
                                    FIRE Engineering Report
                                         August 2003
factor behind the flange is only ~2. The                 penetration       will     reduce     this
average flux values behind the plug and                  enhancement. Table 5.9.12 gives the
the flange are about a factor of 200                     absorbed dose rates in silica and alumina
higher than those without penetrations.                  along the 4-bend penetration and at the
Hence, the presence of the 4-bend                        back of the port flange. The dose rates in
penetration results in enhancing the                     the 4-bend penetration at the back of the
radiation fluxes and dose rates in the test              port plug are about two orders of
cell area by about two orders of                         magnitude less than in the case of a
magnitude. Diagnostic equipment in the                   straight penetration.

    Table 5.9.8. Neutron and Gamma Fluxes during 150 MW DT Fusion Power Pulses for Port Plug
                                      without Penetrations

                                    Total Neutron         Fast Neutron Flux       Total Gamma
                                        Flux                (E>0.1 MeV)               Flux
                                      (n/cm2s)                 (n/cm2s)             /cm2s)
Front of port plug                   8.38x1014                5.85x1014            4.94x1014
Back of port plug                     3.51x107                 1.64x107             3.04x107
Back of port flange                   1.04x107                 6.23x106             1.04x107


        Table 5.9.9. Neutron and Gamma Fluxes for the Case of Straight 10 cm Penetration

                        Total Neutron Flux           Fast Neutron Flux        Total Gamma Flux
                             (n/cm2s)                  (E>0.1 MeV)                 /cm2s)
                                                          (n/cm2s)
                         Along      Average         Along        Average       Along      Average
                      Penetration                Penetration                Penetration
Front of port plug     7.46x1014    8.38x1014     5.34x1014     5.85x1014    4.14x1014    4.94x1014
Back of port plug      2.87x1012    4.06x1011     1.28x10  12
                                                                1.57x1011    2.08x1012    2.79x1011
Back of port flange    5.14x1011    1.27x1011     2.75x10  11
                                                                6.38x1010    3.96x1011    1.03x1011

 Table 5.9.10. Neutron and Gamma Fluxes during 150 MW DT Fusion Power Pulses for the 4-
                                    Bend Penetration

                                    Total Neutron         Fast Neutron Flux       Total Gamma
                                        Flux                (E>0.1 MeV)               Flux
                                      (n/cm2s)                 (n/cm2s)             /cm2s)
Entrance at front of port plug       7.48x1014                5.34x1014            4.15x1014
First bend                           1.20x1014                6.24x1013            7.68x1013
Second bend                          2.06x1013                7.95x1012            1.37x1013
Third bend                           2.57x1012                9.19x1011            1.79x1012
Fourth bend                          2.87x1011                6.75x1010            2.15x1011
Exit at back of port plug            2.92x1010                 6.82x109            2.46x1010
Peak at back of port flange           5.00x109                 1.50x109             5.00x109



                                                5.9-20
                                  FIRE Engineering Report
                                       August 2003
     Table 5.9.11. Average Neutron and Gamma Fluxes at Backs of Plug and Flange with a
                                  Penetration with 4 Bends

                                  Total Neutron          Fast Neutron Flux       Total Gamma
                                      Flux                 (E>0.1 MeV)               Flux
                                    (n/cm2s)                  (n/cm2s)             /cm2s)
Back of port plug                   6.98x109                  2.26x109             5.97x109
Back of port flange                 1.99x109                  8.84x108             2.10x109

Table 5.9.12. Absorbed Radiation Dose Rate in Silica and Alumina along the 4-Bend Penetration

                                                  Dose Rate During 150 MW DT Pulses
                                                                (Rad/s)
                                                      Silica            Alumina
                                                             5
           Front of port plug                       7.90x10             8.48x105
           First bend                               7.69x104            9.85x104
                                                             4
           Second bend                              1.01x10             1.76x104
           Third bend                               1.24x103            2.25x103
                                                             2
           Fourth bend                              1.41x10             2.67x102
           Back of port plug                        1.65x101            3.14x101
           Back of port flange                        3.80                7.07

Fig. 5.9.17 shows the vertical                          biological dose rates behind the port
distribution of the fast neutron flux at the            flange were obtained from activation
back of the port plug for the three cases               calculations performed with a 1.1 m port
considered. While the flux is nearly                    plug without any penetrations [3]. These
uniform for a plug without any                          dose rates are ~10 mrem/hr at shutdown,
penetrations, using a straight penetration              ~3 mrem/hr after 1 hr, ~0.1 mrem/hr
results in significant peaking at the exit              after 1 day, and about 0.05 mrem/hr after
of the penetration. For the penetration                 1 week. These acceptable dose rates can
with 4 bends, the flux peaks at the exit of             be maintained in the cases where
the penetration with a small local                      penetrations are employed in the shield
peaking behind the third bend. The                      plug by increasing the plug thickness.
figure indicates also the relative                      Past activation calculations indicated
enhancement in the average flux level                   that the dose rate scales linearly with the
behind the port plug resulting from                     total neutron flux provided that the
these penetrations.                                     material used in the area where the dose
                                                        is calculated is not changed. We
Since the flux values in the test cell area             performed two-dimensional neutronics
behind the port flange are increased                    calculations     with    different     plug
significantly due to streaming, the                     thicknesses for the cases with straight
biological dose rate will be enhanced                   and 4-bend penetrations. Table 5.9.13
affecting the accessibility for hands-on                gives the results for 1.1 m plug and for
clearing      services       prior       to             the case where all the port is filled by the
removal/installation of port assemblies                 plug. The peak and average flux levels
by remote handling means. Acceptable                    are shown behind the port flange. The

                                               5.9-21
                                                           FIRE Engineering Report
                                                                August 2003
results for the 1.1 m plug without                                            results imply that if a straight 10 cm
penetrations     are      included     for                                    penetration is employed in the port plug,
comparison. For the straight penetration,                                     the plug thickness needs to be increased
the flux peaking factor increases from ~4                                     to ~3.1 m to ensure that the dose rates
with 1.1 m plug to ~15 with 3.4 m plug.                                       after shutdown in the area behind the
On the other hand, for the 4-bend                                             port flange are similar to the acceptable
penetration, the flux peaking factor                                          levels obtained without penetrations. For
increases from ~2 with 1.1 m plug to                                          the case with the 4-bend penetration
~13 with 3.4 m plug. The average                                              employed, the plug thickness should be
neutron flux behind the flange decreases                                      increased to ~2.1 m. Since both
by an order of magnitude by making the                                        penetrations exist in the diagnostic port
shield plug with straight penetration                                         J, it is recommended that the port plug
thicker by ~0.5 m. The same drop in flux                                      thickness should be increased to 3.4 m,
is obtained by ~0.42 m thicker plug for                                       the location of the cryostat interface.
the case with 4-bend penetration. The


                                       1013
                                                  Back of 1.1 m port plug
                                                    150 MW DT pulse                           4-bend penetration
                                       1012                                                   Straight penetration
                                                                                              No penetrations
          Fast Neutron Flux (n/cm s)
          2




                                       1011



                                       1010



                                       109



                                       108                             Exit of straight   Exit of 4-bend
                                                                        penetration         penetration



                                       107
                                              0             50              100                  150                 200
                                                                 Vertical Location (cm)


        Fig. 5.9.17. Spatial distribution of fast neutron flux behind the port plug.


Table 5.9.13. Neutron and Gamma Fluxes at the Back of the Port Flange During 150 MW
               DT Fusion Power Pulses with Different Plug Thicknesses

                                                                     5.9-22
                                FIRE Engineering Report
                                     August 2003


                      Total Neutron Flux        Fast Neutron Flux           Total Gamma Flux
                           (n/cm2s)                (E>0.1 MeV)                   /cm2s)
                                                     (n/cm2s)
                       Peak      Average         Peak      Average           Peak      Average
1.1 m plug           1.04x107    1.04x107      6.23x10 6
                                                          6.23x106         1.04x107    1.04x107
without
penetrations
1.1 m plug with      5.14x1011 1.27x1011 2.75x1011 6.38x1010 3.96x1011 1.03x1011
10 cm straight
penetration
3.4 m plug with      4.18x107    2.64x106      1.08x107       7.51x105     4.28x107    2.96x106
10 cm straight
penetration
1.1 m plug with      5.00x109    1.99x109      1.50x109       8.84x108     5.00x109    2.10x109
10 cm 4-bend
penetration
3.4 m plug with      9.64x104    7.04x103      1.97x104       1.63x103     1.59x105    1.33x104
10 cm 4-bend
penetration

5.9.10 Future Work                                   neutronics MCNP code. Significant
                                                     savings in the modeling effort can be
As FIRE proceeds beyond the current                  achieved if direct interface between the
pre-conceptual design phase, more                    CAD drawings and the MCNP geometry
sophisticated neutronics analysis will be            module is developed. Development of
needed. Detailed three-dimensional                   codes that facilitate such an interface
modeling of the whole machine that                   will be very valuable and will facilitate
includes material heterogeneity,                     performing 3-D neutronics calculations
penetrations and gaps is required to the             with the ability to quickly accommodate
level of detail used during the ITER                 design changes. Activities developing
engineering design activities. The model             these interface codes were initiated at the
should also account for accurate plasma              University of Wisconsin during the pre-
shape and source profile. The machine                conceptual design phase of FIRE and
3-D model will be the basis for                      need to expand as FIRE design proceeds
performing detailed 3-D neutronics                   to the conceptual design phase.
calculations to determine the different
neutronics parameters (e.g., nuclear                 The 2-D analysis performed for the diagnostics
heating, radiation damage, insulator                 port plug J showed that streaming is be an
dose, etc.)                                          important issue even with the relatively small
                                                     10 cm penetration. 3-D analysis is essential to
3-D modeling of ITER required a large                estimate the nuclear environment and radiation
amount of manpower because of the                    damage to the sensitive diagnostic system
need to manually generate the                        components (mirrors, fiber optics, critical
parameters for the surfaces and elements             electronics, etc.) in the diagnostic ports of
in the geometrical input for the                     FIRE. The impact of streaming through the

                                            5.9-23
                                FIRE Engineering Report
                                     August 2003
diagnostics penetrations will have also to be       [4] Technical Basis for the ITER Final
evaluated due to its impact on accessibility             Design Report, Cost Review and
for maintenance behind the cryostat.                     Safety Analysis, ITER EDA
                                                         Documentation       Series,    IAEA,
The divertor region has a geometrically                  Vienna, December 1997.
complex configuration that mandates 3-D             [5] S. Fabritsiev, S. Zinkle, and B.
calculations. Important issues that need to be           Singh, “Evaluation of Copper
addressed for that region are the amount of              Alloys for Fusion Divertor and First
cryopump heating, nuclear heating and                    Wall Components,” Journal of
radiation damage in the divertor components,             Nuclear Materials, 233-237, 127-
and neutron streaming through the divertor               137 (1996).
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                                                         “Copper Alloys for High Heat Flux
The magnet insulator in FIRE is                          Structure Applications,” Atomic
expected to experience significantly                     and Plasma-Material Interaction
higher dose than that in ITER. That puts                 Data for Fusion, Supplement to
premier on the need for insulator R&D                    Nuclear Fusion, 5 (1994) 163-192.
program to qualify insulators for the               [7] D.J. Alexander, S.J. Zinkle and A.F.
high dose level expected in FIRE. In                     Rowcliffe, “Fracture Toughness of
addition, we need to investigate the                     Copper-Base Alloys for Fusion
possibility of using better shielding                    Energy Applications,” J. Nucl.
material (e.g., W, WC, B4C) to reduce                    Mater., 271&272 (1999) 429-434.
the insulator dose particularly if                  [8] S.J. Zinkle and W.S. Eatherly,
extended FIRE operation is feasible.                     “Tensile and Electrical Properties of
                                                         Unirradiated and Irradiated Hycon
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