ACCIDENT DOSES ANALYSIS OF THE SOMBRERO

# INERTIAL FUSION ENERGY POWER PLANT DESIGN

S. Reyes^a,b^, J. F. Latkowski^a^, J. Gomez del Rio^b^,and J. Sanz^b^

^a^ Lawrence Livermore National Laboratory, P. O. Box 808, L-632,

Livermore, CA 94550, Tel: 925-424-5684

^b^ UNED/ Dpto. Ingeniería Energética & Instituto Fusión Nuclear

E.T.S. Ingenieros Industriales, C/Ciudad Universitaria s/n, 28040 Madrid
(Spain)

ABSTRACT

SOMBRERO (solid moving breeder reactor) is a conceptual design of a 1000
MW~e~ laser-driven inertial fusion energy (IFE) power plant. An
important goal of the original study was the achievement of a safe and
environmentally attractive reactor of relatively simple design. However,
recent work has pointed out some key issues involving safety that were
not completely addressed at that time, and which need to be reviewed in
order to maximize the SOMBRERO design attractiveness.

The present work uses a set of computer codes traditionally used for
magnetic fusion safety studies (CHEMCON, MELCOR), which have been
adopted and adapted for use in IFE safety analysis. Here we consider a
loss of flow accident (LOFA) combined with a simultaneous loss of vacuum
accident (LOVA) produced by a breach in the confinement building.
Although confinement failure would be a very unlikely event, it must be
postulated in order to produce significant off-site doses. The CHEMCON
code is used to simulate the long-term thermal transient in the reactor
structures resulting from oxidation and radioactive decay heat. MELCOR
is used to simulate a wide range of physical phenomena including
thermal-hydraulics, heat transfer, aerosol physics and fusion product
release and transport. As specified in the DOE Fusion Safety Standards,
an off-site dose below 1 rem (10 mSv) is the requirement to avoid public
sheltering and evacuation. The SOMBRERO accident analysis results will
be evaluated according to this limit and suggestions will be made for
improvements and future work.

I. INTRODUCTION

Safety and environmental issues in the SOMBRERO IFE power plant design
have been given a strong emphasis since the original report.^1^ The
SOMBRERO concept uses a low activation material (C/C composite) in the
chamber structures. The blanket consists of a moving bed of solid Li~2~O
particles flowing through the chamber, eliminating the risk energetic
chemical reactions usually associated with the use of lithium. The
chamber is surrounded by a 1.7 m thick shield wall at a radius of 10 m
with an additional 1 m thick confinement building wall at a radius of 55
m.

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Since the time of the original study, the nuclear community (fission and
fusion) has reached a consensus and established more restrictive safety
goals. The DOE Fusion Safety Standards for public protection in case of
accident give a limit off-site dose of 1 rem (10 mSv).^2^ This value is
based upon a 1991 Environmental Protection Agency document that
specifies the level at which sheltering and evacuation should be
undertaken.^3^ Also since that time, new experimental data has been
published on tritium retention in carbon/carbon composites,^4^ revealing
the need for an updated safety evaluation of the SOMBRERO design.

In the present work, we have adopted the principle of considering a
worst-case accident scenario. Here we assume a LOFA with loss of flow in
the four circuits of the primary coolant loop. In addition, a 1m^2^
break in the confinement building wall is assumed, leading to a LOVA
simultaneously with an air ingress event. Even though the probability of
such a severe accident would be extremely low we must adopt a
conservative methodology at this early stage, to later perform similar
analyses considering more frequent and less severe accidents.

II\. COMPUTER CODES AND METHODOLOGY

In order to calculate off-site doses in a particular scenario, first we
need to know the source term of radioactivity that is released to the
environment. Neutron transport and activation, heat transfer, and
thermal-hydraulics and aerosol transport calculations have been
performed. Given the activity release and the specific dose of each
isotope, the accident dose may be calculated.

Neutron calculations have been performed using TART^5^ Monte Carlo
transport code. Activation of components and decay heat results were
obtained from the ACAB^6^ activation code following 30 years of plant
operation.

Prior to the calculation of radioactivity release fractions, one needs
to know the temperature evolution of the different reactor components
during the accident. The CHEMCON^7^ code is used to simulate the
time-temperature histories of the structures during the transient,
resulting from oxidation and radioactive decay heat. Some modifications
have been introduced in the CHEMCON oxidation package for an enhanced
tracking of the oxidation front. The temperature excursion of the
different components is then used to determine the activation products
source term available for mobilization, which will later be used as
input data for the MELCOR calculations.^8^

MELCOR thermal-hydraulics code is used to simulate a wide range of
physical phenomena including heat transfer, aerosol physics and fusion
product release and transport. With these data about the radioactivity
released, and the adequate dose conversion factors (DCF), off-site doses
can be obtained. Data on DCF have been calculated using MACCS2 accident
consequences code.^9^

III\. TIME-TEMPERATURE HISTORY OF THE REACTOR COMPONENTS

In order to calculate the temperature excursions of the different plant
components one must account for the various energy sources existing
during the transient. Three possible sources (fusion reactions, decay
heat and oxidation heat) have been identified and will be discussed
next.

Once the accident begins, it is conceivable that fusion reactions could
continue to occur for a short period of time. During a LOFA accident
about 0.44% of the graphite of the first wall is going to evaporate,
increasing the carbon partial pressure of the reactor building by 1 torr
and preventing the laser beam from propagating to the target. This way,
no more fusion reactions would be possible and hence the reactor would
be shutdown. Thus, the two only energy sources present in this scenario
are the radioactive decay heat from activated materials and the
oxidation heat from the exothermic combustion of carbon in presence of
air. The decay heat is low enough to allow a rapid cooling of the first
wall (FW) and blanket structures (the temperature of the FW drops below
1000 ºC in less than a minute). Fig. 1 shows the temperature evolution
of the different components during the transient, due to the radioactive
decay heat.

![](media/image1.wmf){width="3.109722222222222in"
height="2.6118055555555557in"}

Figure 1. Time-temperature history of the various reactor components due
to the radioactive decay heat.

![](media/image2.wmf){width="3.120138888888889in"
height="2.7041666666666666in"}To determine the importance of the
oxidation of the carbon at those temperatures, we developed a CHEMCON
model for SOMBRERO.^1^ The model included the carbon chamber structures,
the Li~2~O blanket, inner shielding and confinement building. A series
of iterations were made between the CHEMCON oxidation calculations and
the MELCOR thermal-hydraulics code, to account for the oxygen diffusion
through a time-dependant carbon-monoxide source (the current version of
MELCOR does not account for oxidation of structures). It was found that
even at temperatures below 1000 ºC the reaction rate was significant
(the 1 cm thick first wall was oxidized in about 7 hours). Fig. 2 shows
the temperature evolution of the various components when accounting for
the air oxidation of the carbon structures.

Figure 2. Temperature evolution in the various reactor components due to
oxidation and radioactive decay heat.

It must be noted that the correlations used by CHEMCON are based on
experimental data of graphite oxidation with air dating more than 10
years ago. However, these results are meaningful given that the
oxidation of the chamber structures would have a great impact on the
radioactive mobilized fractions. All the carbon with its impurities and
activation products would be mobilized in CO form, as would the tritium
trapped in these structures. The activated Li~2~O granules would also be
available for mobilization, but given the size of the particles this
would be a minor contribution to the final released radioactivity
fraction.

In order to avoid the oxidation of the SOMBRERO carbon structures in an
air ingress event, several solutions will be discussed next. First, a
passive safety feature such as an inert gas filled tank with a rupture
disk, should be simple enough to implement. This would allow the gas to
fill up the building when a certain differential pressure is reached,
preventing the oxygen to reach the target chamber. Also, different
oxidation experiments have been done with protective coatings on
carbonaceous composite materials, such as Si-B-C coatings.^10^ Finally,
alternative materials other than C/C composites should also be
considered (some discussion has been already made about the usage of SiC
in the blanket structures, keeping the high-conductivity C/C composite
as first wall material).

IV\. ACTIVATION PRODUCTS SOURCE TERM

Assuming that the carbon structures of the FW/blanket assembly are going
to suffer oxidation during the accident, there will be four main sources
of radioactivity that must be considered in this scenario. First, all
the carbon material will be mobilized as a result of the oxidation.
Conservatively we assume that the impurities and activation products
present in the carbon structures will be also mobilized, dragged along
by the CO gas and finally released through the breach in the building
wall.

Second and as a consequence of the oxidation of the graphite, the
tritium trapped in these structures will be released. In the original
report it was estimated that only 10 grams of tritium would be retained
in the graphite reactor structure. Recent data about tritium trapping in
C/C composites have shown that this number was underestimated. We have
adopted the more realistic value of 1 kg of tritium for the total of
FW/blanket graphite shells. This mass, together with the steady state
inventory of the Li~2~O (162 g), the helium carrier gas (5 g), the
reactor Xe gas (4.6 g) and the target feed system (1 g) results in a
total of 1172.6 g of tritium that will be mobilized during the
transient. We conservatively assume that the tritium is released in the
more radiotoxic form of HTO.

Third, we assume that fraction of the Li~2~O inventory present in the
chamber in the moment of the accident (1/3 of the total 2000 tonnes of
Li~2~O) will be also mobilized as a consequence of the combustion of its
structural support. Given the size of the granules (300 to 500 μm in
diameter) most of them will deposit on the reactor horizontal
structures, resulting in a minor release fraction.

Finally, the Xe gas that fills the reactor will also be mobilized and
finally released through the break in the confinement.

It must be noted that if oxidation could be prevented, then only 172.6 g
of tritium and the atmosphere of Xe gas would constitute the radioactive
mobilized sources.

V. RADIOACTIVITY RELEASES AND OFF-SITE DOSES

In order to estimate release fractions, a MELCOR model of SOMBRERO was
developed. This model includes the FW/blanket assembly, the inner
shielding and the confinement building. The radioactive source term
described in the previous section is used as input data for the MELCOR
calculations . We have considered a severe accident consisting of a
LOFA, with loss of flow in all of the four cooling loops, with
simultaneous LOVA produced by failure of the confinement building and
consequent air ingress. This model is used by the code to simulate the
progression of the accident.

MELCOR heat transfer package considers conduction, convection and
radiation between the structures. The aerosol transport module treats
the aerosol nucleation and agglomeration, vapor condensation, gravity
settling and gaseous/liquid transport. A new module introduced by INEEL
allows simulation of HTO transport and condensation.

A. Results With Graphite Oxidation

Assuming that the graphite reactor structures suffer oxidation, we
calculated the source term mobilization and release fractions to the
environment, and the final doses. We have considered three options for
the chamber filling gas. In the first case we use Xe, including all its
activation products. Second, we report results for the case that the
iodine and cesium isotopes could be removed from the Xe gas by the
chamber vacuum system. Finally, we show the results for the case where
Kr is used instead. It can be noticed that the gaseous species (graphite
in CO form and the chamber filling gas) result in a release fraction of
100%. Due to the size of the Li~2~O granules, most of them will deposit
on the reactor horizontal structures, resulting in a 1 kg release from
the initial 667 tonnes mobilized. Because of the partial condensation of
the HTO on the colder building walls, we get a release fraction of 19%
from the initial 1.173 kg mobilized. With the data about the activity
released and the adequate DCF, off-site doses can be obtained. The DCF
library has been updated to include radionuclides that were missing from
traditional (fission) libraries as some of these radionuclides maybe
important for fusion. Early doses were calculated using typical weather
conditions and ground level release as recommended by the Fusion Safety
Standards.

Results in Table 1 show that the dominant dose comes from the Xe gas,
giving a result of 4.69 rem. The graphite from the FW/blanket structures
results in a dose of 0.17 rem, and the Li~2~O makes a negligible
contribution to the final dose. The design using Xe would lead to a
final accident dose of 5.64 rem if the non-xenon activation products
were included in the release. This value would be reduced to 1.19 rem
assuming that the iodine and cesium isotopes can be separated by the
chamber vacuum system. For a modified SOMBRERO using Kr instead of Xe,
we calculate a total accident dose of 1.06 rem.

Table 1. Mobilized activities, release fractions and off-site doses,
case of graphite oxidation.

+------------------+-------------------+--------------+--------------+
| Radioactive      | Mobilized         | Release      | Off-site     |
| source           | activity          | fraction (%) | dose (rem)   |
|                  |                   |              |              |
|                  | (Bq)              |              |              |
+------------------+-------------------+--------------+--------------+
| C-FW             | 6.1E+14           | 100          | 0.05         |
+------------------+-------------------+--------------+--------------+
| C_blanket        | 1.4E+15           | 100          | 0.12         |
+------------------+-------------------+--------------+--------------+
| Li2O             | 5.8E+18           | 1.5E-4       | 0.00         |
+------------------+-------------------+--------------+--------------+
| HTO              | 4.2E+17           | 19           | 0.78         |
+------------------+-------------------+--------------+--------------+
| Xe               | 1.4E+16           | 100          | 4.69         |
+------------------+-------------------+--------------+--------------+
| Xe^\*^           | 1.4E+16           | 100          | 0.24         |
+------------------+-------------------+--------------+--------------+
| Kr               | 8.8E+15           | 100          | 0.11         |
+------------------+-------------------+--------------+--------------+
| Total w/Xe       | 6.2E+18           |              | 5.64         |
+------------------+-------------------+--------------+--------------+
| Total w/Xe^\*^   | 6.2E+18           |              | 1.19         |
+------------------+-------------------+--------------+--------------+
| Total w/Kr       | 6.2E+18           |              | 1.06         |
+------------------+-------------------+--------------+--------------+

Xe\* = Xe without iodine and cesium activation products

Given that our goal is an accident dose below the 1 rem limit given by
the DOE Standards, other design modifications have been considered.
Other than the activation products from the Xe gas in the chamber, the
most contributing source to the final dose is the tritium. The quantity
of tritium that condenses on the confinement building walls will depend
on the time-temperature history of those structures. In order to keep
the temperature of the building inner surface as low as possible, a high
conductivity material could be useful. The SOMBRERO report used concrete
with 2.8% vol. mild steel as the building material. We have studied
different alternatives, and here we propose two options. First,
increasing the steel content to 5% would reduce the release fraction of
the tritium to 14% and the dose to 0.59 rem. This would result in a
final dose of 1 rem in the case of Xe with removal of iodine and cesium
isotopes, and 0.87 rem in the case of Kr used as chamber gas. Another
alternative would be replacing the material with 97% concrete and 3%
aluminum. This would also affect the thermal conductivity of the
building and thus, the condensation of tritium on the walls. The tritium
release fraction would be reduced to 11% and the off-site dose to 0.4
rem. This would result in a total accident dose of 0.81 rem in the case
of Xe with the iodine and cesium isotopes being removed, and 0.68 rem if
the chamber gas is Kr.

B. Results Without Graphite Oxidation

Assuming that oxidation of graphite does not occur, the only radioactive
source terms would be the 172.6 g of tritium trapped in other parts than
the graphite chamber, and the reactor filling gas (Xe or Kr). The
results in this case are shown in Table 2.

Table 2. Mobilized activities, release fractions and off-site doses,
case without graphite oxidation.

  -------------------- ----------------- ----------------- ---------------
  Radioactive source   Mobilized         Release fraction  Off-site dose
                       activity (Bq)     (%)               (rem)

  HTO                  6.2E+16           2.8               0.02

  Xe                   1.4E+16           100               4.69

  Xe^\*^               1.4E+16           100               0.24

  Kr                   8.8E+15           100               0.11

  Total w/Xe           7.6E+16                             4.71

  Total w/Xe^\*^       7.6E+16                             0.26

  Total w/Kr           7.0E+16                             0.13
  -------------------- ----------------- ----------------- ---------------

Xe\* = Xe without iodine and cesium activation products

In this case the tritium dose is only 0.02 rem, but the Xe would still
produce an off-site dose over 4 rem if the iodine and cesium activation
products were included in the release. If these isotopes were removed by
the chamber vacuum system, the final dose would be 0.26 rem. For the
modified design using Kr, we get an accident dose of 0.13 rem. In these
last two cases with final dose 1 rem, an evacuation plan would not be
needed.

VI\. CONCLUSIONS AND FUTURE WORK

In the present work, accident doses have been analyzed for the SOMBRERO
IFE design. A severe accident consisting on a total LOFA with
simultaneous LOVA because of failure of the confinement building and
subsequent air ingress event has been simulated using fusion safety
codes.

During the accident, graphite structures will suffer oxidation with air,
resulting in mobilization of the carbon, including the tritium trapped
in it, and the Li~2~O granules from the blanket. The oxidation of
graphite with air is a major issue that should be prevented by
introducing passive safety features, protective coatings, or alternate
FW and/or blanket materials.

Assuming that oxidation takes place, the calculated off-site dose
results in 5.64 rem. Removing the iodine and cesium activation products
from the Xe gas would result in a dose of 1.19 rem. If Kr is used as
alternate gas for the reactor atmosphere this value would be reduced to
1.06 rem. If we assume that the iodine and cesium can be removed from
the Xe gas, or that Kr is used instead, doses below 1 rem could be
achieved with simple design modifications such as increasing the steel
content in the confinement building (to 5% vol.) or adding aluminum (3%
vol.) to the concrete. The higher thermal conductivity of the material
would allow a less severe temperature excursion of the building walls
and more tritium could condense on the confinement inner surface. If
oxidation could be prevented, then the off-site dose would be 4.71 rem
if the non-xenon activation products were included in the release, and
0.26 rem if the iodine and cesium isotopes were removed by the chamber
vacuum system. In the case were Kr is used instead of Xe, the final dose
results to be 0.13 rem.

The cases with total dose below 1 rem would meet the requirement given
by the DOE Fusion Safety Standards in order to avoid public sheltering
and evacuation.

Additional work is needed in the areas of C/C composites oxidation and
tritium trapping and retention. Also the use of steam in the He carrier
gas may reduce the tritium inventory but still needs to be evaluated. We
plan to identify and simulate other more credible, less conservative
accident scenarios in order to make a complete safety analysis for the
SOMBRERO IFE power plant design.

## ACKNOWLEDGEMENTS

The authors would like to thank their colleagues at the INEEL and U.
Wisconsin-Madison for their assistance and recommendations that have
been essential to the completion of the above work.

Work performed under the auspices of the U. S. Dept. of Energy by
University of California Lawrence Livermore National Laboratory under
Contract W-7405-Eng-48 and under support Ministerio de Educación y
Cultura, Spain, under Plan Nacional I+D+I (2000-2003) Programa Nacional
de Fusion Termonuclear, and the European Union keep-in-touch Program on
IFE.

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