PROGRESS IN ACCIDENT ANALYSIS OF THE HYLIFE-II

# 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

The present work continues our effort to perform an integrated safety
analysis for the HYLIFE-II inertial fusion energy (IFE) power plant
design. Recently we developed a base case for a severe accident scenario
in order to calculate accident doses for HYLIFE-II. It consisted of a
total loss of coolant accident (LOCA) in which all the liquid flibe
(Li~2~BeF~4~) was lost at the beginning of the transient. Results showed
that the off-site dose was below the limit given by the DOE Fusion
Safety Standards for public protection in case of accident, and that
this dose was dominated by the tritium released during the accident.

In order to further advance a complete safety analysis for HYLIFE-II, a
range of other accident scenarios must be considered. In this work, we
introduce a new version of the MELCOR thermal-hydraulics code recently
developed by the Idaho National Engineering and Environmental Laboratory
(INEEL) that uses flibe as the working fluid. We have focused on a loss
of flow accident (LOFA), with simultaneous failure of the blanket
structure and the beam tubes that connect the chamber with the outside
of the confinement building. This constitutes the bypass needed to
communicate the target chamber with the environment. Once the release
fractions of the various radioactivity sources are known, we calculate
off-site doses under different conditions as a consequence of the
accident.

I. INTRODUCTION

The HYLIFE-II power plant concept^1^ is based on non-flammable,
renewable liquid wall target chambers formed with flibe molten salt
jets, a heavy-ion driver, and double-sided illumination of indirect
drive targets. The liquid wall serves to attenuate fast neutrons before
they strike the first structural wall (FSW), lengthening the lifetime of
components, and also serves to breed tritium fuel. Flibe shielding
circuits inside the target chamber include oscillating flow and steady
flow. Two more flibe circuits are contained in the blanket assembly
procuring cooling and shielding to these structures.

Recently, we modeled a severe accident analysis for HYLIFE-II in order
to calculate accident doses.^2^ It consisted of a total LOCA with
simultaneous breakage of 100% of the beam tubes and failure of both the
inner shielding and confinement buildings. The only flibe mass that was
available for mobilization in this case was the fraction vaporized by
the x-rays from the last ignited target, given that all the liquid flibe
was assumed to be lost. This vapor was then treated as an aerosol in the
thermal-hydraulics calculations in order to obtain the accidental
release fraction. Results showed that the off-site dose to the most
exposed individual (MEI) was below the 1 rem limit given by the DOE
Fusion Safety Standards^3^ for public protection in case of accident.

Here, we consider a range of other accident scenarios, using a new
version of the MELCOR thermal-hydraulics code^4^ recently developed by
INEEL, which introduces flibe as the working fluid.^5^ During a loss of
flow accident, which could be initiated by failure of the pumping
system, the flibe present inside the target chamber (the fraction
vaporized during the last shot plus the flibe from the protective liquid
wall) will be available for mobilization and release. Further, we
estimate the flibe release fraction in the case that there was a failure
in the blanket structure. In this case the entire flibe inventory
present in the first wall/blanket assembly in the moment of the accident
would also be available for mobilization. In order to obtain a
radioactivity fraction released to the environment, we postulate a break
in the beam tubes outside of the reactor building. This constitutes the
bypass needed to communicate the target chamber with the environment.
With the results on the radioactivity release fractions, and the
adequate dose conversion factors (DCF), the accident dose may be
calculated.

II\. COMPUTER CODES AND METHODOLOGY

In order to perform accident doses analysis, first one needs to know the
fraction of the radioactivity source term that is released to the
environment. Neutron transport and activation, heat transfer, and
thermal-hydraulics and aerosol transport calculations have been
performed.

TART^6^ Monte Carlo transport code has been used for neutron transport
calculations. Activation of components and decay heat results were
obtained from the ACAB^7^ activation code following 30 years of plant
operation.

CHEMCON^8^ heat transfer code is used to simulate the long-term
time-temperature histories of the structures during the transient,
resulting from the decay heat of activated structures. 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 for the MELCOR calculations.

MELCOR thermal-hydraulics code is capable of simulating a wide range of
physical phenomena, which include heat transfer, aerosol physics and
fusion product release and transport. The original version of the code,
traditionally used by the fission community, has been adapted for fusion
accident analysis by INEEL. This work uses a new version of the code
that substitutes flibe for water as the working fluid. Once
radioactivity release fractions are known, we calculate the adequate DCF
in order to obtain off-site doses under different accident conditions.
Data on DCF have been calculated using MACCS2 accident consequences
code.^9^

III\. TIME-TEMPERATURE HISTORY OF THE REACTOR COMPONENTS

During a thermal transient in the plant, the potential energy sources
must be identified in order to determine the temperature excursions of
the different components. In this case the only energy source is the
radioactive decay heat from the activated material in the reactor
structures. Once the accident begins it is conceivable that fusion
reactions could continue to occur for a short period of time. Several
factors must be considered: accelerator operation, target injection, and
beam propagation through a varying environment. If accelerator operation
and target injection both continue, then beam propagation will determine
when ignition is no longer possible. We have assumed that the beam tubes
failure occurs outside the confinement building, allowing the air to
penetrate the chamber, and resulting in the inability to propagate the
heavy ion beams to the target.

![](media/image1.wmf){width="2.9430555555555555in"
height="2.5506944444444444in"}Considering then that the decay heat from
activated materials is the only energy source during the transient, a
CHEMCON calculation was developed using a simple HYLIFE-II 1-D
cylindrical model. This model consists of four stainless steel shells
that represent the FSW/blanket structure. It also includes the flibe
trapped in these structures during the LOFA, an inner concrete shield
and the confinement building.^9^ Fig. 1 shows the time-temperature
history of the different components during the transient.

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

Due to the low radioactive afterheat, the first wall temperature,
initially at 675 C, experiences an insignificant peak of tenths of a
degree at about 10 hours, after which it gradually falls. This indicates
that the stainless steel temperature will remain in the surroundings of
the nominal working temperature during the transient. This result
eliminates the possibility of volatilization of activated stainless
steel or any over-stress damage during the accident.

IV\. ACTIVATION PRODUCTS SOURCE TERM

There are three main radioactivity sources to be considered in
HYLIFE-II. First is the flibe with its activation products. When we
studied the total LOCA the only flibe mass to be considered were the 10
kg vaporized by the x-rays from the last ignited target. However, in
this case and as a consequence of the loss of flow, the protective
liquid wall in front of the FSW would also be standing as a pool inside
the target chamber. If additionally we assume a break in the blanket
piping, then the flibe from the blanket cooling and shielding circuits
would leak inside the chamber. That would result in a total mass of
about 140 tonnes of flibe available for mobilization in this case.

Second, we assume that the corrosion of type 304 stainless steel (SS304)
by flibe within the chamber and blanket can be limited to less than 10
m/y via corrosion control methods.^10^ Additionally, we assume that the
flibe clean-up system can maintain the mobilizable mass of corrosion
products to a 1 m-y inventory. Given a total surface area of 1040 m^2^,
we obtain a corrosion product inventory of 8.3 kg in the total flibe
volume. Also, using data from oxidation-driven mobilization experiments
on PCA performed at INEEL, we calculate an additional 0.5 kg of
stainless steel that would be oxidized at the accident temperatures (we
assume that SS304 mobilization will be the same as that from
PCA).^11,12^ Adding up the corrosion and oxidation products, we obtain
10 kg of SS304 in the total flibe volume. Considering that only 5% of
the flibe inventory is present in the chamber at any given time, we
obtain a total of 0.5 kg of SS304 mobilized during the transient. It is
worth noting that we are being quite conservative in assuming that the
corrosion products will be in an aerosol form and available for
mobilization (one could make an argument that the corrosion products
would leave with the flibe during the LOCA).

Finally, it is estimated that approximately 140 g of tritium would be
trapped within the chamber, blanket and piping. We conservatively assume
that the entire tritium inventory is converted to the more radiotoxic
HTO form, yielding a total mass of 1 kg of HTO.

V. RADIOACTIVITY RELEASE AND OFF-SITE DOSES

In order to estimate release fractions, a MELCOR model of HYLIFE-II was
developed. This model consists of the FSW and blanket structures, beam
tubes (considering an updated HYLIFE-II version which includes 96 beams
per side), inner shield and confinement building. The heavy-ion beam
tubes and the target chamber are assumed to be at vacuum. The
radioactive source term described in the previous section is supposed to
be available for mobilization during the transient. This version of the
MELCOR code uses flibe as the operating fluid. 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.  Flibe Release and Off-site Doses

We have performed a series of MELCOR calculations in order to
characterize the behavior of the flibe in accidental conditions not
involving the total loss of the liquid coolant previously addressed. We
assume a LOFA accident with failure in the blanket structure and a break
in the beam tubes that would constitute a bypass between the target
chamber and the environment. We have assumed different severities of the
tubes breakage, ranging from the total failure of the 192 beam tubes to
the case were a single tube breaks. The occurrence of blanket flibe
leakage directly to the beam tubes was also studied. Given that a
special region of concrete shielding for the focusing magnets isolates
the beam tubes from the blanket, we discarded this possibility. We have
obtained the released mass in each scenario and then considered three
sets of accident conditions for off-site dose calculations purposes.
First set is called Typical Case with Ground Release (TCGR) and assumes
average weather conditions (wind speed = 4 m/s, atmospheric stability
class D), and radioactivity release at ground level. Second, the Typical
Case with Elevated Release (TCER), also considers average weather
conditions but the release is produced trough a smokestack 100 m above
ground level (TCHR). Finally the Bounding Case with Ground Release
(BCGR) considers conservative weather conditions (wind speed = 1 m/s,
atmospheric stability class F) and ground level release. Table 1 shows
the results for mass released and off-site doses in the different
described scenarios.

Table 1. Released flibe mass (kg) and off-site doses (mrem)

+-------------------+----------+------------+------------+------------+
| Beam tubes        | Mass     | Dose       | Dose       | Dose       |
|                   |          |            |            |            |
| failure fraction  | released | TCGR       | TCER       | BCGR       |
+-------------------+----------+------------+------------+------------+
| 100%              | 5.54     | 1.24       | 0.23       | 16.6       |
+-------------------+----------+------------+------------+------------+
| 50%               | 4.56     | 1.02       | 0.19       | 13.7       |
+-------------------+----------+------------+------------+------------+
| 25%               | 2.63     | 0.59       | 0.11       | 7.89       |
+-------------------+----------+------------+------------+------------+
| 1 beam tube       | 0.13     | 0.03       | 0.005      | 0.39       |
+-------------------+----------+------------+------------+------------+

Given the low afterheat and the temperature decrease of structures
during the transient, the liquid flibe is not going to suffer
volatilization and will mostly remain in the form a pool in the
chamber/blanket area. The 10 kg of vaporized flibe generated by the last
target, will immediately be converted into fog at the beginning of the
accident, and then transported from the chamber to the beam tubes,
allowing some of this fog to deposit on structures before reaching the
environment. In the first three cases (100%, 50% and 25% of beam tubes
failing), the mass released turns out to be larger that that obtained in
the LOCA study, where we only accounted for the 10 kg of mobilized vapor
and obtained a release fraction of 11%. This is because then, we
considered that the beam tubes would fail inside the inner shield
building, having the mobilized material to travel from here to the
confinement building and next to the environment through the postulated
breaks in the walls. This scenario allowed larger surface areas so as a
longer travel time for the flibe to deposit before leaking through the
breach in the confinement building wall.

Results show that even in the case of 100% beam tubes failure, the dose
in the most pessimistic conditions is less than 17 mrem. It must be
noticed that the isotope ^18^F contributes in a 99% to the final dose.

Other than activation concerns, flibe has some chemical and toxicity
issues. First and foremost, the escaping beryllium compounds, notably
BeF~2~, pose a health hazard. The recommended limit is 25 g/m^3^ for
peak exposure to beryllium and its compounds.^13^ If we assume a release
of 1 kg of flibe (containing 340 g of BeF~2~) that escaped to the
environment under conservative weather conditions over a period of 1
hour, the concentration at a 1 km site boundary would only temporarily
reach 3 g/m^3^.^14^ However, the beryllium concentration within the
plant could require worker protection. Another concern relative to the
use of flibe in HYLIFE-II is the potential of HF formation. When the
lithium is fissioned by neutron interaction, it leaves a free fluorine
atom to combine with available hydrogen. HF poses a corrosion threat and
has its own health risks. One method that has been proposed to control
flibe pH is the presence of beryllium flowing pebbles or fingers which
would allow the HF or free fluoride ions to react back to BeF~2~.^15^

B.  SS304 Release and Off-site Doses

With the previously described MELCOR model we also obtained the release
fractions of SS304 corrosion and oxidation products in the various
scenarios. An initial mass of 0.5 kg was used for the aerosol source
term and the aerosol nucleation and agglomeration, gravity settling and
transport during the accident were simulated. We also calculated the
correspondent DCF for the different accident conditions and results are
shown in table 2.

Table 2. Released SS304 mass (kg) and off-site doses (mrem)

+-----------------+----------+------------+------------+------------+
| Beam tubes      | Mass     | Dose       | Dose       | Dose       |
|                 |          |            |            |            |
| Failure         | released | TCGR       | TCER       | BCGR       |
| fraction        |          |            |            |            |
+-----------------+----------+------------+------------+------------+
| 100%            | 0.27     | 1.40       | 0.13       | 20.9       |
+-----------------+----------+------------+------------+------------+
| 50%             | 0.22     | 1.13       | 0.11       | 16.8       |
+-----------------+----------+------------+------------+------------+
| 25%             | 0.12     | 0.64       | 0.06       | 9.60       |
+-----------------+----------+------------+------------+------------+
| 1 beam tube     | 0.009    | 0.05       | 0.004      | 0.68       |
+-----------------+----------+------------+------------+------------+

It can be observed that in the most pessimistic case (100% of the beam
tubes fail and conservative weather conditions with ground release) the
accident dose to the public would be less than 21 mrem. In this case
about 75% of the total dose comes from the radioisotope ^60^Co, being
followed in importance by the isotopes ^58^Co and ^59^Fe. It must be
pointed out that in the first three beam tubes failure cases, the
radioactivity release results to be larger than that obtained in the
LOCA analysis (only 0.07 kg of SS304 were released then). This is due to
the effect of the smaller available surface for aerosol deposition and
shorter travel time to the environment in this particular bypass
scenario.

C.  HTO Release and Off-site Doses

In order to simulate the HTO transport and condensation on structures we
made use of a former version of the MELCOR code which uses water as the
working fluid in stead of flibe (at this point no-multifluid version of
the code is available yet). An instantaneous release of the tritium from
the SS304 structures was assumed. Calculations of tritium migration from
the stainless steel shells were also developed using TMAP (Tritium
Migration Analysis Program). These calculations showed that at the
operating temperature of 675 ºC, the tritium migration from the steel is
fast enough (\>90% in only 1.5 hour) that there is no difference when
comparing with the results from the instantaneous migration
assumption.^16^ Results show that in this case 100% of the tritium is
released. The non-existence of a cold structure (like the confinement
building wall in the case of the LOCA analysis) where this tritium would
condense and avoid being released, leads to the result of a 1 kg of HTO
release to the environment.

The consequent off-site dose from the tritium in case of average weather
conditions with ground level release would be 0.5 rem. The dose would
result in only 47 mrem if the release occurred through an elevated
stack. If pessimistic weather conditions were assumed instead, then the
accident dose would be 6.4 rem. As occurred in the case of the LOCA
accident, the tritium dominates the total result, being the
contributions of the flibe and SS304 less than 1% of the final accident
dose.

VI\. CONCLUSIONS

Recently, we developed an accident analysis for HYLIFE-II consisting on
a total LOCA. As a consequence of the loss of coolant, no flibe was
involved in the accident except for the fraction vaporized by the x-rays
from the last reactor shot.

In this work we use a new version of the MELCOR thermal-hydraulics code
that substitutes flibe for water as the working fluid. We have developed
an analysis for a LOFA accident, including failure of the blanket
structure and partial/total breakage of the beam tubes that communicate
the chamber with the environment.

Results show that the radioactive afterheat from activated components is
low enough to allow cooling of structures during the transient. Even
though the entire flibe inventory present in the chamber and blanket at
the moment of the accident is available for mobilization, the release
fraction is small enough to result in off-site doses lower than 20 mrem,
demonstrating the inherent radiological safety of HYLIFE-II relative to
the use of flibe. This flibe together with the stainless steel corrosion
and oxidation products contribute to the final off-site dose with less
than the 1% of the total. The final accident dose is clearly dominated
by the 140 g of tritium trapped in the stainless steel structures and
released during the accident.

For the most pessimistic scenario, where a failure of all of the beam
tubes occurs, we obtain an accident dose of 0.5 rem. This dose considers
average weather conditions (as recommended by DOE Fusion Safety
Standards) and assumes ground level release. This result is below the 1
rem limit for public protection in case of accident, and means that an
evacuation plan would not be needed in this case.

In order to maximize the potential of IFE as a safe and environmentally
attractive energy source, more effort is needed in this area. Future
work will include identification of accident initiators and sequences in
order to analyze other accident scenarios, so as work in alternative IFE
concepts such as dry-wall, laser-driven designs.

## ACKNOWLEDGEMENTS

The authors would like to thank their colleagues at the INEEL 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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## 

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