Study of experimental core configuration of the modified STACY for measurement of criticality characteristics of fuel debris

Progress in Nuclear Energy 101 (2017) 321e328

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Progress in Nuclear Energy
journal homepage: www.elsevier.com/locate/pnucene

Study of experimental core configuration of the modified STACY for
measurement of criticality characteristics of fuel debris
Satoshi Gunji a, *, Kotaro Tonoike a, Kazuhiko Izawa b, Hiroki Sono b
a
b

Japan Atomic Energy Agency, Nuclear Safety Research Center, Shirakata 2-4, Tokai-mura, Ibaraki, Japan
Japan Atomic Energy Agency, Department of Fukushima Technology, Shirakata 2-4, Tokai-mura, Ibaraki, Japan

a r t i c l e i n f o

a b s t r a c t

Article history:
Received 26 August 2016
Received in revised form
10 February 2017
Accepted 1 March 2017
Available online 13 April 2017

Criticality safety of fuel debris, particularly MCCI(Molten-Core-Concrete-Interaction) products, is one of
the major safety issues for decommissioning of Fukushima Daiichi Nuclear Power Station. Criticality or
subcriticality condition of the fuel debris is still uncertain; its composition, location, neutron moderation,
etc. are not yet confirmed. The effectiveness of neutron poison in cooling water is also uncertain for use

as a criticality control of fuel debris. A database of computational models is being built by Japan Atomic
Energy Agency (JAEA), covering a wide range of possible conditions of such composition, neutron
moderation, etc., to facilitate assessing criticality characteristics once fuel debris samples are taken and
their conditions are known. The computational models also include uncertainties which are to be clarified by critical experiments. These experiments are planned and will be conducted by JAEA with the
modified STACY(STAtic experiment Critical facilitY) and samples to simulate fuel debris compositions.
Each of the samples will be cladded by a zircalloy tube whose outer shape is compatible with the fuel rod
of STACY and loaded into an array of the fuel rods. This report introduces a study of experimental core
configurations to measure the reactivity worth of samples simulating MCCI products. Parameters to be
varied in the computation models for the experimental series are:
It is concluded that the measurement is feasible in both under- and over-moderated conditions.
Additionally, the required amount of samples was estimated.
© 2017 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND
license ( />
Keywords:
Fuel debris
MCCI product
Modified STACY
Fukushima Daiichi nuclear power station
Critical experiment

 Uranium dioxide with 235U enrichments of 3, 4, and 5 wt.%,
 Concrete volume fraction in the samples of 0, 20, 40, 60, and
80%, and
 Porosity of the samples filled from 0 to 80% where the sample
void is filled with water.

1. Introduction
Criticality safety is one of the major safety issues for defueling of
the damaged reactors in Fukushima Daiichi Nuclear Power Station
(1F-NPS) (Tonoike et al., 2013). Criticality control method of fuel

debris must be established in the mid- or long-term process of
defueling and decommissioning. A significant difference, from the
view point of criticality control, between situations in the 1F-NPS
reactors and the Three Mile Island Unit 2 reactor (TMI-2) is that

* Corresponding author.
E-mail address: (S. Gunji).

cooling water for fuel debris in the 1F-NPS reactors cannot be
poisoned continuously as was done in the TMI-2. Because the
cooling water is not flowing in a closed loop, the destination of
injected water is not known. Therefore, it is difficult to manage
concentrations of the poison in the cooling water for the purpose of
criticality control.
It would be necessary that a mitigation-based criticality control
method is adopted for decommissioning of 1F-NPS. For this purpose, it is necessary to get the criticality characteristics of fuel
debris. However, the actual fuel debris in the reactors has not yet
been observed and it is difficult to obtain accurate information on
its composition, location, neutron moderation, etc. (Tonoike et al.,
2015). This situation leads to large uncertainty in estimation of
criticality characteristics, and criticality or subcriticality condition
of the fuel debris. Therefore, a database of computational models
for possible criticality characteristics of the fuel debris is being built
which will help to predict in which condition critical events may
occur (Tonoike et al., 2015).
Most of criticality characteristics of fuel debris have not been

/>0149-1970/© 2017 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license ( />

322

S. Gunji et al. / Progress in Nuclear Energy 101 (2017) 321e328

evaluated before the accident of 1F-NPS, especially, that of moltencore-concrete-interaction (MCCI) product. Molten core might drop
from the pressure vessels of the reactors on the concrete floors in
the containment vessels, where MCCI products might be produced.
MCCI product has a small neutron absorption cross-section, and
may be porous and contain water when submerged. In fact, it has
been confirmed that the containment vessel floors are submerged
(Status of Fukushima Daiichi Nuclear Power Station (2015)). In past
studies, criticality characteristics of MCCI products had been evaluated only by computations. Some of them suggest small critical
mass and high boron concentration in cooling water that guarantee
subcriticality (Izawa et al., 2012). Detail study of criticality characteristics of MCCI products, therefore, should be conducted,
including criticality experiments, for establishment of criticality
control or criticality risk assessments.
Critical experiments are being planned to validate such highaccuracy computations to support criticality safety or criticality
risk evaluation of defueling in 1F-NPS that will change a volume
ratio of fuel debris and water. It will be conducted at the modified
Static Experiment Critical Facility (STACY) with samples simulating fuel debris compositions (Tonoike et al., 2015). In this paper,
experimental core configurations with samples of MCCI products
were considered based on recent knowledge of criticality characteristics of fuel debris. The amount of samples to be prepared
was determined as function of their reactivity worth and the differences of the core critical water heights. This is important
because there are limitations on the insertion reactivity and the
critical water height in particular specifications of the critical
facility.
2. Analysis and experimental conditions
2.1. Submerged MCCI product
The moderation condition of critical experiments should be
varied, however, more widely because hydrogen in the concrete
would contribute to neutron moderation and because the amount

of hydrogen in actual MCCI products is still unknown. A series of
analyses of criticality characteristics of MCCI products was shown
in Ref. 5, where infinite multiplication factors of MCCI products
with 235U enrichments of 3, 4, and 5 wt.% were computed in homogeneous and heterogeneous conditions. The results indicate that
optimum moderation conditions of those MCCI products would be
at the volume ratios of moderator to fuel (Vm/Vf) of 0.2e4. Vm/Vf is
expressed by following equation (1), this is specifically the modified STACY design based. This value means ratio of water volume to
fuel pellet volume in an active core. Therefore, it does not consider
water contents in samples.

.
Volume of moderator water
Vm V ¼
Volume of UO2
f

(1)

2.2. Outline of the modified STACY
The modification of the STACY is now underway at Japan Atomic
Energy Agency (JAEA) in order to accumulate fundamental experimental data relating to the criticality control for fuel debris
handling in 1F-NPS. The modified STACY is designed to be a tanktype light-water-moderated critical assembly, whose first criticality is expected in FY2018 (Sono et al., 2015; Miyoshi et al., 2015).
An overview of the modified STACY is shown in Fig. 1. Each fuel rod
will consist of UO2 pellets with a diameter of 8.2 mm and a 235U
enrichment of 5 wt.%, and a zircalloy cladding with an outer-

Fig. 1. Concept of modified STACY.

diameter of 9.5 mm. The stack height of the pellets will be
1420 mm.

The modified STACY will be operated by means of filling the tank
with water, which works as neutron moderator and reflector, from
the bottom of the core tank. Reactivity will be adjusted by adjusting
the water height. Vm/Vf of the core will be varied by changing fuel
rod interval. For example, Vm/Vf is 1.2 when a fuel rod interval in a
square lattice is 11.5 mm (Izawa et al., 2015; Sakon et al., 2015),
which was selected as the experimental core configuration in this
study.
2.3. Experimental core configurations
Two experimental core configurations were studied by using the
MCNP 5.1 code system (X-5 Monte Carlo Team, 2003; Brown et al.,
2009) and the nuclear data library JENDL-4.0 (Shibata et al., 2011).
In order to have statistical error of less than 0.015 %Dk (~2 cent1),
5 Â 107 effective neutron histories was used for each calculation.
The effective critical water heights of the two core configurations
were approximately 1000 mm.
One configuration was designed to have a hard neutron spectrum and an under-moderation configuration, which is shown in
Fig. 2. The configuration consists of 701 fuel rods arrayed in a
uniform square lattice with the interval of 11.5 mm. Its critical
water height is estimated to be 990 mm.
The other configuration has a soft neutron spectrum and an
over-moderation configuration. The configuration consists of 400
fuel rods in total and is divided into two regions as shown in Fig. 3.
The “driver region” is an array of fuel rods with the same interval of
11.5 mm that surrounds the “test region”. The “test region” consists
of 85 fuel rods arrayed more sparsely with 84 positions left vacant
(filled with water), an effective interval of 16.3 mm, and Vm/Vf of
about 3.7. Its critical water height is estimated to be 935 mm.
The neutron spectra at the center of each experimental core

1

The effective delayed neutron fraction was calculated by SRAC/TWODANT;

beff ¼ 0.0075.


S. Gunji et al. / Progress in Nuclear Energy 101 (2017) 321e328

323

Fig. 4. Neutron spectra at the center of each core configuration.

Fig. 2. “Under-moderation” experimental core configurations in square lattice of the
modified STACY.

array and a water height of approximately 1000 mm. To determine
the reactivity of the samples, each sample was modeled in the base
array and the reactivity worth was obtained from the change in keff.
Effective multiplication factors (keff) were 1.00035 ± 0.00011 and
1.00357 ± 0.00011, respectively, for the under-moderation and the
over-moderation configurations.
The relation between the water height and keff of each configuration are shown in Fig. 5. The reactivity worth per water level
change of each configuration was estimated to be about 0.62 ¢/mm
or 0.63 ¢/mm at a water height of 1000 mm. The accuracy of the
water height gauge of the modified STACY will be ±0.2 mm.
Therefore, the reactivity worth derived from a water height difference will have an accuracy of ±0.1 ¢, which will be acceptable as
the experimental precision. Thus, the reactivity worth of pseudo
fuel debris samples in each experimental configuration should be
greater than 0.3 ¢ to be distinguished from zero.

Geometrical buckling of each experimental configuration
should be minimally changed. The limitation of the changes was
determined to be less than 1%. This limitation was set to determine
the experimental limit, and there is no reason based on quantitative
consideration. Under these experimental conditions, the change of
the water heights should be less than 100 mm.
Therefore, favorable change of water height in reactivity worth
measurement should be from 0.5 to 100 mm, which corresponds to
reactivity worth of pseudo fuel debris samples from 0.3 to 62 ¢.

Fig. 3. “Over-moderation” experimental core configurations in square lattice of the
modified STACY.

condition are shown in Fig. 4. The thermal neutron flux of the overmoderation configuration was about 3 times as large as that of the
under-moderation configuration.
These are considered the base arrays with only fuel rods in the

Fig. 5. Relations the effective water height and keff of each core.


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S. Gunji et al. / Progress in Nuclear Energy 101 (2017) 321e328

3. Reactivity worth of samples
3.1. Samples for reactivity worth measurements
The samples of pseudo fuel debris simulating MCCI products
were modeled using several compositions which were made of
uranium dioxide with 235U enrichments of 3, 4, and 5 wt.% fuel; and
a concrete. A list of sample types in this study is shown in Table 1.

The composition of the concrete in this study is shown in Table 2
and its density is 2.3 g/cm3. Concrete volume fractions in the
samples were 0, 20, 40, 60, and 80%. Additionally, porosities of
these samples were varied from 0 to 80%, which were filled with
water. Concrete volume fraction is expressed by following equation
(2), and porosity is expressed by following equation (3),
respectively;

Concrete volume fraction %ị ẳ

Volume of concrete
Volume of MCCI product
100
(2)

Porosity %ị ẳ

Volume of moderator water
Volume of MCCI product þ moderator waterÞ
 100
(3)

Ideal sample loading conditions, based on reactivity worths for
each loading, were evaluated for five arrays with 1, 5, 5, 9, and 13
MCCI products samples loaded in the patterns illustrated in Fig. 6.
The in the test region of the under-moderation configuration fuel
rods were replaced with sample rods. For the over-moderation

configuration, the samples were inserted into vacant positions in
the test region.

Values of keff were computed for arrays of fuel rods and the
samples. Reactivity worths were estimated by comparing the keff
values and those of the base arrays. The estimated relative reactivity worth of the pseudo fuel debris whose 235U enrichment is
4 wt.% are shown in the following sub-sections.
3.2. Relative reactivity worths by changing the concrete volume
fraction
Fig. 7 and Fig. 8 show the computation results of relative reactivity worth dependency of the concrete volume fraction in each
configuration. They are the results of the samples based on using
the 235U enrichment of 4 wt.% fuels and their porosities are 0%.
Fig. 7 shows that the changing of the concrete volume fraction
has a big impact on the reactivity worths in the under-moderation
configuration. The reactivity worth of the samples with no concrete, shown in the figure, was negative because the 235U enrichment of 4 wt.% was lower than the 5 wt.% enrichment of fuel rods.
The absolute value was, however, small and the worth turned to
positive if the concrete volume fraction is beyond 40%. There was a
tendency that reactivity worths increase into the positive according
to increase the concrete volume fraction. It is considered that the
water in the concrete contributed to moderate of neutron in these
configuration.
Fig. 8 shows that for the over-moderation configuration, the
reactivity worths are negative for all of the patterns because the
samples excluded the moderator water. There was a tendency that
reactivity worths increase into the negative according to the concrete volume fraction increase. The positive reactivity worths
should have been inserted, because the moderation conditions of

Table 1
A list of the reactivity worth samples and their specifications.
Sample Materials

Length

Loading Patterns

MCCI product with zircalloy cladding
Parameters;
235
U enrichment (3, 4, and 5 wt.%)
Concrete volume fraction (0, 20, 40, 60, and 80%)
Porosity (0, 20, 40, 60, and 80%)

1420 mm

1, 5a, 5b, 9, and 13

Table 2
The composition of the concrete in this study.
Element

Number density [atoms/b cm]

Element

Number density [atoms/b cm]

Element

Number density [atoms/b cm]

H
O
C

Na

1.374E-02
4.592E-02
1.153E-04
9.640E-04

Mg
Al
Si
K

1.239E-04
1.741E-03
1.662E-02
4.606E-04

Ca
Fe

1.503E-03
3.451E-04

Fig. 6. Loading patterns of the reactivity worth samples into the test regions.


S. Gunji et al. / Progress in Nuclear Energy 101 (2017) 321e328

Fig. 7. Relative reactivity of the pseudo fuel debris samples by changing the concrete
volume fraction in the under-moderation configuration.

325

change in the critical water height is too much for a larger number
of samples.
Table 3 and Table 4 summarize the reactivity worth per each
sample in each moderator condition. In addition, samples of 100%
concrete and water with zircalloy cladding are shown as references.
In the under-moderation configuration, positive reactivity
worths were inserted by increase of the concrete volume fraction,
and more reactivity worth was inserted by insertion of the 100%
concrete sample. Table 3 shows the effect of the replacement of the
fuel rod of the water sample is approximately 12 ¢, and that of the
100% concrete sample is approximately 3.3 ¢ in each insertion
pattern. And 4 wt% fuel rods (see Concrete volume 0%) have
negative reactivity worths in each insertion pattern. Moreover, the
maximum reactivity worth was inserted by swapping the fuel
rods for water holes. In this configuration, the reactivity worths
per rod were almost the same for sample types in each insertion
pattern.
On the other hand, in the over-moderation configuration, small
negative reactivity worths were inserted by increase of the concrete
volume fraction. Table 4 shows the effect of the replacement of the
fuel rod of the water sample is approximately À2.5 ¢, and that of the
100% concrete sample is approximately À10¢ in each insertion
pattern. Sensitivities of both the insertion pattern and the concrete
volume fraction for the reactivity worths were small.
3.3. Relative reactivity worths by changing the porosity

Fig. 8. Relative reactivity of the pseudo fuel debris samples by changing the concrete

volume fraction in the over-moderation configuration.

this case were close to the suitable moderation condition by
removing water. However, contrary to expectations, the reactivity
worths remained negative. Maybe, these results show that dry
condition is nearly optimum moderation condition. Furthermore,
in this configuration, it is also concluded that loading of up to 5
samples will be suitable to measure their reactivity worth because

Fig. 9 and Fig. 10 show the computation results of relative
reactivity worth depend on changing the porosities of the sample in
“Pattern 5a” for several concrete volume fraction in each configuration. They are the results of the samples based on using the 235U
enrichment of 4 wt.% fuels. The relative reactivity worths in each
configuration has proportional relations to porosities.
Fig. 9 shows that the increasing of the porosities have moderation effects, therefore, the samples has a positive reactivity worth.
About 40 ¢ positive reactivities occurred by the porosity increased
from 0 to 80% in the under-moderation configuration. The effect of
porosity changing is dominant than that of the concrete volume
fraction changing, because the amount of hydrogen differ by one
order of magnitude between two parameters.
Fig. 10 shows that about 25 ¢ positive reactivities occurred by
the porosity increased from 0 to 80% in the over-moderation
configuration. This results show that this experimental core
configuration is not enough “over-moderation”, because the positive reactivity worths were inserted by increasing of water content.
3.4. Additional analysis for over-moderation core configuration
In section 3.3, it has turned out that “over-moderation” core
configuration was not have enough moderation ability. Therefore, a
new over-moderation experimental core configuration which

Table 3

The reactivity worth per sample rod in the under-moderation configuration. (Unit: cent/rod).
Samplea
Concrete
Concrete
Concrete
Concrete
Concrete
Concrete
Water
a
b

Volume
Volume
Volume
Volume
Volume
Volume

0%
20%
40%
60%
80%
100%

b

All sample has zircalloy cladding.
References.

Pattern 1

Pattern 5a

Pattern 5b

Pattern 9

Pattern 13

1.1
0.8
ỵ0.7
ỵ0.7
ỵ2.0
ỵ3.4
ỵ11.9

0.5
0.2
0.1
ỵ0.6
ỵ1.4
ỵ3.4
ỵ11.8

0.6
0.4
ỵ0.2

ỵ0.6
ỵ1.4
ỵ3.2
ỵ12.0

0.8
0.4
ỵ0.1
ỵ0.8
ỵ1.4
ỵ3.3
ỵ11.5

0.7
0.3
ỵ0.1
ỵ0.7
ỵ1.3
ỵ3.0
ỵ10.9


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S. Gunji et al. / Progress in Nuclear Energy 101 (2017) 321e328

Table 4
The reactivity worth per sample rod in the over-moderation configuration. (Unit: cent/rod).
Samplea
Concrete

Concrete
Concrete
Concrete
Concrete
Concrete
Waterb
a
b

Volume
Volume
Volume
Volume
Volume
Volume

0%
20%
40%
60%
80%
100 %b

Pattern 1

Pattern 5a

Pattern 5b

Pattern 9

Pattern 13

À7.4
À8.9
À9.5
À10.7
À10.2
À9.8
À2.4

À9.7
À10.0
À10.3
À10.8
À11.2
À10.5
À2.5

À8.3
À8.9
À9.5
À9.7
À10.5
À10.1
À2.6

À9.6
À10.0
À10.1

À10.6
À11.0
À10.2
À2.6

À11.0
À11.1
À11.1
À11.3
À11.4
À10.4
À2.7

All sample has zircalloy cladding.
References.

Fig. 9. Relative reactivity of the samples in the under-moderated “Pattern 5a”
configuration.

Fig. 11. A new “Over-moderation” experimental core configurations in square lattice of
the modified STACY.

Fig. 10. Relative reactivity of the samples in the over-moderated “Pattern 5a”
configuration.

shifted array of fuel rods was considered. This configuration is
shown in Fig. 11. In this configuration, local Vm/Vf (¼3.7) at the test
region do not change by insertion of the reactivity worth samples.
The relative reactivity worths of the samples by changing the
concrete volume fraction and the porosities are shown in Figs. 12

and 13, respectively. As considered in section 3.3, some features
of the over-moderation were seen in this core configuration. The
increase of moderator water by increasing of the concrete volume

Fig. 12. Relative reactivity of the pseudo fuel debris samples by changing the concrete
volume fraction in the new over-moderated configuration.


S. Gunji et al. / Progress in Nuclear Energy 101 (2017) 321e328

327

prepared. It was revealed that the experimental “over-moderation”
core conditions in this study was not enough over-moderation
condition for the sample of pseudo fuel debris. Therefore the
“new” over-moderation core configuration was analyzed in this
paper. This configuration was good to evaluate of the criticality
characteristics for high concrete volume fraction samples.

5. Further studies

Fig. 13. Relative reactivity of the samples in the new over-moderated “Pattern 5a”
configuration.

The experiment plans drafting in the modified STACY is carried
out continuously. Further discussion is necessary on criticality
characteristics of the 1F-NPS fuel debris. For example, water content of fuel debris, MCCI products, and usage of neutron absorber
materials should be studied before the experiment using the
modified STACY. In this paper, a combination of enriched uranium
fuel, concrete and water was considered as a first plan, other

combinations (burnup, cladding, steel construction, control rod,
and so on) should be studied in near future. It is scheduled to
conduct the actual measurements of reactivity worth for those
materials using the modified STACY after FY 2020.

Table 5
The reactivity worth per sample rod in the new over-moderated configuration. (Unit: cent/rod).
Samplea
Concrete
Concrete
Concrete
Concrete
Concrete
Concrete
Waterb
a
b

Volume
Volume
Volume
Volume
Volume
Volume

0%
20%
40%
60%
80%

100%b

Pattern 1

Pattern 5a

Pattern 5b

Pattern 9

Pattern 13

À4.2
À4.7
À7.4
À9.4
À11.5
À13.1
À8.1

À3.7
À5.5
À7.8
À11.0
À15.5
À20.6
À18.5

À3.4
À5.1

À6.7
À8.9
À11.7
À13.7
À8.8

À3.3
À4.9
À7.2
À9.9
À13.6
À17.3
À14.0

À3.5
À5.5
À8.2
À12.3
À18.8
À29.4
À31.2

All sample has zircalloy cladding.
References.

fraction or the porosities caused insertion of negative reactivity
worths. Especially, in these graphs, the relations of the concrete
volume fraction and the reactivity worth or the porosity and the
reactivity worth are characterized by not being linear.
Table 5 shows the reactivity worth per each sample in each

moderator condition with references. In this configuration, negative reactivity worths were inserted by increasing of the concrete
volume fraction, and more negative reactivity worth was inserted
by insertion of the 100% concrete sample. Moreover, minimum
negative reactivity worths were inserted by swapping the fuel rods
for the 100% uranium fuel without water. These features in the over
moderation configuration have not seen in the past “over-moderation” configuration described in section 3.3.
4. Conclusions
As a part of design works of critical experiments, core configurations to measured reactivity worth of MCCI products were studied. It was found that the measurements using the modified STACY
in under- and over-moderation configurations with pseudo fuel
debris simulating MCCI products are feasible because the worth can
be estimated with enough accuracy from change of the critical
water height. The suitable loading numbers of the samples were
estimated. From these results, it is possible to determine the
amount of the pseudo fuel debris sample which should be

Acknowledgments
This report includes results of the contract work funded by the
Nuclear Regulation Authority (NRA)/the Secretariat of NRA of Japan.

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