Conversion Studies Based on Caramel Fuel

As already described above, the Caramel fuel plate takes the form of two thin sheets of zircaloy, enclosing a regular array of rectangular pieces of U0₂, separated by small pieces of zircaloy. These plates are assembled in parallel between two side plates to which they are welded, and equipped with a

foot and a handling head to form the fuel element.

It results from this description that these Caramel fuel elements are quite similar in geometry to the currently used UA1 elements of the flat or

slightly curved MTR type. Therefore, the Caramel assemblies can fit very easily with a wide range of research and test reactors. Especially for the 10 MW

reactor under consideration there is no difficulty in implementing such a conversion.

Caramel fuel has strong nonproliferation characteristics due to the high specific weight of the U0₂ fuel (10.3 g/cm³); it can lead to a core with a volumetric uranium weight as high as 2 kg U/dm³ and thus to an enrichment in

the range as low as 3% to 10% 23⁵U. The calculations performed by the CEA have confirmed the feasibility of this solution for converting HEU cores directly to LEU cores in most cases. Furthermore the Caramel has several specific advantages:

- fuel compartmented and operated at a rather low temperature - negative Doppler coefficient in case of power excursion

- good chemical behavior in demineralized water which suppresses the necessity of a hot layer at the pool surface (this can be a great

advantage in countries with hot and wet summers).

In the case of this 10 MW reactor one must find the optimum design in balancing neutronic and thermal-hydraulic aspects. For such a power range one has to increase the heat surface area, and therefore to divide the lattice with

rather numerous thin fuel plates and thin channels adjusted to keep a reasonable moderation ratio and a good reactivity. These evaluations have been performed for the OSIRIS reactor conversion; they have resulted in the following data used as convenient for this 10 MW reactor:

fuel plate thickness 2.25 mm

oxide thickness 1.45 mm

coolant channel thickness 2.6 mm plate number per assembly 16

active plate length 60 cm uranium weight per assembly 7.65 kg U

From a neutronics point of view, the enrichment has been determined to reach the same cycle length as the HEU design. This is a low value of the enrichment, and the fluxes are similar to those achieved with the HEU.

From a thermal-hydraulics viewpoint, on the basis of the present safety criteria required by the French Regulation Authorities, the conversion appears feasible in very good conditions.

Table 2-14 summarizes the characteristics of both the original HEU core and the Caramel converted one.

Table 2-14. Summary of Characteristics of Original 10 MW HEU Core and Caramel Converted Core.

93 % U235 Caramel 6.5% U₂₃₅ fuel element dimensions (mm) 76 x 80 76 x 80 number of plates in standard fuel element 23 16

water gap (mm) 2.1 2.75

plate thickness " 1.27 2.25

meat thickness n 0.51 1.45

23 5U weight by element (g) 280 500

active height (mm) 600 600

average cycle length (days) 16.4 16.4

fast flux in central water hole(at the center) 13 13 1 >0,9 MeV (averages height) (n/s/cm²) 5.40 10 5.81 10 thermal flux in central water hole

(at the center) 2 14 14

*4 ^<0,625 eV (averages height (n/s/cm ) 2.8 10 2.6 10

thermal flux in water reflector 14 14

*4 <0,625 eV (averages height) (n/s/cm²) 1.0 10 1.0 10

water velocity (m/s) 1.8 2

flow rate (m³/h) 600 630

59

2.4 "BENCHtARK" CALCULATIONS

2.4.1 Definitions and Aims

In order to compare reactor physics methods used in various research centers, benchmark problems were calculated by seven international centers for well defined reactor conditions. Since the emphasis of these calculations is on

the comparison of the results, rather than on their absolute values, the reactor configurations were idealized and simplified as much as possible. Thus, these calculations may not correspond to realistic conditions and conclusion about actual reactor performance with REU fuels should not be drawn from them, even though some results are very similar to the results of the generic studies.

The specifications of the benchmark problems are provided in Table 2-15 and in Fig. 2-25. Briefly, they correspond to a 10 MW, 6 x 5 element core

reflected by a graphite row on two opposite sides, and surrounded by water. The standard MTR-elements contain 23 fuel plates. The enrichments considered are 93%, 45%, and 20%, and each of these correspond to a 235U content of 280, 320, and 390 grams per element, respectively. The calculations were to be carried out with Xe-equilibrium and for various burnup conditions. The main data to be

calculated were the absolute reactivities keff as well as the subsequent reactivity differences and the flux distributions.

Some ambiguity was caused by the fact that the burnup states were specified in terms of % (i.e., percentage loss of the number of ^{2 3 5}U atoms). Cores of

different enrichments contain different amounts of ^{23 5}U, and their burnup in MWd is very different when they have the same burnup in %. Since it is the burnup in MWd that is more significant, some technical groups assumed that the % burnup of the specifications applied only to the 93% enriched case, and used the corresponding MWd burnup for all other cases. Other groups used the % burnup in all cases. Results for both choices are used in this summary.

2.4.2 Results

The results of the seven contributors can be divided into two parts, i.e., absolute reactivities plus reactivity steps and absolute fluxes plus flux ratios for different enrichments.

The starting point for the comparison of the reactivities is the koo-behaviour for the three enrichments. The great number of results forces to plot the infinite reactivity versus burnup in percentage loss of U-235 for each

enrichment separately (Fig. 2-26 - 2-28). The overall impression of these three figures is a good agreement within the majority of the contributors. Deviations exist for the JAERI-calculations* and some small deviations for the EIR-results.

Based on this agreement the effective reactivities for the core calcula-tions as given by Table 2-16 show similarly small deviacalcula-tions below 1% Ap from

each other (with only one exception) as may be seen from Table 2-17. The INTERATOM results are an arbitrary choice of a basis. These reactivities are backed by

another interesting keff comparison which was obtained by ANL by running detailed 3-D, continuous energy Monte Carlo calculations, and comparing their results with those of diffusion-theory calculations. The results are listed in Table 2-18.

*For the reasons see Appendix F-6, Section 1.3.

Table 2-15. Specifications for the Methodical Benchmark-Problem Aims: Comparison of the different calculation methods and cross-section data

sets used in different laboratories, limited conclusions for real con-version problems.

Data and Specifications Agreed Upon:

Active Core Height 600 mm

Extrapolation Length 80 mm (in 80 mm distance from the core, the cosine-shaped flux goes to zero)

X-Y Calculations only

Identification of the remaining plate positions of the control element:

4 plates of pure aluminum PA1 = 2.7 g * cm^-3, each 1.27 mm thick in the position of the first, the third, the twenty-first, and the twenty-third standard plate position; water gaps between the two sets of aluminum plates.

Specifications of the different fuels (UAlx-Al Fuel) for HEU, MEU, LEU corresponding to the previous definitions:

HEU: Enrichment 93 w/o (weight Z) U-235

Homogeneous Xenon content corresponding to average-power-density

Results

1 Y

outside boundary condition = 0

A

3fuel edement width of water reflector , 231mm

graphite grophite water

rrmm 25% 5% water

i_

BOL- Core

outside -- 3fuel element widfh boundary

of water reflector condiion

-243 mm

45% 25%

Control Elemenr

5%

e45- x45 % 25%

. . .~~~~~~~~~~~ ___- X

81mm

Burnup step 5%

graphite gaphite water

30% 10% water 50% 30%° _Control 30% 10%

Element_

[__ r r

EOL- Core

graphite block cross section 77 mm x 81 mm graphite density 1.7gcrm

³

1

_{jjb iV o I V u}

1

r""Ii I

Burnup definition : (%) means the percentage of loss of the number of U 235 -Atoms

METHODICAL BENCHMARK 10 MW CASE

CORE CROSS SECTION

I FIG. 2-25

Ko

18

- -ANL

--- INTERATOM -- -dOSGAE

-- ·- CEA '(Adaption MWd--% by EIR - Table) --- EIR

-*-.... . CNEA"

JAERI

mXe-free-values not available

Xe-1,6

15

1.4

13

1,2

0 5 10 15 20 25 30 35 40 45 50 W [%

10 MW - BENCHMARK

Koo FOR 93% ENRICHMENT AS A

FUNCTION OF THE BURNUP W [%]

fig. 2- 26

63

K,,

1,8

ANL - INTERATOM

- - OSGAE

- CEA' (Adaption MWd-% by EIR -Table ) EIR

... CNEA ^" (not calculated)

-- JAERI

' Xe-free-values not available

Xe-J free

-1,6

-1,5

1,4

1,3

1,2

0 5 10 15 20 25 30 35 40 45 50 W[%]

10 MW - BENCHMARK

Koo FOR 45 % ENRICHMENT AS A fig. 2-27

FUNCTION OF THE BURNUP W [%]

K. A

1,8

t

ANL INTERATOM

OSGAE

CEA (Adaption MWd-% by EIR-Table ) EIR

CNEA X JAERI

' Xe-free-values not available

Xe-free

1,6

1,5

1

N.

'N

N,'N 'N

-».

NN\

N '»._Ns

NN, N'

NN

1,3

1,2

0 5 10 15 20 25 30 35 40 45 50 W [%

10 MW - BENCHMARK

Koo FOR 20 % ENRICHMENT FUNCTION OF THE BURNUP

AS A

W [%] fig.2-28

65

TABLE 2-16: REACTIVITY LEVEL (KEFF) (INTERATO) USA SWITZERL. AUSTRIA FRANCE ARGENTINA JAPAN 93 BOL 1.0328 - .90 % + .37 % - 0,08 % + ,71 X - 0,45 % + O,5 X

TAR] P )- IQ. _{--DLL L IV -n rll} ANI -P$:(Zil TcZ _{-- LUVLI-} cn~thptznN CI_{V \»wn n -l vI --} MfnNTF CRI _{1-r-.-. rv -} n* ANn nrFFIISTRN TIHFFRY FIRFNVAI IIFR_{" --- * ' ---- w . ... L--- v---"} v--CORE MONTE CARLO KC_ 6' DIFFUSION KD KMC - KD (KMC - KD)/ 6

93 % FRESH 1.189 ± .0033 1.18343 + .006 1.82

20 % FRESH 1.168 ± .0033 1.16830 .0 0,

93 % EOL** 1.045 ± .0036 1.03366 + ,011 3.06 20 % EOL^'* 1.048 ± .0034 1.03934 + .009 2.65 20 % EQ MWD** 1.072 + .0027 1.06847 + .004 1.48

100,000 HISTORIES PER CALCULATION

** THESE CALCULATIONS DID NOT INCLUDE LUMPED FISSION PRODUCTS

TABLE 2-19: COMPARISON OF REACTIVITY-LOSS (a_ V[ S( ) -BY BURNUPBRNILLE

ENRICH- EQUAL USA GERMANY AUSTRIA SWITZL, FRANCE JAPAN ARGENTINA MENT BURNUP (ANL) (INTER- (OSGAE) (EIR) (CEA) (JAERI) (CNEA)

IN ATOM)

93 W/O MWD 2,24 % 2.18 % 2,21 X 2.19 % 2.21 % 1.88 % 2,22 %

MWD 1.61 % 1.53 X

45 W/O ..--- .--- .____ _ __.. -- --- ._.--- --- _ X 2.08 X 1.92 X 2.09 % 1.99 % 2.06 X 1.70 X

MWD 1.10 % 1.03 X

20 W/O --- _______---- ._ _ --- .- ._--- ._--- . .---.-- ---X 1.94 % 1.81 1,92 % 1.75 X 1.91 X 151 X 1,93 X

67

As the main purpose of this check of methods is the comparison of the reactivity differences for given burnup-steps at the different enrichments as well as for the reactivity differences caused by the enrichment reduction these results are compiled within the Tables 2-19 and 2-20, respectively. These differences are of great importance for the determination of the U-235-loading necessary when reducing the enrichment. With one exception all reactivity differences are in good agreement as far as the loss by burnup is concerned.

Table 2-20 shows some relatively great deviations which supply the extreme

figures far away from each other. Nevertheless the majority of the contributors are fairly close together.

To compare the different flux distributions of the contributors flux ratios were plotted for three cases (Figs. 2-29, -30, -31). All comparisons were carried out for the core state BOL with Xenon-equilibrium and along the

x-axis only. For the fast flux the ratio of the LEU-case to the HEU-case was plotted only (Fig. 2-29) which delivers excellent agreement in two sets of results which must be distinguished. With equal burnups in % loss of U-235 for both enrichments all ratios end at 1.0 in the reflector with very small differ-ences inside the core area except the OSGAE-results. With equal burnups in MWd for both enrichments there exists a clear difference of 5% loss of fast flux in the reflector area as well as a somewhat different behavior inside the core.

For the ratios of thermal fluxes the comparisons were carried out for both reductions under consideration, i.e., to MEU-fuel (Fig. 2-30) as well as to LEU-fuel (Fig. 2-31). The differences between the two sets of burnups (equal

percentage loss of U-235 and equal MWd, respectively) are found again. Within these sets there exists excellent agreement. Only the OSGAE-results are to be found within the gap between the two sets in the reflector area whereas the two sets are clearly separated for the other contributors.

From the figures one can get the rough values for the reduction of the thermal flux caused by the enrichment reduction as

- 14-17% in case of MEU

in the fuel area - 31-38% in case of LEU

and

0-2% in case of MEU

- 0-4% in case of LEU in the reflector area 0-4% in case of LEU

for equal burnup in MWd only. Using the equal %-loss of U-235 these small

reductions in the reflector are reduced to zero. But it must be emphasized that at the position of the thermal flux peak in the reflector (x ~ 27 cm) there exists a clear reduction of the thermal flux. A similar comparison for the y-direction with somewhat different results due to the graphite reflector elements used there may be carried out by interested users on basis of the various results of the different laboratories.

A last aspect worth mentioning from a proliferation point of view is the plutonium content of the burned fuel. The data on ^{2 3 9}pu content shown in

Table 2-21 were compiled from the results of cell calculations at 50% burnup for the three enrichments. They are intended only for comparing results obtained at the various laboratories. More accurate data for real reactors can be obtained from the results of the generic studies (see, for example, Appendix A, Section A.6).

TABLE 2-20: COMPARISON OF REACTIVITY-DIFFERENCES (4?) BY ENRICHMENT REDUCTION (10 MW-BENCHMARK) AUSTRIA FRANCE GERMANY JAPAN SWITZERL, USA ARGENTINA (OSGAE) (CEA) (INTER- (JAERI) (EIR) (ANL) (CNEA)

ATOM)

93 W/O FRESH FUEL - 0,49 X - 0.49 - 0,70 % 0.00 % - 1.05 % - 0.38 %

% + 0.13 % + 0,03 % - 0,16 % + 0.63 % - 0,58 % + 0.13 % BOL --- -.-. _ --- _____ _---___ ____---.-______ -__ __

MWD + 1.35 X + 1,65 %

% + 0.25 % + 0.19 % + 0,06 % + 0,82 % - 0.38 % + 0,29 %

45 W/O EOL --- __ - --

MWD + 2.00 % + 2.29 %

93 W/O FRESH FUEL - 1.08 X - 1,05 % - 1,47 % + 0.17 X - 2,49 % - 1,09 % - 1.32 % X

~

+ 0.00 X - 0,09 % - 0,47 % + 1.44 % - 1,80 % - 0.20 % - 0.42 % BOL --- .---. _ _ --- _ ._ _ __ _

_---IWD + 2.48 % + 2.84 %

%

X + 0.29 % + 0.20 % - 0.10 % + 1,81 - 1,36 % + 0.10 % - 0,12 %

20 W/O EGL ---. __ - ---

---;i'.iD .+ 3,63 % + 3,98 %

Table 2-21. Pu-239-Content at 50% Burnup in Grams per (Obtained from Cell Calculations)

Fuel Assembly

OSGAE 93 w/o U^{2 3 5}

45 w/o U^{2 3 5} 20 w/o U^{23 5}

0.42 4.34

ANL

0.44 4.24

INTERATOM 0.42 4.41

EIR

0.45 5.50

JAERI 0.37

CNEA 0.43 3.32

12.30 12.17 11.92 14.80 9.13 12.71

In conclusion, all the comparisons which have been performed as part of the benchmark studies, including k,, keff, Ap, and flux-distributions as a function of burnup and enrichment indicate that the calculations carried out at the different laboratories and companies are in good agreement with each other.

Some minor exceptions may bring these contributors to a recheck of their methods and calculations to find the reasons for the deviations.

69

20/f3 o20/013

1,1 +

1,05

0,95 MWd

ANL MWd

Q9 . -- INTERATOM MWd

--- OSGAE %

-. CEA %

0,85 0.,85^ - -.... EIR ~CNEA % %

-fuel INTERATOM %

0

t -- JAERI %

0 5 10 15 20 25 30 35 40 45 50 X [cm]

' ANL %

10 MW - BENCHMARK A

th th Flux Ratios (f0s'f'f9c) vs. X - Distance FIG 2-29 450/93 Core State: BOL, Xe - Equil.

1,05

1,0 -'-- %

09 \ ANL MWd

z\1l /,/' ~ --- INTERATOM MWd

\V

i'/ --

OSGAE

^%

85- EIR %

…/--- t-rough ----

INTERATOM %

position -- JAER

0,8 of the

thermal .A N L %

-fuel --- flux peak

0,75

^{, ,. , ,__,,, ,, ,,_}

0 5 10 15 20 25 30 35 40 45 50 X [cm]

10 MW - BENCHMARK

Flux Ratio (0h/0th ) vs. X - Distance FIG 2-30

Core State: BOL, Xe - Equil.

1

th / th Q20/ 93 1,1

1,0

0,9

0,8

0,7

}%

} MWd

- ANL MWd

-- INTERATOM MWd

---- OSGAE %

--.- CEA %

---- EIR %

rough ... CNEA %

position ---- INTERATOM %

of the

thermal --- JAERI %

flux peak

--.-- ANL %

0,5

0 5 10 15 20 25 30 35 50 X cm]

10 MW - BENCHMARK

Flux Ratio (a, /0) Core State: BOL, Xe

vs. X- Distance - Equil.

FIG 2-31

3.0 STATUS AND DEVELOPMENT POTENTIAL OF RESEARCH AND TEST REACTOR FUELS 3.1 OVERVIEW

Table 3-1 summarizes the status, as of March 1980, of reduced enrichment fuel availability from commercial research reactor fuel suppliers. Further data on the availability and development potential of these fuels can be obtained from Tables 1-1 and 1-2 in Section 1 and from Appendix H.

3.2 STATUS OF PLATE-TYPE FUEL TECHNOLOGY

Development of high density fuels for high flux/power research reactors has already led to considerable fabrication and irradiation experience with high uranium density plate-type fuels (see Table 3-2). At the moment the highest

uranium densities routinely used are in the range 1.1 - 1.7 g/cm³ (ATR, HFIR, BR-2, RHF, ORPHEE).

In recent years, with the prospect of enrichment reductions to 45% and 20% instead of 93%, important research and development work has been started in Europe by the companies CERCA and NUKEM, in Argentina by the CNEA, and in the United States by the Department of Energy (DOE) under the Reduced Enrichment

71

ESTIMATED SCHEDULE

Table 3-1

OF TESTS ON REDUCED ENRICHMENT FUELS FOR RESEARCH REACTORS (Status of March 1980)

Fuel Type Element Uranium Density 79 80 81

Configuration in Fuel Meat g/cm³

U-ZrH rod <0.75 D

1.3 C D

2.2 C

3.7 C

U-A1 alloy plate 1.1 C

UAlx-Al plate 1.6-1.7 C D

2.0 A B C

2.2/2.8 A B C

U₃0₈-Al plate 1.7 C D

2.1-2.5 A B C

-3.0 A B

U0₂ plate 4.5 A

plate ~9 C D

rod ~9 C D

U₃Si-A1 plate 4-8 A B

LEGEND: A Beginning of small-sample irradiation tests.

B Results from small-sample irradiation tests available.

C Results from full-size element irradiation tests available.

Table 3-2. Reactors Currently Using Fuels With High Uranium Density

*Without hot channel factors.

Research and Test Reactor (RERTR) Program managed by Argonne National Laboratory.

The objective of the above programs is twofold: firstly to increase the uranium content of existing plate-type fuels, e.g. alloy, and aluminide and U₃0₈ dis-persions; and secondly to examine and develop newer high density fuels, such as U₃Si. At the same time, fabrication development is underway to produce thicker fuel meats for those reactors which can accept them.

CERCA, with more than 170,000 plates delivered up to now, has already fabricated (in 1972) for the French CEA UAlx-Al prototype elements with a uranium density of 1.7 g/cm³, which were qualified with very good results at

burnups of 58% (mean value, maximum attained-70%). Plates made of U₃0₈-Al dispersions with a uranium density of 1.7 g/cm³ were also tested with satis-factory results.

NUKEM, with a capacity of 20,000 plates per year, has delivered up to now almost 200,000 fuel plates within 18 years of MTR fuel program. In this period more than 2 1/2 tons of HEU in the form of UF₆ and additional amounts of

recoverable scraps have been converted to uranium metal. Using this metal, UAl-alloy and UAlx-Al fuel elements have been fabricated for various research

and test reactors, mainly in Europe and the U.S. Different core conversions in European research reactors have been done with NUKEM advanced type fuels in

order to increase power and neutron flux. For this purpose dispersed fuels (aluminides and oxides) and dead burned U₃0₈, partly mixed with burnable poisons, have been developed and successfully tested.

In the U.S., both Texas Instruments (TI) and Atomics International (AI) operate plate-fuel fabrication lines for the DOE. These facilities together fabricate about 20,000 plates per year for U.S. reactors, principally the ATR and HFIR, with some elements also supplied to U.S. universities, under a DOE university assistance program.

3.2.1 UAlx-Al Fuel

In the course of the development program, CERCA is preparing this year two elements with 45% enriched uranium and a uranium density of about 1.7 g/cm³ for irradiation in the ORR reactor. With this same enrichment of 45%, CERCA will also fabricate 300 plates with a uranium density of 1.6 g/cm³ in

the fuel meat for the Kyoto University Critical Assembly (KUCA) in Japan. The next step is to increase the uranium density to about 2.1 g/cm³ in thin (1.3 mm) or thick plates (2.2 mm). An irradiation experiment with two elements is planned in the EURATOM reactor at Petten (The Netherlands) within the scope of a cooperative agreement between ANL, CERCA, NUKEM, and the European Community.

The technology for uranium densities up to 2.2 - 2.3 g/cm³ is presently well mastered with UAlx-Al, so that irradiation of elements with 20% enriched uranium are planned in the ORR reactor for 1980.

At NUKEM, prototype fuel elements, with UAlx-Al fuel and a uranium density of up to 2.2 g/cm³ in the fuel meat, without changing the geometry of

fuel plates and number of plates per element, will be delivered to several reactors in Europe in 1980. In connection with the national R & D program in Germany for using reduced enriched fuels, NUKEM is going to insert prototype

fuel elements with up to 2.6 g U/cm³ in the fuel meat in the ORR and European research reactors for examination of irradiation behaviour, life time, core physics, reactivity studies, etc. This is planned step by step starting soon and within the following 12 to 15 months.

CNEA-Argentina has obtained good results in fabricating miniplates with 2.2 g U/cm³ in the fuel meat using natural uranium. Plans are being made for irradiation testing of miniplates in the ORR and a prototype fuel element in the RA-3 reactor.

The U.S. effort for UAlx-Al is divided into two areas: fuel development and irradiation testing at EG&G, Idaho; and full-scale element fabrication develop-ment at Atomics International. Fuel developdevelop-ment is underway with the irradiation

of mini plates containing up to 2.6 g U/cm³ beginning early in 1980 in the ORR.

Fabricability tests already completed at AI have indicated that full scale plates containing at least 2.2 g U/cm³ in the fuel meat can be fabricated easily.

3.2.2 U₃0₈-A1 Fuel

CERCA is also developing the technology of U₃0₈-A1 dispersions, as it appears that UAlx-Al fuel plates may be limited in the future, from technical and economical reasons, to uranium densities in the range 2.5 - 2.8 g/cm³. Now the U₃0₈-A1 technology is at hand for 3.0 g U/cm³ in the meat, and irradiation

experiments are planned in ORR. Uranium densities in the range 3.3 - 3.8 g/cm³ are expected to be reached in the near future.

NUKEM has set up and scheduled a similar program for U₃0₈-A1 dispersion fuel. Irradiation experiments will be conducted in the ORR and in European research reactors with fuels containing up to 3.2 g U/cm³ in the fuel meat.

The CNEA-Argentina fuel development program for U₃0₈-A1 fuel is similar to that for its UAlx-Al fuel. Irradiation tests on miniplates in the ORR and prototype elements in the RA-3 are being planned with uranium densities in the range 2.4 - 3.0 g/cm³.

The CNEA-Argentina fuel development program for U₃0₈-A1 fuel is similar to that for its UAlx-Al fuel. Irradiation tests on miniplates in the ORR and prototype elements in the RA-3 are being planned with uranium densities in the range 2.4 - 3.0 g/cm³.

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