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Fission product release from Triso-coated UO2 particles at 1940 to 2320°C
Journal of Nuclear Materials 98 (1981) 107-115 North-Holland Publishing Company
107
FISSION PRODUCT RELEASE FROM TRISO-COATED UO? PARTICLES AT 1940 TO 2320°C
Yuji KURATA, Katsuichi IKAWA and Kazurnl IWAMOTO Japan Atomic Energy Research Institute, Tok-ai-mura,Naka-gun, Ibaraki-ken, 319-l 1 Japan Received 16 December 1980
The fiiion product release from TRISO-coated UOa particles was measured by postactivation heating at 1940 to 2320°C for use in a safety analysis. The results are analyzed mathematically with effective diffusion coefficients in each medium. lo3Ru, ggMo and gsNb are released at 1940 to 2320°C and have high effective diffusion coeffldents. Although 140Ba and 137Cs are retained in TRJS0coate.d particies at 205O”C, they are released rapidly at 2320°C. This is attriiuted to the transition of beta to alpha SE at 2320°C. r4tCe, 140La and gs2r are released little if any at 2320°C. Rare gas nuclides, iodine and tellurium seem to be retained in coated particles at this high temperature.
1. Introduction It has been shown by the results of irradiation tests that coated fuel particles for the high-temperature gas-cooled reactor (HTGR) behave fairly well under normal reactor conditions. According to a safety analysis [l] of the experimental HTGR, there is a possibility of fuel temperature excursion exceeding 2OOO’C under hypothetical accident conditions such as blocking of coolant channels or a reactivity accident. There were several studies on the behaviour of coated fuel particles during a thermal excursion above 2OOO’C. Some data on failure behaviour of coated particles are already available in the literature [2-S]. Most of them are of BISO-coated particles. Data on the failure of TRISOcoated particles with a Sic layer for use in the experimental HTGR being developed in Japan are also reported [6]. As recently reported [7], heat treatment above 2OOO’Chas an effect on density and microstructure of Sic of the coated fuel particles. For a safety analysis, it is essential to evaluate the amount of fission products released from fuels. Many works have been made on fusion product release at temperatures under normal operating conditions of HTGR. But few works have been made at temperatures under hypothetical accident conditions. Schenk and Naoumidis [3] reported the fission product release in HTGR fuel above 2OOO’Cunder the condi-
0 022-3 115/8 1/OOOO-0000/$02.50 0 North-Holland
tions of unrestricted core heatup events. Their fuels were spherical fuel elements that had uranium-thorium mixed oxide particles with a pyrocarbon coating deposited from methane (I-III-BISO). It is expected that the behaviour of fwion product release in TRISOcoated particles with UOa kernel differs frorr that in BISOcoated particles. An experiment of post-activation heating above 2OOO’Cwas carried out to examine the fission product release from slightly-irradiated TRISOcoated fuel particles. Then the results were analyzed mathematically with effective diffusion coefficients in UOa, pyrolytic carbon and Sic as parameters.
2. Experimental The samples used were the same TRISOcoated particles as in the previous experiment [7]. The materials, dimensions and densities of the components of the particles are listed in table 1. The innermost layer is the buffer layer deposited from acetylene, the second and the outermost layers LTI PyC (low temperature isotropic pyrolytic carbon) from propylene and the third layer Sic from methyltrichlorosilane. The sample particles were irradiated at ambient temperature (approximately 8O’C) to a thermal neutron dose of 6 X 1016 or 6 X 10” n/cm?. After irradiation, the particles were allowed to cool for a fured period.
Y. Kurata et al. /Fission product release porn UOz particles
108
Table 1 TRISO particle parameters Component
Material
Thickness OImf
Density Q./cm31
Fuel kernel Innermost layer Second layer Third layer Outermost layer
uoz PyC
600 w 60 30 25 45
10.53 1.10 1.85 >3.20 1.85
PYC
sic P,C
Q,= diameter.
The period was one week or more for measuring the short-lived fusion products such as “MO, 133Te, t3sI, 133Xe, “‘Ba and “‘Ru, two or three months for 95Zr, “Nb and “Ice and about ten moths for 13’Cs. Following this, the intact particles were dispersed in graphite powders, placed into a graphite crucible and heated in the apparatus shown in fig. 1. The crucible was evacuated to 10m5bar. Solid fission products were absorbed on the graphite components, which were composed of graphite powders, a graphite crucible, a graphite supporter and a graphite felt sleeve as shown in fig. 1. Gaseous ftion products were monitored by providing a charcoal trap cooled with dry ice at the outlet of the vacuum pump. As the sample particles failed in heating at 2450 to 2500°C for 15 to 30 min in this apparatus, post-activation heating was made between 1940 and 2320%
Temperatures were measured by an optical pyrometer through a hole in the lid of the graphite crucible and corrected for the absorption of lens, prism and quartz plate. The sample particles were assayed by gammaray spectrometry before and after heating. The graphite components and the charcoal trap that had absorbed fission products released from the sample particles were also assayed similarly. Fractional release of each fusion product was calculated from these data. Most of solid fmion products were absorbed on the graphite components. But some fission products, for example “‘Cs and to3Ru, were found on the graphite wool or zirconia when the fractional release became high. In this case, the fractional release was calculated from the remainder of activities counted before and after heating. It was difficult to measure the fractional release of rare gas nuclides and iodine directly in this experiment, because an effective collector for these gaseous nuclides was not prepared in this apparatus. Therefore, the activity of each fission product in coated particles was measured before and after heating and compared with a decay curve of each fission product in the experiment of short-lived fission product release.
to-’
I
0
’
too
I
’
200
Heating
time
I
300
’
I
400
(min)
Fig. 2. Fractional release of fission products from TRISOFig. 1. Apparatus for postrctivation heating of coated fuel IXUtiCleS. coated fuel particles at 1940°C.
Y. Kurata et al. /Fission
productrelease from UOspurticles
109
3. Results 3.1. Release of solid fission products
Fractional release of fission products from the sample particles was measured at 1940, 2050, 2200 and 2320°C. Heating at 1940 and 2050°C was made up to 400 min. The results are shown in figs. 2 and 3. Only release of lo3Ru, 99Mo and “Nb of the solid fission products is detected. The rate of increase in the fractional release tends to decrease at 205O’C with heating time. Fig. 4 shows the fractional release of fission products at 2200°C up to 180 min. The fractional release of rsoBa is measured at 2200°C in addition to that of ro3Ru, 99Mo and “Nb. The fractional release of r3’Cs in heating at 2200°C for 180 min is almost the detection limit. The release of “‘Cs which is one of the important nuclides from the pbint of view of a safety analysis, is kept low at 2200°C. Fractional release at 2320°C up to 100 min is shown in fig. 5 similarly. The fractional release
‘7’
Heating
0
50
100
150
200
l-taoting time (min) Fig. 4. Fractional release of fdn coated fuel particles at 2200°C.
time(tin)
Fig. 3. Fractional release of fission products from TRlSOcoated fuel particles at 2050°C.
products
from TRlSO-
of 13’Cs and r4rCe is measured in addition to that measured at 2200°C. The rate of increase in the fractional release of ro3Ru, 99Mo and “Nb at 232O’C decreases with heating time in the same way as that measured below 2200°C. But it is noticed that the fractional release of r4’Ba and “‘Cs increases rapidly at about 30 min in heating at 232O’C. This implies that a change of the diffusion behaviour of 14’Ba and 13’Cs in coated particles occurs at 232O’C. 9sNb was observed in the graphite components that absorbed fission products, but it was not 9sZr, which is a parent nuclide of 95Nb. It is found that 95Nb is released from coated particles but 95Zr is retained in coated particles. Both 14’Ba and r40L.a were observed in the graphite components. Since r4’La is a daughter nuclide of 14’Ba, the radioactive equilibrium of r40Ba-r40La was calculated. The result indicates that “‘La found in the graphite components almost depends on lroBa released from
110
Y. Kumta et al. /Fission product release from U02 prrrticles
I Time
(day)
a 0
20
40
60
80
loo
420
Heating time (min) Fig. 5. Fractional release of fission products from TRISOcoated fuel particles at 2320°C.
coated particles and that the release of 14’La is not significant. According to these experimental results, solid fission products released at this high temperature are classified into the following groups: (1) 103Ru, 99Mo and 95Nb that have high fractional release at 1940 to 2320°C, (2) 14’Ba and 13’Cs that begin to be released at 22OO’Cand are released rapidly at 2320°C, and (3) 141Ce, 140La and 9sZr that are released little if any at 232O’C.
Fig. 6. Activity of each fission product in coated fuel particles and heating history. Dotted lines indicate decay curves.
high fractional release deviate from each decay curve when the coated particles are heated and fission products are released. But the curves of activities of these nuclides versus time are parallel to each decay curve as shown in fig. 6 while the coated particles are not heated. Activities of “‘Xe, “‘1 and 132Te measured after heating lie on each decay curve. Similarly it is shown in fig. 7 that the activities of ‘O’Ru, 99Mo and 14’Ba are apart from each decay curve but those of “‘Xe, 1311and 132Te he on each decay curve in heating at 2320°C up to 100 min. Then neither lJ3Xe, 1311, 132I nor 132Te was observed on the charcoal trap and graphite components. These results indicate that rare gas nuclides and iodine seem to be retained in coated particles even at this high temperature.
3.2. Release of gaseous fission products Release of gaseous fission products was examined by the indirect method, which is the comparison of the activity of each fission product in coated particles with the decay curve. Fig. 6 shows the result of heating at 22OO’C up to 180 min. Dotted lines indicate decay curves. Activities of lo3Ru and 99Mo that have
4. Mathematical analysis and discussion 4.1. Mathematical analysis The mathematical analysis of the fission product release from coated fuel particles has been under-
Y. Kumta et al. /Fission product release from UOz particles
111
where C,(r, t) is the concentration of fusion products in the medium i, D1 and &(r) are the diffusion coefficient and the birth rate in the medium i, respectively, r is the radial distance in the coated particle, h is the decay constant and t is the time. Numerical calculation for unsteady diffusion in multi-layers are done in the code FECUND by means of the Crank-Nicolson method. B,(r) is zero in the case of post-irradiation heating. Since the heating time was short, X could be also neglected. Then, eq. (1) becomes
ac,(r,r)=D a2Ci(r, t) 2 aCAr, f) at
i
(
ar2
+~~
1.
(2)
The initial condition is
/:[i-i/ 1
2 3 4 Time (day)
5
6
7
Fig. 7. Activity of each ftion product in coated fuel particlesand heatinghistory. Dotted linesindicatedecay curves.
taken by several workers [8-l 31. Walther [9] treated the steady state fractional release of fission products from coated particles. The mathematical expressions by Baurmann [ 121 and Rosenberg et al. [ 131 were for the particles with single-layered coating. These mathematical analyses are insufficient for the fission product release from TRISOcoated particles used in this experiment. The code FECUND [14] (Fission product Evaluation Code for UNsteady Diffusion) is used for the evaluation of fission product release from coated particle fuels at Japan Atomic Energy Research Institute. It is assumed that the fmion product release from intact coated particles is due to diffusion that obeys Fick’s law. The following equation is obtained for the migration of ftion products in a spherical coated particle:
aw
0=
(
aV,(r, t)
at
I
- WC
0 + Bf(d ,
D
ar2
2 ackr, t)
+TT
) (1)
where C is the initial concentration, II is the radius of the kernel and b is the radius of the coated particle. As for the boundary conditions, the partition coefficients at the interface between the layers are postulated to be 1.Oand the evaporation coefficients at the surface of the outermost layer are infinite. The evaporation coefficients can not be fed into the code FECUND now. Although these two assumptions are not correct actually, they are made so that the mathematical analysis of fusion product release is carried out with diffusion coefficients in each mediium as parameters. As the diffusion coefficients of fission products have not been known at this high temperature, the effective diffusion coefficients for the calculation by the code FECUND were set arbitrarily in accordance with a following line referring to data obtained from experiments at lower temperatures [l&18]. If D1, D2, D3, D4 and Ds denote effective diffusion coefficients in UO1 kernel, buffer layer, inner PyC, Sic and outer PyC, respectively: (1)D2>Ds=Ds>D1>D4, (2) the effective diffusion coefficients in each medium increase with temperature, (3) the nuclide which has a high fractional release has high effective diffusion coefficients in each medium (4) D2 = lODs, and (5) D4 becomes larger than D, or D, at 232O’C. Fractional release was computed by the data from the code FECUND feeding these effective diffusion
Y. Kurata et al. /Fission product release from lJO2 particles
-Q-
Enpwimantol Qto
-
Caladatim
S 0
Heating time (min) Fig. 8. Fraction of 9sNb released from TRISOcoated particles versus heating time.
Calculation
’ I ’ 100 200
8
’ 300
I
’ 400
t-krting time (mid fuel
coefficients. Then, data fitting was made by repeating the calculation with various effective diffusion coefficients. The results for “Nb are shown in fig. 8, for 99Mo in fig. 9 and for lo3Ru in fig. 10. Input data of effective diffusion coefficients for the code FECUND are shown in table 2. Good agreement is seen between calculation and experiment at low temperatures. When temperatures and the fractional release become high, experimental data deviate from calculated results. Fig. 11 shows the results for r4’ Ce, ’ 3’Cs and 140Ba at 232OOC.The agreement between the experimental data and the calculated result for 141Ce is good. But it was difficult to fit the experimental data to a single calculated curve in the case of “‘Cs and 140Ba Experimental data at short heating time for 13’Cs ;ie on the calculated curve E-l and at long heating time on E-2. The difference of input data between E-l and E-2 is only that the effective diffusion coefficient in Sic (Da) is about sixty times higher in E-2 than in E-l as shown in table 2. The experimental data for 140Ba are between the calculated curve F-l
Fig. 9. Fraction of 99Mo released from TRISOcoated particles versus heating time.
fuel
and F-2. These results indicate that diffusion coeffrcients of fission products in Sic become high at halfway time of heating at 232O’C and this implies the rapid change of the structure of Sic. Diffusion in Sic is a ratedetermining step in the case of lroBa and 13’Cs, but not in the case of 141Ceat 232O’C. 4.2. Discussion Although ‘03Ru, 99Mo and 95Nb have not been considered to belong to the fission product group which cause contamination of the cooling circuit in HTGR at normal operating temperatures [15], the fractional release of these nuclides is high and they have large effective diffusion coefficients at this high temperature. Ruthenium is present in the metallic state in the oxide fuel [20]. Since the Gibbs energy of formation of the oxides Moo2 and Nb20s is much higher than that of uranium oxide [20], molybdenum and niobium will be present as metals. Therefore these nuclides will be able to migrate into the coating
113
Y. Kurata et al. / Fission product release from UO? particles
0
20
40
60
a0
loo
120
Heating time (min)
lb’
’
0
’
-Q-
Ex~lmalltal
-
culalbtlon ’
too
’ 200
’
’ 300
data
’
Fig. 11. Fractional release of 141Ce, lJ7Cs and 140Bafrom TRISO-coated fuel particle at 2320°C.
’ 400
Heating time (min)
Fig. 10. Fraction of to3Ru released from TRISO-coated fuel particles versus heating time.
rapidly at high temperatures. Then ro3Ru and 9sNb are transported more easily in PyC than 95Zr and 14’Ce [15]. As the diffusion coefficient of lo3Ru in SIC, which is the barrier for solid fission products, is fairly large [18,19], those of 99Mo and 9sNb will be high at this high temperature. These nuclides are considered to have high diffusion coefficients in UOs, PyC and Sic at this high temperature as shown in the mathematical analysis. It was reported that 55% of the cesium inventory was released at 2OOO’Cwith 5 h from intact BISOcoated particle fuels and barium behaved qualitatively similar to cesium [3]. On the other hand, 14’Ba and “‘Cs are not released at 2050°C and begin to be released a little at 2200°C from TRISOcoated particles in our experiment. TRISOcoating with Sic is considered to be effective for the retention of 14’Ba and 13’Cs at 2000°C from these results. Fukuda et al. [17] reported the release of 14’Ba from SiCcoated fuel particles with UC2 kernel and diffusion coeffi-
cients of 14’Ba in Sic at 1650 to 185O’C. Barium is one of the elements which form oxides in UOs kernel of the coated fuel particles. BaO is detected at the grain boundaries and in inclusions in the fuel matrix and is retained well in the inclusions (20,211. It might be due to the difference in kernel compositions or characters of Sic that barium is not released up to 205O’C in our experiment. The data for fractional release of 14’Ba and r3’Cs at 2320°C are between the two calculated curves as shown in the mathematical analysis. Cesium is mobile in UOs [ 151 and barium will not be retained in UOs at this temperature. Then these nuclides are transported easily in PyC [ 15). They reach the Sic layer at short heating time. Then they are transported in Sic slowly first and rapidly at halfway time of heating at 2320°C. This indicates that the diffusion behaviour in Sic changes rapidly in heating at 2320°C. As it was found recently that the structure of Sic changed rapidly and the transition of beta to alpha Sic occurred rapidly at 2320°C [7], the rapid change of the effective diffusion coefficients in Sic is considered to be due to proceeding of the transition of beta to alpha Sic at this temperature. 141Ce 14’La and 95Zr belong to the most immobile fission products in coated particles with oxide kernels at normal operating temperatures [ 151 and
114
Y. Kurata et al. / Fission product release from UOz particles
Table 2 Input data of effective diffusion coefficients for calculatingfissionproduct releasefrom TRISOcoated fuel particles by the code FECUND D1
D2
03
04
DS
A-l 2 3 4
3.5 x 1.2x 6.0 x 2.0 x
lo-‘3 10-12 lo-l2 10-t l
7.0 x 3.0 x 1.6x 5.0 x
lo-‘2 lo-” lo-” lo-to
7.0 x 3.0x 1.6x 5.0x
10-13 10-12 lo-” lo-”
3.0 x 2.0x 2.0 x 3.0 x
10-14 10-13 lo-t2 lo-”
7.0 X 3.0X 1.6 X 5.0 x
lo-l3 lo-l2 lo-” lo-”
B-l 2 3 4
5.0 x 1.8x 9.0 x 3.0 x
10-13 lo-l2 10-12 lo-”
1.0 x 3.5 x 1.9x 6.0 x
lo-” lo-” lo-to lo-lo
1.0x 3.5 x 1.9x 6.0 x
10-12 10-12 lo-” 10-l 1
5.0 3.5 3.5 5.0
x x x x
10-14 lo-‘3 10-12 10-t 1
1.0x 3.5 x 1.9x 6.0 x
10-12 10-12 lo-” lo-”
C-l 2 3 4
8.0 x 8.0x 4.0 x 1.0x
lo-l3 lo-l2 lo-” lo-to
2.0x 1.6 x 8.0x 2.1 x
lo-‘1 lo-lo lo-lo 10-9
2.0 x 1.6 X 8.0x 2.1 x
10-12 10-l’ lo-” lo-to
1.2x 2.0x 1.0 x 1.5 x
10-13 lo-‘2 lo-” lo-”
2.0 1.6 8.0 2.1
10-12 lo-” 10-l’ lo-to
lo-l3
x X X x
D-l
8.0x
1.4 x lo-”
1.4 x 10-12
1.1 X lo-”
1.4 x 10-12
E-l 2
1.5 x lo-to 1.5 x lo-to
5.0 x lo-to 5.0 x lo-to
5.0x lo-” 5.0 x lo-”
1.4 X lo-l2 8.5 x lo-”
5.0 x lo-” 5.0 x lo-”
F-l 2
2.0 x lo-to 2.0 x lo-to
1.0 x 10-a 1.0 x 10-9
1.0 x lo-to 1.0 x 10-10
3.0 x 10-12 1.6 x 10-10
1.0 X 10-10 1.0 X lo-to
Dl, D2, D3.04 respectively.
and D5 denote effective diffusioncoefficients (m2/s) in UO2 kernel, buffer layer, inner PyC, SIC and outer PyC,
are released little if any at 232O’C. These nuclides form thermodynamically stable oxides in U02 kernels [20]. Then they are transported more slowly in PyC than 14’Ba, ’ “Cs and lo3Ru [ 151. Since Sic does not retain Won products at 232O’C as mentioned above, diffusion in U02 or PyC is the ratedetermining step for the release of 14’Ce, 14’L.a and “Zr. As rare gas nuclides and iodine were also retained in the BISOcoated particles at this high temperature as long as the coating remained intact [3], it is suggested that PyC holds the.gas tightness and retains 13jXe, 1311, 1321and 132Te. Effective diffusion coefficients in U02, PyC and Sic (Dl, D3 or Ds and D4, respectively) used for calculation at 1940 to 232O’C are between lo-l4 and lo-” (m2/s). But these values are much higher than the diffusion coefficients obtained from extrapolation of data in refs. [15-181 to this high temperature. It is seen in ref. [22] that the experimental data at 2000°C deviate from the calculated results by extrapolation. The change in characters of the medium where fission products diffuse might occur at this high temperature. It is difficult to discuss the abso-
lute value of the effective diffusion coefficients used for calculating the fission product release by the code FECUND now, and further studies are necessary to measure true diffusion coefficients in each medium and elucidate the mechanism of diffusion,
5. conclusions
From the experiment on fmion product release from TRISOcoated particles by post-activation heating at 1940 to 2320°C, the following is concluded: (1) lo3Ru, 99Mo and 9sNb are released at 1940 to 232O’C. The fractional release of these nuclides is very high at 232O’C. They have high effective diffusion coefficients in coated particles at this high temperature . (2) 14’Ba and 13’Cs are retained in TRISOcoated particles at 2050°C. These nuclides begin to be released at 22OO’C and are released rapidly at 232O’C. The rapid release at 2320°C is considered to be due to proceeding of the transition of beta to alpha Sic.
Y. !&rata et al. f R&m
product releasefromi.&
(3) 14’Ce, 140La and “Zr are released little if any at 2320°C. Diffusion in UO? or PyC is considered to be the ratedetermining step for the release of these nuclides at this high temperature, (4) Rare gas r-&ides, iodine and te~u~urn seem to be retained in coated particles at this high temperature as long as the coating does not fail. Acknowledgements The authors wish to thank Dr. J. Shimokawa, Head of the Division of Nuclear Fuel Research for his interest and encouragement, and Dr. S. Mitake for his permission to use the code FECUND.
[l] T. Aochi, J. Shimokawaet al., 1AERl-M 6895 (1976) p. 157. [2] A. Naoumidis, &L-l465 (1977). [3] W. Schenh and A. Naoumidis, Nucl. Technol. 46 {1979) 228. [4] K. Petersen, H. Barthels, H.E. Drescher et al., Nucl. Technol. 46 (1979) 306.
particies
115
(51 P. Soo, G. Uneberg, B. Rocks et al., BNL-NUREG50931 (1978) p. 10. [6] K. Ikawa, F. Kobayashi and K. Iwamoto, J. Nucl. $5. Technol. 15 (1978) 774. [7) Y. Kurata, K. lkawa and K. Iwamoto, J. Nucl. Mater. 92 (1980) 351. [8] R.W. Dunlap and T.D. Gulden, Nucl. Sci. Eng. 32 (1.968) 407. 191 H. Walthex, Nukleonih ll(l968) 171. (lo] R.B. Evans III, J.O. Stiegler and J. Truitt, ORNL-3711 (1964). [ll] P.D. Smith, R.G. Steinhe, D.D. Jensen and T. Hama, Nucl. Technol. 35 (1977) 475. [ 121 K.W. Baurmann, JoL68SPA (1970). [ 131 H.S. Rosenberg, D.L. Morrison, C.W. TownIey and D.N. Sunderman, BMI-1734 (1964). [14] M. Mahino and S. Yasuhawa, JAERI-M 4883 (1972). (15 ] J. En@hard, JijL-752-RG (1971). (161 P.E. Brown and R.L. Faircloth, J. Nucl. Mater. 59 (1976) 21. [17] K. Fukuda and K. Iwamoto, J. Nucl. Sci. Technol. 12 (1975) 181. [ 181 JAERI internal document on research and development of multi-purpose VHTR (1978). [19] D.R. Boyle, F.L. Brown and J.E. !Sanedri, J. Nucl. Mater. 29 (1969) 27. [20] BY. Jeffrey, J. Nucl. Mater. 22 (1967) 33. [21] F.E. Van&lager, WE. Bell, 0. Siaman and T. Morgan, GA-10073 (1970).
107
FISSION PRODUCT RELEASE FROM TRISO-COATED UO? PARTICLES AT 1940 TO 2320°C
Yuji KURATA, Katsuichi IKAWA and Kazurnl IWAMOTO Japan Atomic Energy Research Institute, Tok-ai-mura,Naka-gun, Ibaraki-ken, 319-l 1 Japan Received 16 December 1980
The fiiion product release from TRISO-coated UOa particles was measured by postactivation heating at 1940 to 2320°C for use in a safety analysis. The results are analyzed mathematically with effective diffusion coefficients in each medium. lo3Ru, ggMo and gsNb are released at 1940 to 2320°C and have high effective diffusion coeffldents. Although 140Ba and 137Cs are retained in TRJS0coate.d particies at 205O”C, they are released rapidly at 2320°C. This is attriiuted to the transition of beta to alpha SE at 2320°C. r4tCe, 140La and gs2r are released little if any at 2320°C. Rare gas nuclides, iodine and tellurium seem to be retained in coated particles at this high temperature.
1. Introduction It has been shown by the results of irradiation tests that coated fuel particles for the high-temperature gas-cooled reactor (HTGR) behave fairly well under normal reactor conditions. According to a safety analysis [l] of the experimental HTGR, there is a possibility of fuel temperature excursion exceeding 2OOO’C under hypothetical accident conditions such as blocking of coolant channels or a reactivity accident. There were several studies on the behaviour of coated fuel particles during a thermal excursion above 2OOO’C. Some data on failure behaviour of coated particles are already available in the literature [2-S]. Most of them are of BISO-coated particles. Data on the failure of TRISOcoated particles with a Sic layer for use in the experimental HTGR being developed in Japan are also reported [6]. As recently reported [7], heat treatment above 2OOO’Chas an effect on density and microstructure of Sic of the coated fuel particles. For a safety analysis, it is essential to evaluate the amount of fission products released from fuels. Many works have been made on fusion product release at temperatures under normal operating conditions of HTGR. But few works have been made at temperatures under hypothetical accident conditions. Schenk and Naoumidis [3] reported the fission product release in HTGR fuel above 2OOO’Cunder the condi-
0 022-3 115/8 1/OOOO-0000/$02.50 0 North-Holland
tions of unrestricted core heatup events. Their fuels were spherical fuel elements that had uranium-thorium mixed oxide particles with a pyrocarbon coating deposited from methane (I-III-BISO). It is expected that the behaviour of fwion product release in TRISOcoated particles with UOa kernel differs frorr that in BISOcoated particles. An experiment of post-activation heating above 2OOO’Cwas carried out to examine the fission product release from slightly-irradiated TRISOcoated fuel particles. Then the results were analyzed mathematically with effective diffusion coefficients in UOa, pyrolytic carbon and Sic as parameters.
2. Experimental The samples used were the same TRISOcoated particles as in the previous experiment [7]. The materials, dimensions and densities of the components of the particles are listed in table 1. The innermost layer is the buffer layer deposited from acetylene, the second and the outermost layers LTI PyC (low temperature isotropic pyrolytic carbon) from propylene and the third layer Sic from methyltrichlorosilane. The sample particles were irradiated at ambient temperature (approximately 8O’C) to a thermal neutron dose of 6 X 1016 or 6 X 10” n/cm?. After irradiation, the particles were allowed to cool for a fured period.
Y. Kurata et al. /Fission product release porn UOz particles
108
Table 1 TRISO particle parameters Component
Material
Thickness OImf
Density Q./cm31
Fuel kernel Innermost layer Second layer Third layer Outermost layer
uoz PyC
600 w 60 30 25 45
10.53 1.10 1.85 >3.20 1.85
PYC
sic P,C
Q,= diameter.
The period was one week or more for measuring the short-lived fusion products such as “MO, 133Te, t3sI, 133Xe, “‘Ba and “‘Ru, two or three months for 95Zr, “Nb and “Ice and about ten moths for 13’Cs. Following this, the intact particles were dispersed in graphite powders, placed into a graphite crucible and heated in the apparatus shown in fig. 1. The crucible was evacuated to 10m5bar. Solid fission products were absorbed on the graphite components, which were composed of graphite powders, a graphite crucible, a graphite supporter and a graphite felt sleeve as shown in fig. 1. Gaseous ftion products were monitored by providing a charcoal trap cooled with dry ice at the outlet of the vacuum pump. As the sample particles failed in heating at 2450 to 2500°C for 15 to 30 min in this apparatus, post-activation heating was made between 1940 and 2320%
Temperatures were measured by an optical pyrometer through a hole in the lid of the graphite crucible and corrected for the absorption of lens, prism and quartz plate. The sample particles were assayed by gammaray spectrometry before and after heating. The graphite components and the charcoal trap that had absorbed fission products released from the sample particles were also assayed similarly. Fractional release of each fusion product was calculated from these data. Most of solid fmion products were absorbed on the graphite components. But some fission products, for example “‘Cs and to3Ru, were found on the graphite wool or zirconia when the fractional release became high. In this case, the fractional release was calculated from the remainder of activities counted before and after heating. It was difficult to measure the fractional release of rare gas nuclides and iodine directly in this experiment, because an effective collector for these gaseous nuclides was not prepared in this apparatus. Therefore, the activity of each fission product in coated particles was measured before and after heating and compared with a decay curve of each fission product in the experiment of short-lived fission product release.
to-’
I
0
’
too
I
’
200
Heating
time
I
300
’
I
400
(min)
Fig. 2. Fractional release of fission products from TRISOFig. 1. Apparatus for postrctivation heating of coated fuel IXUtiCleS. coated fuel particles at 1940°C.
Y. Kurata et al. /Fission
productrelease from UOspurticles
109
3. Results 3.1. Release of solid fission products
Fractional release of fission products from the sample particles was measured at 1940, 2050, 2200 and 2320°C. Heating at 1940 and 2050°C was made up to 400 min. The results are shown in figs. 2 and 3. Only release of lo3Ru, 99Mo and “Nb of the solid fission products is detected. The rate of increase in the fractional release tends to decrease at 205O’C with heating time. Fig. 4 shows the fractional release of fission products at 2200°C up to 180 min. The fractional release of rsoBa is measured at 2200°C in addition to that of ro3Ru, 99Mo and “Nb. The fractional release of r3’Cs in heating at 2200°C for 180 min is almost the detection limit. The release of “‘Cs which is one of the important nuclides from the pbint of view of a safety analysis, is kept low at 2200°C. Fractional release at 2320°C up to 100 min is shown in fig. 5 similarly. The fractional release
‘7’
Heating
0
50
100
150
200
l-taoting time (min) Fig. 4. Fractional release of fdn coated fuel particles at 2200°C.
time(tin)
Fig. 3. Fractional release of fission products from TRlSOcoated fuel particles at 2050°C.
products
from TRlSO-
of 13’Cs and r4rCe is measured in addition to that measured at 2200°C. The rate of increase in the fractional release of ro3Ru, 99Mo and “Nb at 232O’C decreases with heating time in the same way as that measured below 2200°C. But it is noticed that the fractional release of r4’Ba and “‘Cs increases rapidly at about 30 min in heating at 232O’C. This implies that a change of the diffusion behaviour of 14’Ba and 13’Cs in coated particles occurs at 232O’C. 9sNb was observed in the graphite components that absorbed fission products, but it was not 9sZr, which is a parent nuclide of 95Nb. It is found that 95Nb is released from coated particles but 95Zr is retained in coated particles. Both 14’Ba and r40L.a were observed in the graphite components. Since r4’La is a daughter nuclide of 14’Ba, the radioactive equilibrium of r40Ba-r40La was calculated. The result indicates that “‘La found in the graphite components almost depends on lroBa released from
110
Y. Kumta et al. /Fission product release from U02 prrrticles
I Time
(day)
a 0
20
40
60
80
loo
420
Heating time (min) Fig. 5. Fractional release of fission products from TRISOcoated fuel particles at 2320°C.
coated particles and that the release of 14’La is not significant. According to these experimental results, solid fission products released at this high temperature are classified into the following groups: (1) 103Ru, 99Mo and 95Nb that have high fractional release at 1940 to 2320°C, (2) 14’Ba and 13’Cs that begin to be released at 22OO’Cand are released rapidly at 2320°C, and (3) 141Ce, 140La and 9sZr that are released little if any at 232O’C.
Fig. 6. Activity of each fission product in coated fuel particles and heating history. Dotted lines indicate decay curves.
high fractional release deviate from each decay curve when the coated particles are heated and fission products are released. But the curves of activities of these nuclides versus time are parallel to each decay curve as shown in fig. 6 while the coated particles are not heated. Activities of “‘Xe, “‘1 and 132Te measured after heating lie on each decay curve. Similarly it is shown in fig. 7 that the activities of ‘O’Ru, 99Mo and 14’Ba are apart from each decay curve but those of “‘Xe, 1311and 132Te he on each decay curve in heating at 2320°C up to 100 min. Then neither lJ3Xe, 1311, 132I nor 132Te was observed on the charcoal trap and graphite components. These results indicate that rare gas nuclides and iodine seem to be retained in coated particles even at this high temperature.
3.2. Release of gaseous fission products Release of gaseous fission products was examined by the indirect method, which is the comparison of the activity of each fission product in coated particles with the decay curve. Fig. 6 shows the result of heating at 22OO’C up to 180 min. Dotted lines indicate decay curves. Activities of lo3Ru and 99Mo that have
4. Mathematical analysis and discussion 4.1. Mathematical analysis The mathematical analysis of the fission product release from coated fuel particles has been under-
Y. Kumta et al. /Fission product release from UOz particles
111
where C,(r, t) is the concentration of fusion products in the medium i, D1 and &(r) are the diffusion coefficient and the birth rate in the medium i, respectively, r is the radial distance in the coated particle, h is the decay constant and t is the time. Numerical calculation for unsteady diffusion in multi-layers are done in the code FECUND by means of the Crank-Nicolson method. B,(r) is zero in the case of post-irradiation heating. Since the heating time was short, X could be also neglected. Then, eq. (1) becomes
ac,(r,r)=D a2Ci(r, t) 2 aCAr, f) at
i
(
ar2
+~~
1.
(2)
The initial condition is
/:[i-i/ 1
2 3 4 Time (day)
5
6
7
Fig. 7. Activity of each ftion product in coated fuel particlesand heatinghistory. Dotted linesindicatedecay curves.
taken by several workers [8-l 31. Walther [9] treated the steady state fractional release of fission products from coated particles. The mathematical expressions by Baurmann [ 121 and Rosenberg et al. [ 131 were for the particles with single-layered coating. These mathematical analyses are insufficient for the fission product release from TRISOcoated particles used in this experiment. The code FECUND [14] (Fission product Evaluation Code for UNsteady Diffusion) is used for the evaluation of fission product release from coated particle fuels at Japan Atomic Energy Research Institute. It is assumed that the fmion product release from intact coated particles is due to diffusion that obeys Fick’s law. The following equation is obtained for the migration of ftion products in a spherical coated particle:
aw
0=
(
aV,(r, t)
at
I
- WC
0 + Bf(d ,
D
ar2
2 ackr, t)
+TT
) (1)
where C is the initial concentration, II is the radius of the kernel and b is the radius of the coated particle. As for the boundary conditions, the partition coefficients at the interface between the layers are postulated to be 1.Oand the evaporation coefficients at the surface of the outermost layer are infinite. The evaporation coefficients can not be fed into the code FECUND now. Although these two assumptions are not correct actually, they are made so that the mathematical analysis of fusion product release is carried out with diffusion coefficients in each mediium as parameters. As the diffusion coefficients of fission products have not been known at this high temperature, the effective diffusion coefficients for the calculation by the code FECUND were set arbitrarily in accordance with a following line referring to data obtained from experiments at lower temperatures [l&18]. If D1, D2, D3, D4 and Ds denote effective diffusion coefficients in UO1 kernel, buffer layer, inner PyC, Sic and outer PyC, respectively: (1)D2>Ds=Ds>D1>D4, (2) the effective diffusion coefficients in each medium increase with temperature, (3) the nuclide which has a high fractional release has high effective diffusion coefficients in each medium (4) D2 = lODs, and (5) D4 becomes larger than D, or D, at 232O’C. Fractional release was computed by the data from the code FECUND feeding these effective diffusion
Y. Kurata et al. /Fission product release from lJO2 particles
-Q-
Enpwimantol Qto
-
Caladatim
S 0
Heating time (min) Fig. 8. Fraction of 9sNb released from TRISOcoated particles versus heating time.
Calculation
’ I ’ 100 200
8
’ 300
I
’ 400
t-krting time (mid fuel
coefficients. Then, data fitting was made by repeating the calculation with various effective diffusion coefficients. The results for “Nb are shown in fig. 8, for 99Mo in fig. 9 and for lo3Ru in fig. 10. Input data of effective diffusion coefficients for the code FECUND are shown in table 2. Good agreement is seen between calculation and experiment at low temperatures. When temperatures and the fractional release become high, experimental data deviate from calculated results. Fig. 11 shows the results for r4’ Ce, ’ 3’Cs and 140Ba at 232OOC.The agreement between the experimental data and the calculated result for 141Ce is good. But it was difficult to fit the experimental data to a single calculated curve in the case of “‘Cs and 140Ba Experimental data at short heating time for 13’Cs ;ie on the calculated curve E-l and at long heating time on E-2. The difference of input data between E-l and E-2 is only that the effective diffusion coefficient in Sic (Da) is about sixty times higher in E-2 than in E-l as shown in table 2. The experimental data for 140Ba are between the calculated curve F-l
Fig. 9. Fraction of 99Mo released from TRISOcoated particles versus heating time.
fuel
and F-2. These results indicate that diffusion coeffrcients of fission products in Sic become high at halfway time of heating at 232O’C and this implies the rapid change of the structure of Sic. Diffusion in Sic is a ratedetermining step in the case of lroBa and 13’Cs, but not in the case of 141Ceat 232O’C. 4.2. Discussion Although ‘03Ru, 99Mo and 95Nb have not been considered to belong to the fission product group which cause contamination of the cooling circuit in HTGR at normal operating temperatures [15], the fractional release of these nuclides is high and they have large effective diffusion coefficients at this high temperature. Ruthenium is present in the metallic state in the oxide fuel [20]. Since the Gibbs energy of formation of the oxides Moo2 and Nb20s is much higher than that of uranium oxide [20], molybdenum and niobium will be present as metals. Therefore these nuclides will be able to migrate into the coating
113
Y. Kurata et al. / Fission product release from UO? particles
0
20
40
60
a0
loo
120
Heating time (min)
lb’
’
0
’
-Q-
Ex~lmalltal
-
culalbtlon ’
too
’ 200
’
’ 300
data
’
Fig. 11. Fractional release of 141Ce, lJ7Cs and 140Bafrom TRISO-coated fuel particle at 2320°C.
’ 400
Heating time (min)
Fig. 10. Fraction of to3Ru released from TRISO-coated fuel particles versus heating time.
rapidly at high temperatures. Then ro3Ru and 9sNb are transported more easily in PyC than 95Zr and 14’Ce [15]. As the diffusion coefficient of lo3Ru in SIC, which is the barrier for solid fission products, is fairly large [18,19], those of 99Mo and 9sNb will be high at this high temperature. These nuclides are considered to have high diffusion coefficients in UOs, PyC and Sic at this high temperature as shown in the mathematical analysis. It was reported that 55% of the cesium inventory was released at 2OOO’Cwith 5 h from intact BISOcoated particle fuels and barium behaved qualitatively similar to cesium [3]. On the other hand, 14’Ba and “‘Cs are not released at 2050°C and begin to be released a little at 2200°C from TRISOcoated particles in our experiment. TRISOcoating with Sic is considered to be effective for the retention of 14’Ba and 13’Cs at 2000°C from these results. Fukuda et al. [17] reported the release of 14’Ba from SiCcoated fuel particles with UC2 kernel and diffusion coeffi-
cients of 14’Ba in Sic at 1650 to 185O’C. Barium is one of the elements which form oxides in UOs kernel of the coated fuel particles. BaO is detected at the grain boundaries and in inclusions in the fuel matrix and is retained well in the inclusions (20,211. It might be due to the difference in kernel compositions or characters of Sic that barium is not released up to 205O’C in our experiment. The data for fractional release of 14’Ba and r3’Cs at 2320°C are between the two calculated curves as shown in the mathematical analysis. Cesium is mobile in UOs [ 151 and barium will not be retained in UOs at this temperature. Then these nuclides are transported easily in PyC [ 15). They reach the Sic layer at short heating time. Then they are transported in Sic slowly first and rapidly at halfway time of heating at 2320°C. This indicates that the diffusion behaviour in Sic changes rapidly in heating at 2320°C. As it was found recently that the structure of Sic changed rapidly and the transition of beta to alpha Sic occurred rapidly at 2320°C [7], the rapid change of the effective diffusion coefficients in Sic is considered to be due to proceeding of the transition of beta to alpha Sic at this temperature. 141Ce 14’La and 95Zr belong to the most immobile fission products in coated particles with oxide kernels at normal operating temperatures [ 151 and
114
Y. Kurata et al. / Fission product release from UOz particles
Table 2 Input data of effective diffusion coefficients for calculatingfissionproduct releasefrom TRISOcoated fuel particles by the code FECUND D1
D2
03
04
DS
A-l 2 3 4
3.5 x 1.2x 6.0 x 2.0 x
lo-‘3 10-12 lo-l2 10-t l
7.0 x 3.0 x 1.6x 5.0 x
lo-‘2 lo-” lo-” lo-to
7.0 x 3.0x 1.6x 5.0x
10-13 10-12 lo-” lo-”
3.0 x 2.0x 2.0 x 3.0 x
10-14 10-13 lo-t2 lo-”
7.0 X 3.0X 1.6 X 5.0 x
lo-l3 lo-l2 lo-” lo-”
B-l 2 3 4
5.0 x 1.8x 9.0 x 3.0 x
10-13 lo-l2 10-12 lo-”
1.0 x 3.5 x 1.9x 6.0 x
lo-” lo-” lo-to lo-lo
1.0x 3.5 x 1.9x 6.0 x
10-12 10-12 lo-” 10-l 1
5.0 3.5 3.5 5.0
x x x x
10-14 lo-‘3 10-12 10-t 1
1.0x 3.5 x 1.9x 6.0 x
10-12 10-12 lo-” lo-”
C-l 2 3 4
8.0 x 8.0x 4.0 x 1.0x
lo-l3 lo-l2 lo-” lo-to
2.0x 1.6 x 8.0x 2.1 x
lo-‘1 lo-lo lo-lo 10-9
2.0 x 1.6 X 8.0x 2.1 x
10-12 10-l’ lo-” lo-to
1.2x 2.0x 1.0 x 1.5 x
10-13 lo-‘2 lo-” lo-”
2.0 1.6 8.0 2.1
10-12 lo-” 10-l’ lo-to
lo-l3
x X X x
D-l
8.0x
1.4 x lo-”
1.4 x 10-12
1.1 X lo-”
1.4 x 10-12
E-l 2
1.5 x lo-to 1.5 x lo-to
5.0 x lo-to 5.0 x lo-to
5.0x lo-” 5.0 x lo-”
1.4 X lo-l2 8.5 x lo-”
5.0 x lo-” 5.0 x lo-”
F-l 2
2.0 x lo-to 2.0 x lo-to
1.0 x 10-a 1.0 x 10-9
1.0 x lo-to 1.0 x 10-10
3.0 x 10-12 1.6 x 10-10
1.0 X 10-10 1.0 X lo-to
Dl, D2, D3.04 respectively.
and D5 denote effective diffusioncoefficients (m2/s) in UO2 kernel, buffer layer, inner PyC, SIC and outer PyC,
are released little if any at 232O’C. These nuclides form thermodynamically stable oxides in U02 kernels [20]. Then they are transported more slowly in PyC than 14’Ba, ’ “Cs and lo3Ru [ 151. Since Sic does not retain Won products at 232O’C as mentioned above, diffusion in U02 or PyC is the ratedetermining step for the release of 14’Ce, 14’L.a and “Zr. As rare gas nuclides and iodine were also retained in the BISOcoated particles at this high temperature as long as the coating remained intact [3], it is suggested that PyC holds the.gas tightness and retains 13jXe, 1311, 1321and 132Te. Effective diffusion coefficients in U02, PyC and Sic (Dl, D3 or Ds and D4, respectively) used for calculation at 1940 to 232O’C are between lo-l4 and lo-” (m2/s). But these values are much higher than the diffusion coefficients obtained from extrapolation of data in refs. [15-181 to this high temperature. It is seen in ref. [22] that the experimental data at 2000°C deviate from the calculated results by extrapolation. The change in characters of the medium where fission products diffuse might occur at this high temperature. It is difficult to discuss the abso-
lute value of the effective diffusion coefficients used for calculating the fission product release by the code FECUND now, and further studies are necessary to measure true diffusion coefficients in each medium and elucidate the mechanism of diffusion,
5. conclusions
From the experiment on fmion product release from TRISOcoated particles by post-activation heating at 1940 to 2320°C, the following is concluded: (1) lo3Ru, 99Mo and 9sNb are released at 1940 to 232O’C. The fractional release of these nuclides is very high at 232O’C. They have high effective diffusion coefficients in coated particles at this high temperature . (2) 14’Ba and 13’Cs are retained in TRISOcoated particles at 2050°C. These nuclides begin to be released at 22OO’C and are released rapidly at 232O’C. The rapid release at 2320°C is considered to be due to proceeding of the transition of beta to alpha Sic.
Y. !&rata et al. f R&m
product releasefromi.&
(3) 14’Ce, 140La and “Zr are released little if any at 2320°C. Diffusion in UO? or PyC is considered to be the ratedetermining step for the release of these nuclides at this high temperature, (4) Rare gas r-&ides, iodine and te~u~urn seem to be retained in coated particles at this high temperature as long as the coating does not fail. Acknowledgements The authors wish to thank Dr. J. Shimokawa, Head of the Division of Nuclear Fuel Research for his interest and encouragement, and Dr. S. Mitake for his permission to use the code FECUND.
[l] T. Aochi, J. Shimokawaet al., 1AERl-M 6895 (1976) p. 157. [2] A. Naoumidis, &L-l465 (1977). [3] W. Schenh and A. Naoumidis, Nucl. Technol. 46 {1979) 228. [4] K. Petersen, H. Barthels, H.E. Drescher et al., Nucl. Technol. 46 (1979) 306.
particies
115
(51 P. Soo, G. Uneberg, B. Rocks et al., BNL-NUREG50931 (1978) p. 10. [6] K. Ikawa, F. Kobayashi and K. Iwamoto, J. Nucl. $5. Technol. 15 (1978) 774. [7) Y. Kurata, K. lkawa and K. Iwamoto, J. Nucl. Mater. 92 (1980) 351. [8] R.W. Dunlap and T.D. Gulden, Nucl. Sci. Eng. 32 (1.968) 407. 191 H. Walthex, Nukleonih ll(l968) 171. (lo] R.B. Evans III, J.O. Stiegler and J. Truitt, ORNL-3711 (1964). [ll] P.D. Smith, R.G. Steinhe, D.D. Jensen and T. Hama, Nucl. Technol. 35 (1977) 475. [ 121 K.W. Baurmann, JoL68SPA (1970). [ 131 H.S. Rosenberg, D.L. Morrison, C.W. TownIey and D.N. Sunderman, BMI-1734 (1964). [14] M. Mahino and S. Yasuhawa, JAERI-M 4883 (1972). (15 ] J. En@hard, JijL-752-RG (1971). (161 P.E. Brown and R.L. Faircloth, J. Nucl. Mater. 59 (1976) 21. [17] K. Fukuda and K. Iwamoto, J. Nucl. Sci. Technol. 12 (1975) 181. [ 181 JAERI internal document on research and development of multi-purpose VHTR (1978). [19] D.R. Boyle, F.L. Brown and J.E. !Sanedri, J. Nucl. Mater. 29 (1969) 27. [20] BY. Jeffrey, J. Nucl. Mater. 22 (1967) 33. [21] F.E. Van&lager, WE. Bell, 0. Siaman and T. Morgan, GA-10073 (1970).