Evaluation of Carbon 14 Release from Irradiated Zircaloy Fuel Cladding Through aLong-Term Static Leaching Test
Hiroyoshi Ueda, Tomofumi Sakuragi, Naoki Fujii and Hitoshi Owada
Radioactive Waste Management Funding and Research Center (RWMC)
6-4 Akashicho, Chuo-ku, Tokyo, 104-0044, Japan
Key words; geological disposal, spent nuclear fuel, TRU waste, C-14, Zircaloy, corrosion

Abstract
Zircaloy cladding of the spent nuclear fuel, which is highly activated and contaminated, is expected to bedisposed of in a deep underground geological repository, both in the once-through and recycling fuel cycles.
Carbon 14 (C-14), mainly originated from the nitrogen impurity in Zircaloy cladding and formed through the14N(n, p)14C reaction in the reactor, is a typical activation product. In a preliminary performance assessment of
the deep geological disposal in Japan, C-14 gave a significant dose impact, which is due to the relatively large
inventory, relatively long half-life, i.e., 5730 years, higher release rate and the chemical speciation andconsequent migration parameters. In a preliminary Japanese safety case, relatively high IRF (instant releasefraction), 20%, was assumed regarding the oxide film formed on the metal surface as a source of the instantrelease.
With respect to this source term issue, we have prepared an irradiated BWR fuel cladding (Zircaloy-2, averagerod burnup of 41.6 GWd/t) which has an external oxide film of 25.3 μm thickness. Although the specific activity
in oxide was approximately 3 times higher than the base metal activity, due to the additional C-14 generation by
the 17O(n, α)14C reaction, the C-14 abundance in the oxide was less than 10% of total inventory. A static leaching
test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOHsolution (pH of 12.5) at room temperature. After 6.5 years, C-14 was found in each leachate fraction of gas phaseand dissolved organics and inorganics, the total of which was less than 0.01% of the entire C-14 inventory of theimmersed cladding tube. Both the C-14 abundance and the low leaching rate suggests that C-14 in oxide does
not have a significant impact on the IRF in the safety case.

1. IntroductionOne of the typical activation products in theradioactive waste from the nuclear fuel cycle isCarbon 14 (C-14) which is mainly generated by a
14N(n, p)14C reaction in the nuclear reactor. In apreliminary performance assessment of the deep
geological disposal in Japan [1], C-14 provided asignificant exposure dose due to the large inventory,
relatively long half-life, i.e., 5730 years, higher
release rate, and the speciation and consequentmigration parameters. The oxide film formed on theZircaloy cladding surface was regarded as a source ofinstant release while the release of radionuclides fromthe metal matrix was regarded as the corrosion-related
congruent release. The Zircaloy corrosion rate of 0.02
µm/y and the 20% IRF (instant release ratio) of thespent cladding waste were set as the source term
parameters in this preliminary safety case, in which
the speciation of C-14 was assumed to be an aqueousorganic.
Despite the presence of many studies on this issue,
the basic information is still sparse on radionuclide
inventory, release behavior of radionuclides andZircaloy corrosion. For this source term issue, wehave prepared an irradiated BWR fuel cladding(Zircaloy-2, average rod burnup of 41.6 GWd/tU)
which has an external oxide film with a thickness of
25.3 μm. The inventory of C-14 was determined by a
wet chemical treatment and liquid scintillation
counting of C-14 activity using different parts of thecladding such as the base Zircaloy metal, oxide, andmetal with and without oxide film. Then in the next
step, a static leaching test using cladding tube whoseinternal oxide was removed was carried out undersimulated geological disposal conditions for anexposure period of 6.5 years. This paper describes
some details of the experimental work shown in theprevious paper [2] on this study.
2. Measurements and tests(1) Specific radioactivity measurements of C-14
In advance to the leaching test, the specificradioactivity of C-14 was measured as the C-14
inventory in the claddings. The irradiated claddings
for this study were obtained from a BWR fuel rod
located at the circumference of the fuel assembly.
The fuel type is STEP I type, an older fuel type inJapan that uses an 8×8 array in a lattice configuration.

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

The average burnups are 39.4 GWd/tU for the fuelassembly and 41.6 GWd/tU for the fuel rod,respectively. The irradiation ended in March, 1990.
The specimen used in the present work was locatedbetween 582 mm and 882 mm from the rod top(Sample ID: F3GT3-H7-481). Table 1 summaries thecharacteristics and irradiation conditions of the fuelrod. The claddings had approximately 18 years ofcooling-off at the beginning of the leaching test.
Table 1Characteristics of irradiated cladding.

Reactor type

BWR

Fuel type

STEP I 8×8 array

Material

Zircaloy-2 (Zr liner)

Irradiation period

5 cyclesBurnup (Bundle ave., Rod ave.)39.4 GWd/tU, 41.6 GWd/tU

Cooling-off period

Approximately 18 years

After the fuel components were removed, thecladdings were washed with 6 M of HNO3 in a warm
bath and cut to an appropriate size. The thickness ofthe external oxide layer was measured, based on
optical microscope observations as shown in Fig. 1.
The thickness of the cladding specimen wasdetermined by the average value of measurements at
9 points in each of the 8 regions, i.e., at 72 points.

both internal and external oxide layers; (b) thecladding with the external oxide layer kept after
removing the internal oxide by grinding; (c) theZircaloy base metal without the internal and externaloxide layers which were removed by grinding; and(d) the fragment of the external oxide layer. The
external oxide layer of Zircaloy was peeled off
through vertical pressing of the columnar cladding(approximately 2 cm high) whose interior layer hadpreviously been removed. As a consequence,approximately 0.05 g of oxide was recovered as apowdery fragment.
(a)(b)(c)(d)

Fig. 2 External appearance of irradiated claddings for
C-14 measurement: (a) with internal andexternal oxide layers in place, (b) with internaloxide removed, (c) with internal and externaloxide removed, (d) (external) oxide fragment
removed from the cladding.

Fig. 1 Measurement of thickness of the external oxide
layers.

The C-14 radioactivity in the cladding specimens
was measured as shown in Fig. 3. First, the wholebody of the cladding or oxide sample was dissolvedin an HNO3 + HF solution and then heated. Carbon
components were evaporated by air bubbling for 30
min and oxidized by CuO in CO2 at 1073 K. Thematerial was then passed through three steps of adry-ice cold trap to remove tritium and three steps ofan alkaline trap (1 M of NaOH) to collect carbon
dioxide (14CO2). The potential contamination byradioiodine in a volatile state was removed by an AgI
precipitation. The C-14 radioactivity in the alkaline
traps was measured with a liquid scintillation counter(PerkinElmer Tri-Carb 2900TR).

Four types of samples, i.e., approximately 1 mm
thick, ring-shaped claddings and oxide fragment,
were obtained for specific activity measurements ofC-14, as shown in Fig. 2: (a) the cladding, which has

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Fig. 3 Measurement procedures of the C-14 specificradioactivity.

(2) Static leaching test
The static leaching test was carried out using theirradiated Zircaloy-2 cladding (F3GT3-H7-481),
fore-mentioned, of a 2 cm high cylinder (4.03 g) withan external oxide layer kept, but with its internaloxide layer removed. Final polishing of the cladding
interior was done with a 0.02 mm abrasive. A glassvial with a simple cap was filled with 20 ml of
diluted NaOH solution adjusted to a pH of 12.5. The
solution was deoxygenated before use to achieve aninitial ORP (oxidation-reduction potential) of −253
mV. The glass vial was placed on the inside of astainless steel container and sealed by gasket tomaintain the initial inert condition and to prevent
leakage of volatile C-14. This immersion procedure
was carried out inside a simple glove box with anitrogen atmosphere. The external view of the test
system is shown in Fig. 4.

After 6.5 years of immersion and being kept atroom temperature of approximately 293 K, theouter and inner containers were opened in a glove
bag with a vacuum tube attached. The gaseous
C-14 was collected through the tube by using thethree steps of an alkaline trap after passing throughan oxidation furnace. A portion of the liquid phase
was provided for measurement of total dissolved
C-14 and dissolved organic and inorganic C-14.
The measurement equipment of the dissolved C-14
is shown in Fig. 5. In the total C-14 measurement,
an aliquot of leachate was mixed with a carriercarbon (Na2CO3 and C2H5OH), an oxidant(potassium persulfate, K2S2O8), a catalyst (AgNO3),
and an acid (H2SO4) in a warm vessel. In theorganic/inorganic separation method, first theinorganic C-14 was fractionated from anotheraliquot by bubbling for 30 minutes underacidification. After that, the remaining organics
were oxidized with the same reagents used for totalC-14 measurement and the volatile C-14 (14CO2)
was collected by an alkaline trap in the samemanner as described for total C-14 measurement.
Leached C-14 in the alkaline traps was measured
by a liquid scintillation counter (PerkinElmer
Tri-Carb 3100TR).

Fig. 5 Measurement equipment of the dissolve C-14.

Fig. 4 External view of the test system for the leaching
test: the outer stainless steel container (left) andthe glass vial in which the cladding is immersed(right) in the photo.

3. Results and discussions(1) Distribution of C-14 in irradiated cladding
Table 2 shows the results of C-14 specific activity(Bq/g) measurements for the irradiated cladding
samples shown in Fig. 2: (a) the cladding withinternal and external oxide, (b) the cladding withexternal oxide, (c) the cladding base metal, i.e.,
Zircaloy-2, and (d) the external oxide. The measured
activities were normalized to the values at thebeginning of the leaching test in Table 2 because themeasurement was conducted through two years. For
the external oxide, units of grams in the specificactivity of oxide are converted into grams ofzirconium metal in order to facilitate comparison
with the base metal. The average values of (a), (b),6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan(c) and (d) are 1.49×104, 1.53×104, 1.49×104 and4.04×104 Bq/g, respectively.

The radionuclides in cladding oxide areregarded as a source of instant release in thepreliminary Japanese safety case, in which 20% ofC-14 in cladding was assumed as IRF and theremaining
80%
wasassumedas
thecorrosion-related congruent release from the base
Zircaloy metal [1]. In Table 2, the specific activity
between oxide and base Zircaloy differ by a factorof 3. The C-14 in Zircaloy is mainly generated
from the nitrogen impurity as a result of the14N(n,p)14C reaction. Given that the nitrogencontent in oxide and Zircaloy base metal are equalin amount, the activation calculation using theORIGEN code suggests that the C-14 content inoxide is larger than the content in base Zircaloy
metal due to the additional 17O(n,α)14C reaction [3].
Based on the cladding geometry (thickness of thebase metal and oxide layer in Table 2) and theTable 2

density of Zircaloy and the oxide layer, theabundance of C-14 in the oxide and the base
Zircaloy can be estimated as 7.5% and 92.5%,
respectively, where the oxide density is assumed tobe 5.24 g/cm3 which corresponds to 90% of thetheoretical density of zirconia, ZrO (5.82 g/cm3)[4]. This C-14 distribution corresponds roughly to
the C-14 inventory in waste claddings estimated by
the fore-mentioned ORIGEN calculation, in which
the respective percentages of 3.5% and 96.5%
were suggested for the BWR STEP I cladding [3].
It can therefore be concluded from the results ofthese measurements and calculations that theassumption of 20% IRF in the safety case wassomewhat over-conservative. However, the C-14
distribution in oxide for the PWR cladding wasmeasured as 17% [5], which can be attributed tothe oxide thickness of the claddings and that theoxide layer of PWR cladding in the literature isespecially thick at 80 µm.

Specific activity of C-14 for irradiated claddings and oxide.

Thickness (µm)
Sample ID
Base metal

F3GT3-H7-481

Specific activity (Bq/g)

Externaloxide layer(a) Cladding with internaland external oxide(b) Cladding withexternal oxide(c) Cladding
base metal(d) Externaloxide*

25.3

1.54 × 104

1.53× 104

1.49 × 104

4.25 × 104

1.49 × 104

1.50 × 104

3.83 × 104

44704.7

1.43 × 10

1.47 × 10* Units of grams in the oxide are converted to grams of zirconium, not grams of ZrO2.

(2) Released C-14 from irradiated cladding after
6.5 years aqueous immersion
Table 3 shows the leached amount of C-14 forthe respective fractions measured after 6.5 years ofaqueous immersion of irradiated cladding. The
gaseous released C-14 (0.317 Bq) is less than thedissolved C-14 species (2.00 Bq in total fraction).
The sum of the dissolved inorganic and organic
C-14 is 1.64 Bq, which corresponds roughly to thetotal dissolved C-14. The leached organic C-14 isapproximately 3 times of the amount of theinorganic C-14. On the other hand, the speciationof released C-14 was all dissolved organics,
reported by previous short-term leaching tests
using irradiated PWR cladding [5]. The reason forthe discrepancy is not clear, but it could be adifference in the analysis in that the previous
conclusion was based on the simple anion
exchange property of dissolved C-14. The leachingratio, which represents the leached amount
Table 3(gaseous + dissolved) divided by the initial
inventory, is very small at 3.8×10-5 or less, i.e. lessthan 0.01%. Past short-term leaching tests which
involve PWRs [5], fore-mentioned, or BWR STEP
III [6] cladding also showed small C-14 leachingratios of less than 4×10-5.
The oxide layer might be a dominant source ofC-14 release from irradiated cladding, taking into
account the limited C-14 release from the basemetal due to the low corrosion rate of Zircaloy [2].
However, regarding the C-14 in oxide as a sourceof instant release would be too conservative
because the released C-14 is less than 3.8×10-5 ofthe total after 6.5 years of immersion. Together
with the corrosion behavior of Zircaloy,
understanding the mechanism of how the C-14 isreleased from oxide layer, for example, thesolubility of zirconia under deep disposal
conditions is a challenge that needs to beaddressed.

Released C-14 in different fractions measured after 6.5 years aqueous immersion, in Bq.

Initial inventory
6.17×104Gas phase

Total dissolved

Dissolved inorganic

Dissolved organic

0.317

2.00

0.403

1.24

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

4. Conclusion
The specific activity of C-14 in the irradiated
BWR fuel cladding was measured and found to be
1.49×104 Bq/g in the base Zircaloy. The specificactivity in the external oxide that might be activatedfrom nitrogen impurity and oxygen wasapproximately 3 times higher than that in base
Zircaloy, which leads to the C-14 distribution of
7.5% to the oxide layer and 92.5% to the base
Zircaloy in the cladding. From the 6.5 years of
aqueous immersion, the leached C-14 from theirradiated cladding with the adhering external oxide
layer was found in gaseous form, as dissolved
inorganics and as dissolved organics, whose
composition ratios are 16%, 21%, and 63%,respectively. The total C-14 released was less than
0.01% of the C-14 inventory of the cladding. Boththe low amount of C-14 in oxide and the low leachingrate indicate that the C-14 in oxide does not have amajor impact on the instant release fraction in thesafety case. To improve the understanding of these
phenomena, knowledge of the radionuclide leaching
behavior from the oxide layer is needed. Specifically,
the estimation of the gradual release rate from thecladding oxide can be facilitated from the solubilityof zirconia under the deep repository ground waterconditions. Another future challenge is to clarify theassumed congruence behavior between the corrosion
and radionuclide release.

corrosion of Zircaloy fuel rods. EPRI NP-2789,1982.
[5] Yamaguchi, T., Tanuma, S., Yasutomi, I.,
Nakayama, T., Tanabe, H., Katsurai, K.,
Kawamura, W., Maeda, K., A study on chemicalforms and migration behavior of radionuclides inhull waste, Proc. of the 7th International
Conference on Environmental Remediation andRadioactive Waste Management, ICEM1999,1999.
[6] Yamashita, Y., Tanabe, H., Sakuragi, T.,
Takahashi, R., Sasoh, M., C-14 release behavior
and chemical species from irradiated hull waste
under geological disposal conditions, Mater. Res.
Soc. Symp. Proc. 1665, 2014.

Acknowledgement

This research is a part of the “Research anddevelopment of processing and disposal techniquefor TRU waste” program funded by the Agency for
Natural Resources and Energy in the Ministry ofEconomy, Trade and Industry of Japan.
References[1] Federation of Electric Power Companies (FEPC)
and Japan Atomic Energy Agency (JAEA).
Second progress report on research anddevelopment for TRU waste disposal in Japan,
2007.
[2] Sakuragi, T., Yamashita, Y., Akagi, M.,
Takahashi, R., Carbon 14 Distribution in
irradiated BWR fuel cladding and released
Carbon 14 after aqueous immersion of 6.5 years,
Procedia Chemistry, Volume 21, 2016.
[3] Sakuragi, T., Tanabe, H., Hirose, E., Sakashita,A., Nishimura, T., Estimation of Carbon 14
inventory in hull and end-piece wastes fromJapanese commercial reprocessing operation,
Proc. of the 15th International Conference on
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M., Parry, G. W., Smerd, P. G., Waterside

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan