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Determination of the 14C Content in Activated Steel Components from a Neutron Spallation Source and a Nuclear Power Plant Dorothea Schumann,*,† Tanja Stowasser,† Benjamin Volmert,‡ Ines Günther-Leopold,† Hanspeter Linder,† and Erich Wieland† †

Paul Scherrer Institute, Villigen, Aargau 5232, Switzerland National Cooperative for the Disposal of Radioactive Waste (Nagra), Wettingen, Aargau 5430, Switzerland

‡

ABSTRACT: The 14C content in activated steel components from the Swiss Nuclear Power Plant (NPP) Gösgen and the Spallation Neutron Source SINQ at the Paul Scherrer Institute is determined using a wet chemistry digestion technique and liquid scintillation counting for 14C activity measurements. The 14C activity of an activated fuel assembly steel nut from the NPP is further compared with theoretical predictions made on the basis of a Monte Carlo reactor model for this NPP. Knowledge of the 14C inventory in these activated steel materials is important in conjunction with future corrosion studies on these materials aimed at identifying the 14 C containing organic compounds possibly formed in the cementbased near field of a repository for radioactive waste.

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volatile organic carbon compounds as well as alcohols, aldehydes, and carboxylic acid remaining in solution during steel corrosion.2 All these species feature a high radiological risk due to their high mobility either in the liquid or in the gas phase. The main sources of 14C in neutron-activated steel are the interactions of neutrons with nitrogen (14N(n,p)14C), carbon (13C(n,γ)14C), and oxygen (17O(n,α)14C), with the first one being the most important due to the isotopic abundance of 14N and the large capture cross-section for neutrons. Depending on the kind of steel, the amount of the involved reaction partners is variable. The nitrogen content in steel can typically vary between 0.04 and 0.1 wt %. Only a few experimental studies exist for the determination of 14C in reactor steel. The most common method for the chemical digestion of the sample material is combustion,3,4 although acid treatment was also employed.5 Hou et.al.6 used the commercially available Packard Sample Oxidizer for the analysis of graphite and concrete, which was also applied for steel and paint samples. The present study is aimed at determining experimentally the 14 C content in activated steel samples relevant to the Swiss disposal program, originating from the NPP Gösgen (KKG) and from the spallation neutron source SINQ at the Paul Scherrer Institut (PSI) for the purpose of comparison. We applied the wet acid chemistry method for sample digestion and modified it slightly to serve our needs. The results obtained by liquid scintillation counting (LSC) for the NPP sample are compared with theoretical predictions of the inventory of

ver the past decades, there has been an increasing requirement for quality and reliability of information on the characteristics (physical, chemical, radiological) of radioactive waste for disposal.1 While the chemical and physical properties of nuclear fission products and actinides formed in spent fuel assemblies of nuclear power plants (NPP) and their impact to society and the environment have been studied extensively already for several decades, comparably minor attention had been paid to the potential hazard of long-lived radionuclides contained in structural materials of an NPP, for instance in steel components. One of the open questions in this research area is the contribution of 14C to the dose released from a backfilled repository, especially when a partial or complete release has to be expected due to anoxic corrosion of activated metals over time. Knowledge of the radionuclide inventories is a prerequisite for the quantitative evaluation of the radiological hazard of radionuclides released from a repository for radioactive waste. In the case of nuclear facility hardware like steel components, the main short- and midterm dose rate contributors 60Co and 54 Mn, produced by neutron activation with comparable high cross sections, can easily be determined by measuring the intensity of their γ rays. The determination of long-lived isotopes, however, and in particular of those which do not have measurable γ rays, such as 14C with a half-life of 5730 y and a maximal energy of beta emissions of 156 keV, requires cost intensive and time-consuming chemical separation techniques in the course of the radioanalytical measurements. 14C is not only of particular interest due to its nuclear properties but also because of the chemical behavior of the 14C containing compounds expected to be formed during anoxic steel corrosion. Several studies indicate the possible formation of © 2014 American Chemical Society

Received: February 17, 2014 Accepted: May 9, 2014 Published: May 9, 2014 5448

dx.doi.org/10.1021/ac500654a | Anal. Chem. 2014, 86, 5448−5454

Analytical Chemistry

Article

Figure 1. KKG MCNP reactor model and resulting distribution of the total neutron flux; the approximate location of the guide tube nut at the bottom of a FA is indicated.

activation products, using as reference also the values for 54Mn and 60Co. The obtained and evaluated values for the 14C inventory of the activated steel samples from KKG shall serve as basic information in conjunction with future corrosion experiments.7

characteristic gamma lines emitted from the activated component, allowing a comparison with the model predictions. Prediction of the Nuclide Inventory for a KKG Fuel Assembly Guide Tube Nut. To determine the nuclide inventory of a KKG fuel assembly (FA) guide tube nut (“Brennelement-Führungsrohrmutter”), it is necessary to calculate the neutron radiation field that it has been exposed to during its time within the reactor pressure vessel. Since an MCNP (Monte Carlo N-Particle)8 reactor model for KKG had been already set up by Nagra for the “Swiss NPP Decommissioning Study 2011”,9 it was decided to use this already available Monte Carlo model for an approximate inventory determination for the FA guide tube nut. Although the development of a specialized reactor model being optimized for the task at hand could deliver more accurate information about the neutron radiation field at the FA nut’s location, the additional efforts would have been certainly beyond the scope of the present study. Figure 1 shows the Nagra MCNP KKG model and the corresponding result of the Monte Carlo simulation for the total neutron flux distribution being itself an input to the ensuing calculation of the radionuclide inventory for the Swiss decommissioning study 2011.9 Although the MCNP model is primarily optimized for the determination of the neutron transport outside the core and the reactor pressure vessel up to the surrounding drywell and basin area walls, it still can be used to estimate the neutron activation characteristic for areas near the active core. For this reason, the position of a typical FA as well as that of the FA guide tube nut are highlighted in Figure 1. It shows that the neutron total flux has been dropped at the nut’s position by already 1 to 2 orders of magnitude compared to the level at the active core (around 2 × 1014 n/cm2s).

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THEORETICAL PREDICTIONS Boundary Conditions and Description of the Method. For the theoretical determination of the nuclide inventory of an exposed internal reactor component, the following data has to be acquired or estimated: the material composition of the component including any impurity concentrations, the dimension of the component, the neutron spectrum and the corresponding total flux at the component’s location, the irradiation history, i.e., the change of the neutron radiation field at the component’s location over the time interval of exposure and the decay time after removal of the component from the reactor. If the dimension of the component is small compared to the gradient of the neutron radiation field (i.e., the change of the neutron flux and spectrum within the component can be neglected), the determination of a representative neutron flux and spectrum is sufficient. If additionally, the time dependence of the neutron radiation field can be neglected because the exposure conditions are approximately constant at the component’s location, the determination of a total exposure time is sufficient. By using an adequate nuclear burn-up code, the information above is used for the calculation of the total nuclide inventory of the component for any point in time during its exposure and after its removal from the reactor. Certain nuclides of the calculated inventory can be verified by measurement of the 5449



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After the material testing, three types of stainless steel samples were cut into suitable pieces with masses ranging between 100 and 300 mg (Figure 3). Six of these samples were used for the determination of the 14C inventory. A detailed description of the sample materials is given elsewhere.13

Taking the exact location of the nut into account, the corresponding neutron radiation field was interpolated by the neutron flux data being already available in the context of the existing KKG decommissioning study (i.e., total neutron flux for the nut = ca. 8 × 1012 n/cm2s). The interpolation had to be rather coarse since the resolution of the underlying KKG MCNP calculation is limited to node spacing intervals of ca. 20 cm. Thus, the resulting interpolated flux information is considered as an approximation only and should therefore only be used to support the order of magnitude of the presented 14C activity measurements. After retrieving the corresponding neutron spectrum (in 84 energy groups discretization) at the FA nut’s location, a burn up code (GRS-AKTIV210) was used to calculate the induced nuclide inventory (i.e., the resulting activities for 60Co, 14C, and 54 Mn) of the nut for the given exposure and decay time.

Figure 3. (a) STIP samples after the material test. (b−d) Cutting into suitable pieces for 14C determination.

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Chemical Separation Procedure. Digestion in acidic solution was applied for extracting carbon from the metallic sample material. The flow scheme for the procedure is shown in Figure 4 while a schematic view of the apparatus as well as an image is depicted in Figure 5. About 100 mg of steel sample was placed into a two-neck reaction vessel and flushed with nitrogen. The gas flow was led through a reflux condenser, followed by a first gas-sampling unit filled with 50 mL of 1 M HCl for acid adsorption and finally through two gas-sampling units filled with 20 mL of 1 M NaOH each, for the absorption of CO2. A mixture of concentrated HNO3 and HCl, 5 mL each (aqua regia), was added via an acid reservoir which was tightly connected to the apparatus (Figure 5). The acid solution in the reaction vessel was heated for about 10 min until the steel was dissolved. The insoluble residue of carbon, a component of steel, was further dissolved by using a mixture of H2SO4/HClO4/HNO3 instead of aqua regia. For the final dissolution step, all gassampling units were replaced. All fractions were analyzed separately using LSC to determine the 14C activities. The background was checked by taking an aliquot of the NaOH containing absorption solutions and removing 14C by adding acid. The measurement of the remaining solution provided the information on disturbing background counts. Quenching effects due to the different acid content can be neglected because of the relatively high dilution of the solutions. In particular, we used 0.1 mL of 1 M NaOH + 0.9 mL of H2O + 10 mL of Aquasafe for the samples and 0.1 mL of 1 M NaOH + 0.2 mL of 1 M HCl + 0.7 mL of H2O + 10 mL of Aquasafe for the background, resulting in concentrations of 0.01 M NaOH and 0.01 M HCl, respectively. Analytical Methods. γ-Emitting Radionuclides. Each sample was first analyzed to determine the γ-activity using a commercially available HPGe detector (ORTEC) and the data processing program GENIE2000 (CANBERRA). Liquid Scintillation Counting. LSC measurements were carried out using a Canberra Packard Tricarb 2250CA liquid scintillation analyzer (PerkinElmer, USA). As scintillator cocktail, we used Aquasafe 500 Plus from Zinsser Analytic. The standardized 14C solution (CFZ640) was purchased from Eckert & Ziegler Isotope Products, USA. Measurements of 14C samples and the background were conducted using Polyvials20. Each sample was measured 5 × 10 min in the energy window of 12−156 keV. Three cocktails with the following compositions were applied to each of the four NaOH solutions sampled from the gas-sampling units, i.e., dissolution with aqua regia and the H2SO4/HClO4/HNO3 mixture: (a) sample: 100 μL aliquot from NaOH gas-sampling

EXPERIMENTAL SECTION Sample Description and Treatment. Samples from KKG. Five guide tube nuts were retrieved from the storage pool of the KKG and transferred to the PSI Hot Laboratory. The nuts are principally made from three different kinds of stainless steel: 1.4541 (X6CrNiTi18-10), 1.4550 (X6CrNiNb18-10), or 1.4571 (X6CrNiMoTi17-12-2), differing only slightly with respect to the content of Mo as well as Ti and Nb.11 They were situated at the top and bottom end of fuel rods, and therefore, their activation cycle, which was terminated on June 4, 2011, is well-known. Each nut (see Figure 2a) had a weight of ∼5 g

Figure 2. (a) Activated steel nuts obtained from KKG. (b) Segment prepared from a steel nut. (Inset in b) Cutting scheme of the segment.

(diameter: 1 cm; height: 1.1 cm) and a contact dose rate of ∼100 mSv/h. Two nuts were processed to prepare small specimens for laboratory experiments (Figure 2b). A slice with approximately 2 mm thickness was cut into 6 pieces as indicated in Figure 2. The six segments had masses between 50 and 100 mg. Three of them were used for the analytical study. Samples from SINQ. PSI operates one of the most powerful continuous spallation sources worldwide, the so-called spallation neutron source SINQ. Protons are accelerated in a ring cyclotron up to 590 MeV and are fully stopped in a heavy metal target, where the neutron production takes place. The samples for this study were secured in the framework of the socalled “SINQ target-irradiation program (STIP)”. This program is aimed at investigating material properties of structure materials under conditions representative for a high-power spallation source.12 The SINQ target consists of several rods filled with lead, some of them filled with other material foreseen for scientific study. We used 3 types of ferritic/martensitic (FM) steels (Optifer, Optimax A, and Optimax C), which where irradiated in the STIP I program from July 1998 until December 1999. After finishing the material research program, these samples were regarded as waste materials, and therefore, they could be used for the present study. 5450



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Figure 4. Flowchart for the determination of 14C activity in the stainless steel samples.

Figure 5. Schematic diagram of the experimental design (left) complemented by an image of the experimental setup (right).

unit, 900 μL of water, and 10 mL of scintillator; (b) sample + 14 C standard: 100 μL aliquot from NaOH gas-sampling unit, 3.7 Bq 14C standard solution (10 μL), 900 μL of water, and 10 mL of scintillator; (c) background sample: 100 μL aliquot from NaOH gas-sampling unit, 200 μL of 1 M HCl, 700 μL of water, and 10 mL of scintillator. The 14C activity in the NaOH solution from the gas-sampling units was calculated as follows: activity =

dilution with 0.2 M HNO3 using inductively coupled plasma optical emission spectroscopy (Vista Pro AX ICP-OES, Varian Inc., USA). Nitrogen and Oxygen Measurements. The determination of nitrogen and oxygen in the KKG steel sample was carried out using the inert gas fusion technique with the LECO-TC-436 system. The weighed sample is melted in a graphite crucible under helium atmosphere. The oxygen in the sample combines with carbon and is converted to CO2 in a catalytic reaction. A solid-state infrared absorption detector determines the CO2 content from which the weight % of oxygen in the sample was calculated. Nitrogen released from the sample is reduced to N2, which is determined by a thermal conductivity cell.

(cpm a − cpm c) × 3.7 Bq × V [mL] (cpm b − cpm a) × 0.1 mL

where cpm a, b, and c are counts per minute of cocktail a, b, and c and V is volume of the NaOH in the gas sampling unit. Inductively Coupled Plasma Optical Emission Spectroscopy. The concentrations of stable Mn and Co were determined after dissolution of the steel in aqua regia and subsequent 5451



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combined uncertainties of the final values, are summarized. One γ-measurement per sample has been performed. The averaged statistical errors of the γ-measurements represent the uncertainty of the peak area determination, which can vary depending on the count rate. The LSC measurements were repeated 5 times per vial, resulting in the averaged statistical errors shown in Table 1. These values include three types of measurement per sample: the 14C content in the first absorption vessel, the 14C content in the same vessel including a known added amount of 14C standard, and the background/blank sample. For each sample, both the first and the second absorption vessel were considered by summing up the activity values. Additionally, the uncertainties for the volume measurements and the chemical yield determinations have to be considered as listed in Table 1. Finally, we determined the mean values of the multiple determinations (3-fold for the NPP steel nut and 2-fold for SINQ samples). Determination of the 44Ti, 54Mn, 60Co, and 14C Content in STIP Samples. The activities of the γ-emitters 44Ti, 54Mn, and 60Co as well as the β-emitter 14C are listed in Table 2. The 14 C activities associated with the metallic and the graphite fractions are listed separately. In all cases, the 14C activities in the second flask yielded ≤3% on average. These values were utilized to predict the sample processing yield. Note that two pieces (replicates) of the same sample were processed for each of the three STIP samples with the aim of discerning possible deviations due to inhomogeneities of the sample. The γmeasurements show reasonable agreement between the replicates for all radionuclides while the activities of the three different STIP samples are different. These differences are attributed to differences in the composition of the steels, especially differences in the Co and Mn contents. 13 Furthermore, the differences in the activities result from differences in the irradiation position, which has an influence on the neutron or proton-induced nuclear reactions, respectively, as well as the particle density dependent on the irradiation position. The presence of 44Ti as the main dosecontributing radionuclide indicates that the type of involved nuclear reactions in a spallation neutron source is very different from that in a NPP where 60Co dominates. The uncertainties of the 54Mn measurements are relatively high because of the low counting rate in the peak. The 14C activities for the samples 17(A/B) and 25(c/d) fairly agree while a rather high deviation of about 25% was found between the two samples 22-6 and 22-7. Note that the difference in activity is due to the significantly higher 14C content observed in the graphite fraction of sample 22-7. Inhomogeneity of the

RESULTS AND DISCUSSION Determination of the Chemical Yield. The chemical yield was determined using 1 g of nonactivated stainless steel standard sample with a known (certified) amount of carbon (0.864%). The standard material was treated in the same way as the activated steel samples. However, the CO2 capture in the gas-sampling units had to be modified. The carbon absorbed in the NaOH solutions was precipitated with 1 M Ba(OH)2 as BaCO3. The precipitate was separated by centrifugation and, following phase separation, washed with bidistilled water, dried in a desiccator, and weighed. Yields between 95% and 99% for the carbon recovery were obtained. The chemical yield was further checked by adding 14C standard solution to nonirradiated steel samples prior to the dissolution process. The recovery rate determined from eight replicates ranged between 83% and 99%. The recovery rates obtained by these two independent methods demonstrate the reliability and reproducibility of the wet digestion method as used in this study. Determination of the Oxygen and Nitrogen Content in the KKG Steel Sample. For performing theoretical calculation of the expected 14C content and comparison with the experimental results, the knowledge of the impurity content in the steel material is mandatory. In particular, the content of oxygen (17O) and nitrogen (14N) should be known, because these two elements are the main sources of 14C production during neutron activation of steel. The measurements yielded 0.0112 (±0.0031) elemental weight % for nitrogen and 0.0372 (±0.0033) elemental weight % for oxygen in the nonirradiated steel material. The concentrations of stable Mn and Co were determined to be 2.001 (±0.04) weight % and 0.0176 (±0.0004) weight %, respectively. Evaluation of Uncertainties. In Table 1, the estimated budget of individual uncertainties, contributing to the Table 1. Uncertainty Budgeta γ-measurement, %

contributors certified uncertainty of the calibration source (γ) or standard solution (14C) certified uncertainty of the C content in steel chemical yield by precipitation of Ba(CO3)2 chemical yield by use of 14C standard volume measurement error averaged statistical error combined standard uncertainty a

LSC, %

3

0.75

1.4−12.2 3.3−12.6

0.864 2 8 3 3.5 11.6

All uncertainties are combined standard uncertainties with k = 1.

Table 2. Results of LSC and γ-Measurements for Three Different STIP Samplesa 14

no. 17-A 17-B 22-6 22-7 25-c 25-d

material Optimax Optimax Optifer Optifer Optimax Optimax

C C

A A

mass [g] Σ14C [Bq/g] 0.2247 0.2906 0.1896 0.2231 0.1802 0.2031

12 303 14 034 21 464 30 494 4084 4118

C without graphite residue [Bq/g] 9344 10 357 15 066 16 006 3778 3814

14

C in graphite residue [Bq/g] 2958 3676 6398 14 487 306 304

44

Ti [kBq/g]

5162 5120 7284 7248 2347 2166

(4.1%) (3.8%) (3.6%) (3.6%) (4.0%) (4.0%)

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Mn [kBq/g]

223 (11.1%) 244 (9.4%) 390 (7.0%) 377 (6.4%) 144 (11.8%) 123 (12.6%)

60

Co [kBq/g]

1375 1325 881 847 477 497

(3.7%) (3.7%) (3.6%) (3.6%) (4.5%) (3.6%)

a Activities are related to the mass of sample material. Date of γ-measurements: 22-6/7: 15.03.2013; 17-A/B, 25c/d: 07.05.2013; date of 14C determination: May 2013; combined standard uncertainties for the γ-measurements in parentheses; for the 14C determination, a combined statistical uncertainty of 11.6% applies for all values.

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Table 3. LSC and γ-Measurements for Three Samples from the Steel Nut from KKGa sample no.

mass [g]

Σ14C [Bq/g]

A4 A5 A6

0.0956 0.0849 0.0693

19 399 17 328 16 797

14

14

C without graphite residue [Bq/g]

C in graphite residue [Bq/g]

17 279 15 009 14 878

2120 2318 1919

54

60

11 736 (4.3%) 11 779 (4.2%) 12 136 (4.1%)

60 973 (3.6%) 62 191 (3.6%) 61 847 (3.3%)

Mn [kBq/g]

Co [kBq/g]

Activities are related to the mass of sample material. Date of γ-measurements: 07.05.2013; date of 14C determination: May 2013; combined standard uncertainties for the γ-measurements are in parentheses; for the 14C determination, a combined statistical uncertainty of 11.6% applies for all values.

a

Table 4. Mean Values of the Radionuclide Activities for the Investigated Steel Samplesa sample no.

Σ14C [Bq/g]

STIP 17 STIP 22 STIP 25 KKG steel nut prediction KKG steel nut

13 168 ± 1527 25 979 ± 6385b 4101 ± 475 17 841 ± 2070 4500

54

Mn [kBq/g]

234 ± 24 384 ± 25 134 ± 16 11 884 ± 499 9700

60

Co [kBq/g]

date EOA

EOA 54Mn [GBq/g]

EOA 60Co [kBq/g]

1350 ± 50 864 ± 31 487 ± 20 61 670 ± 2158 27 000

20.12.1999 20.12.1999 20.12.1999 04.06.2011 04.06.2011

12.1 ± 1.3 17.6 ± 1.2 6.9 ± 0.8 0.057 ± 0.002 0.021

7855 ± 290 4932 ± 178 2834 ± 128 79 450 ± 2780 34 784

a

Activities are related to the mass of sample material. EOA: end of activation. Uncertainties are the combined standard uncertainties taken from Table 1. Using the conservative approach, the higher value is quoted as uncertainty. bThe standard deviation of the averaged value derived from the two individual determinations was higher than the estimated combined standard uncertainty taken from Table 1.

sample can be excluded since the γ-measurements of the samples agree very well. Furthermore, it should be noted that the 14C ratios between the metallic and the graphite fraction range from 10:1 to nearly 1:1 for the three samples, being in total much higher than expected from the total amount of graphite normally in stainless steels (e.g., the standard sample described in the Determination of the Chemical Yield section) and the comparable low production rate directly from carbon. Probably, this phenomenon can be attributed to the fact that 14 C formed via the nuclear reactions with nitrogen and oxygen similarly does not dissolve well in aqua regia like the stable graphite in steel. Therefore, an enrichment of 14C in the graphite fraction was observed. The reason for the higher 14C activity in the graphitic residue of sample 22-7 is presently unknown. We refrained from carrying out theoretical calculations for the STIP samples as the complex character of the involved nuclear reactions with high-energetic protons and the influence of a variety of parameters (e.g., steel composition, particle intensity, position in the target) makes predictions of the radionuclide inventories very difficult. Determination of the 54Mn, 60Co, and 14C Content of the Steel Nut from KKG. The results for the three segments gained from the steel nut are shown in Table 3. The activities of the γ-emitting nuclides are in good agreement. This is also true for the 14C activities. Additionally, the ratios between metallic and graphite fractions are consistent. Discussion of the Experimental Results and Comparison with Theoretical Predictions. The mean values of the decay-corrected activities (end of activation) were determined for all samples which are listed along with the theoretical predictions for the steel nut from KKG in Table 4. The results show that the 14C activity in the steel nut is comparable to the 14 C activities determined in the different steel samples irradiated in the spallation neutron source. Differences in activities reach a factor of 4 at best. In contrast, the 54Mn activity of the KKG steel nut is more than two orders of a magnitude lower than the 54Mn activity in the STIP samples while the 60Co activity is about a factor of 10 higher. While in a reactor 60Co is mainly produced by the (n,γ) reaction of the 59 Co present as an impurity in the steel, 54Mn is produced by

the (n,2n) of 55Mn, requiring slightly higher energies of the neutrons and, thus, resulting in lower production rates. In contrast, in the SINQ facility spallation reactions of the target material Fe with high-energetic protons and neutrons enhance the production of 53Mn (the atomic number of Mn is lower than that of Fe) but not that of 60Co, because the atomic number of Co is higher than that of Fe. The theoretical predictions of the radionuclide activities of the steel nut from KKG are based on the calculations and assumptions outlined in the Theoretical Predictions section. Note that the calculated 60Co activity agrees well with the 60Co activities determined by a Microshield14 analysis of a gamma dose rate measurement of the FA nut as well as a 60Coinventory calculation for a KKG core components waste container including several thousand similar FA nuts and crews with mixed radiation histories being formerly characterized in the framework of the KKG core waste conditioning campaign COSKO.15 The experimental data for the 14C inventory agree within a factor of 4 with the theoretical predictions, which is acceptable with a view to the coarse estimates available for the neutron flux at the guide tube nut’s location in the reactor core.

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SUMMARY AND CONCLUSIONS

Using a wet digestion method combined with LSC detection, it was possible to quantify the 14C content in irradiated stainless steel samples from the SINQ facility at PSI and from the NPP Gösgen, Switzerland. The irradiation times for both the steel nut from the NPP and the SINQ samples are comparable (around two years), but the nuclear reactions leading to the radioactive products we measured are to a great extent completely different. The differences in the radionuclide inventories between the SINQ samples are believed to be due to their different irradiation positions. Thus, the inventories cannot be compared directly since the spectrum of possible nuclear reactions in a spallation target is very complex and depends strongly on the actual energy at the irradiation position. A comparison of the 14 C activities of the materials from the two different irradiation sources, however, reveals a surprisingly good agreement. The 14 C activities of the SINQ samples (with the exception of STIP sample 25) are comparable to the activity in the steel nut from 5453



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the NPP. In contrast, the 54Mn and 60Co activities are very different for the steel samples from the two sources. While the 60 Co content at the end of activation is a factor of 10 higher for the NPP sample, the SINQ samples show significantly higher 54 Mn activities (at least 2 orders of magnitude). Taking into account the different steel compositions and the irradiation positions and on the basis of similarities in the 14C activities, we infer that the activation of steel in the high power spallation source (SINQ) is comparable to that in the NPP. This is a result which is often neglected in connection with the transmutation of nuclear waste in an Accelerator Driven System (ADS). Such a system consists of a high-power particle accelerator hitting a heavy metal target for neutron production. These neutrons are used to transmute radioactive waste from NPPs in a safe way, because it is not a self-sustained nuclear reaction. However, highly activated target materials are produced inside such an ADS facility, which also have to be handled, stored, and disposed. The 14C activity determined in the steel nut from KKG agrees within a factor 4 with the theoretical predictions made on the basis of a Monte Carlo reactor model used for activation calculations in conjunction with Swiss decommissioning studies. This difference is acceptable considering the uncertainties in the underlying assumptions. Knowledge of the 14C activities in the studied steels is important in connection with future corrosion studies to be carried out on these materials. The planned corrosion study will focus on the identification and quantification of 14C containing organic compounds generated during the anaerobic corrosion of activated steel under the simulated near field conditions of a repository for radioactive waste. The study is expected to provide important information on the release paths of 14C labeled organic compounds from a cement-based repository for radioactive waste. The developed radiochemical analysis method not only is practicable and reliable for the described task but also can be applied in future investigations with similarly activated materials. For example, in the frame of the EC funded project “CAST” (Carbon Source Term), a round robin test has been proposed with the aim of estimating the 14C activation in the steel samples from reactor cores on an international level. Availability of experimental data for benchmarking these calculations will be mandatory. Therefore, the present study provides basic information for a more general survey of nuclear burn-up codes.

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REFERENCES

(1) IAEA. Strategy and methodology for radioactive waste characterization; IAEA-TECDOC-1537; IAEA: Vienna, Austria, 2007. (2) Wieland, E.; Hummel, W. The speciation of 14-C in the cementitious near field of a repository for radioactive waste; PSI report TM44-10-01; Villigen PSI: Switzerland, 2010. (3) Bushuev, A. V.; Verzilov, Y. M.; Zubarev, V. M.; Kachanovskii, A. E.; Matveev, O. V.; Proshin, L. M.; Bidulya, L. V.; Ivanov, A. A.; Kalugin, A. K. At. Energy 1992, 73, 959. (4) Berg, J. F.; Fonnesbeck, J. E. Anal. Chim. 2001, 447 (1−2), 201. (5) Raymond, A. Analyse des radioisotopes èmetteurs bêta purs (3H, 14C, 36Cl, 63Ni) dans le graphite irradiè; Rapport techniques CEART-DSD-20; CEA: France, 1990. (6) Hou, X. L. J. Radioanal. Nucl. Chem. 2007, 273, 43−48. (7) Wieland, E. Swiss Nuclear Proposal Nr. 2 “Investigation on the chemical speciation of C-14 released from activated steel: Development of the analytical methods for speciation studies and the set-up for long-term corrosion experiments”, 2011. (8) X-5 Monte Carlo Team. MCNP; A general Monte Carlo nparticle transport code, version 5, LA-UR-03-1987; Los Alamos National Lab: Los Alamos, NM, (2003, revised 2008). (9) Volmert, B.; Pantelias, M. Nuclear power plant activation analysis for the Swiss decommissioning study 2011. Proceedings of Jahrestagung Kerntechnik, Berlin, May 14−16, 2013. (10) Hesse, U.; Hummelsheim, K. GRSAKTIV-II: Programmsystem zur Berechnung der Aktivierung von Core-Bauteilen in Vielgruppendarstellung, Programmbeschreibung UNIX-Version 05/01, GRS-A-3002; Gesellschaft fü r Anlagen-und Reaktorsicherheit (GRS): Kö ln, Germany, 2001. (11) DIN EN 10088-3 Stainless steel − part 3: Technical delivery conditions for semifinished products, bars, rods, wire, sections and bright products of corrosion resisting steels for general purposes; DIN Deutsches Institut für Normung e. V.: Berlin, 2005. (12) Dai, Y.; Bauer, G. S. J. Nucl. Mater. 2001, 296, 43−53. (13) Dressler, R.; Ayranov, M.; Bemmerer, D.; Bunka, M.; Dai, Y.; Lederer, C.; Fallis, J.; StJ Murphy, A.; Pignatari, M.; Schumann, D.; Stora, T.; Stowasser, T.; Woods, P. J. J. Phys. G: Nucl. Part. Phys. 2012, 39, 105201. (14) Microshield 9 − User’s Manual; Grove Software, Inc.: Lynchburg, VA, 2011. (15) Bitterli, B.; Braun, D.; Hummel, W. Core-Schrott-Konditionierung im Kernkraftwerk Gösgen (COSKO). Proceedings of KONTEC, Dresden, Germany, March 21−23, 2007.

AUTHOR INFORMATION

Corresponding Author

*E-mail: [email protected]. Notes

The authors declare no competing financial interest.

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ACKNOWLEDGMENTS

We gratefully acknowledge the NPP Gösgen (KKG), Switzerland, for providing the stainless steel nuts. Thanks are extended to Dr. Yong Dai (PSI) for providing the samples from the SINQ target-irradiation program (STIP). We thank M. Martin and R. Grabherr (PSI) for processing the materials and the preparation of small specimens for laboratory experiments. Partial financial contribution by swissnuclear is kindly acknowledged. 5454


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