- Email: [email protected]
Production of 242mAm
ARTICLE IN PRESS
Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485 www.elsevier.com/locate/nima
Production of
242m
Am
P. Benettia, A. Cesanab, A. Dodaroc, S. Mongellib, G.L. Rasellia,, M. Terranib, F. Troianid a
Istituto Nazionale di Fisica Nucleare (INFN) Sezione di Pavia and Universita` di Pavia, via Bassi 6, I-27100 Pavia, Italy b Dipartimento di Ingegneria Nucleare, Politecnico di Milano, Via Ponzio 34/3, I-20133 Milano, Italy c Radioactive Waste Management Department, ENEA, Italy d Ente per le Nuove tecnologie, l’Energia e l’Ambiente (ENEA), Italy Received 2 December 2005; received in revised form 13 March 2006; accepted 5 April 2006 Available online 2 May 2006
Abstract The isotope 242mAm is an interesting nuclide as it shows one of the known highest neutron fission cross-section. This could be useful in some special applications, like nuclear reactors for space propulsion, electricity production by direct energy conversion and small-core nuclear reactors. However its availability, in particular at high isotopic abundance, is at present a serious drawback. We have experimentally tested a growth model for the production of 242mAm by means of 241 Amðn; gÞ 242m Am reaction in thermal nuclear reactors. The obtained results are found to be in agreement with the model itself. r 2006 Elsevier B.V. All rights reserved. PACS: 27.90.+b; 28.41.Bm; 28.80+s Keywords:
242m
Am; Nuclear fuel; Enrichment
1. Introduction Several properties of 242mAm, e.g. thermal neutron cross-section, half life and number of prompt neutrons per fission, make this nuclide an ideal nuclear fuel. Its use was suggested not only for small-core nuclear reactors [1], but also as fuel for reactors designed for space propulsion [2] and for electricity production by direct energy conversion [3]. Perhaps the only true disadvantage from this point of view is that this nuclide is difficult to obtain in amounts useful for the above mentioned purposes. It is present in the waste of LWR nuclear reactors, at a level of some tenth per cent of the total americium, a concentration too low for the above mentioned applications. 242m Am can be on purpose obtained by neutron capture reaction from 241Am which is present in large quantity in nuclear fuel waste. Although the process is not straightforward, neutron irradiation of waste leads, on one hand, to its conversion into a valuable material and, on the other Corresponding author. Tel.: +39 0382 987 410; fax: +39 0382 423241.
E-mail address: [email protected] (G.L. Raselli). 0168-9002/$ - see front matter r 2006 Elsevier B.V. All rights reserved. doi:10.1016/j.nima.2006.04.029
hand, to the disposal of an isotope which is one of the most radiotoxic nuclides. In a previous work we considered theoretically the production of 242mAm by irradiation in a thermal nuclear reactor [4]. Now we report the experimental results of an irradiation test carried out to check our computations and to validate the available crosssection libraries. 2. Sample preparation and irradiation We prepared four samples of 241Am. Americium nitrate was extracted from U/Pu mixed oxides according to the process described in Ref. [5]. A drop of solution was deposited on a nuclear grade aluminium substrate and then dried. The resulting 241Am mass of each sample was of the order of 0.3 mg; a more precise value was determined a posteriori by gamma-ray counting. The 242mAm content before the neutron irradiation was also measured by gamma-ray counting and found to be less than 1.5 ppm. Two samples were enveloped in a cadmium foil, 1 mm thick, in order to suppress the thermal neutron flux, which would be very effective in fissioning the 242mAm just produced [4].
ARTICLE IN PRESS P. Benetti et al. / Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485
Finally, all samples were grouped in two pairs, each including a Cd-lined unit and a bare one. Both pairs were placed in the central channel of the TRIGA nuclear reactor operating at ENEA [6] (Casaccia, Italy) in an integral neutron flux of 2:6 1013 cm2 s1 . The irradiation times were 30 hours and 119 hours for each pair, respectively.
3. Activity measurement After irradiation, the 241Am and 242mAm masses were determined measuring the sample activities by means of high resolution gamma-ray counting. We used a coaxial intrinsic Ge detector (r ¼ 2:85 cm, h ¼ 4:71 cm) with a nominal efficiency, at 1.33 MeV, of 25% with respect to a 300 300 NaI. The source to detector distance was 10 cm and a Cd filter 6 mm thick was used in order to shield the detector against the strong 60 keV gamma-line from 241Am. For the determination of the produced 242mAm mass we measured the intensities of the lines at 984.5 keV, 1025.9 keV and 1028.5 keV emitted in the decay of 238Np coming from 242mAm alpha decay; for the 241Am mass we used the lines at 322.5, 332.4, 335.4, 376.6 and 722.0 keV emitted in the decay of 241Am [7]. In order to reduce the effect of fission and/or activation products, the samples were counted one year after the irradiation. Despite this long cooling time, several spurious nuclides (e.g. 137Cs, 60Co, 110mAg, 54Fe) survived with an activity much larger than that of interest. The lines used for the analysis, listed above, were free from any interference apart from a strong continuous (Compton) background which imposed very long counting times, about 5 105 s for each sample, in order to get a statistical precision better than 10%. The detector absolute efficiency as a function of energy in the above experimental conditions was evaluated using 152 Eu, 60Co, 131Ba and 110mAg calibrated sources; data are reported in Fig. 1.
Table 1 Quantities used for mass evaluation. 432 y and 152 y, respectively Energy (keV)
984.45 1025.87 1028.54 322.52 332.35 335.37 376.65 722.01
Branching [7]
0.278a 0.096a 0.203a 1:518 106 1:490 106 4:960 106 1:383 106 1:960 106
241
Am and
483
242m
Am half lives are
Efficiency Bare sample
Cd-lined sample
0.00128 0.00125 0.00124 0.00253 0.00253 0.00253 0.00245 0.00159
0.00122 0.00118 0.00118 0.00224 0.00224 0.00224 0.00220 0.00148
a
Values refer to 238Np decay. The branching value adopted for alpha decay from 242mAm to 238Np is 0.00459.
4. Determination of the masses The quantities relevant for the determination of the masses are listed in Table 1. In this kind of measurements the gamma-ray self-absorption is a source of systematic errors which must be taken into account. Since in case of alleged uniform samples the maximum average thickness deduced a posteriori was about 1 mg cm2 , self-absorption factor less than 1% was expected. Furthermore, we did not found any significant variation of 241Am mass when using in the calculation both low and high energy gamma-rays; hence we considered the samples uniform and the gamma self-absorption disregarding. In conclusion the actual 241Am mass in the samples were evaluated as follows: 0:299 0:071 mg and 0:385 0:006 mg in bare samples; 0:380 0:009 mg and 0:460 0:046 mg in Cd-lined samples.
5. Expected results 0.005
In a previous paper we investigated the 242mAm production through 241 Amðn; gÞ242m Am reaction [4]. The ratio between the number of 242mAm and 241Am atoms after an irradiation time t with the initial condition N 42m ð0Þ ¼ 0 is
0.004 Bare targets
Efficiency
0.003
0.002
N 42m ðtÞ R41 b ¼ ð1 eðl42m R41 Þt Þ N 41 ð0Þ eR41 t l42m R41
Cd-lined targets
(1)
where R41 is the 241Am capture rate per atom, l is the Am depletion rate, i.e. the sum of decay constant (1:56 1010 s1 [7]) and reaction rates per atom for fission and neutron absorption respectively:
0.001 0.0009 0.0008 0.0007 0.0006
242m
0.0005 0.1
1 Energy (MeV)
Fig. 1. Detector absolute efficiency versus energy.
l42m ¼ l42m þ R42m;f þ R42m;a ( 8:59 108 s1 Bare ¼ 1:39 109 s1 Cdlined
ARTICLE IN PRESS P. Benetti et al. / Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485 10-1 Bare targets Cd-lined targets 10-2
6. Experimental results and discussion
10-3
10-4
10-5 1
10-2
10-4
R41 ðs1 Þ
R42m;f ðs1 Þ
R42m;a ðs1 Þ
Bare Cd-lined (thickness ¼ 1 mm)
8:32 109 7:64 1010
7:12 108 1:08 109
1:46 108 1:69 1010
10-3 Enrichment
Bare targets
10-4
10-5
10-6 10
Cd-lined targets
100 Irradiation time (hours)
Fig. 2. Theoretical (continuous lines) atomic ratio expected at different irradiation times. The experimental data (dots) are discussed in Section 6.
Am enrichment in low neutron flux.
10-1
Table 2 Mean reaction rates per atom for Sample
242m
Bare targets Cd-lined targets
10-3
Am production and depletion
1000
100
The measurements carried out on the irradiated samples give the results shown in Table 3, where the calculated 242m
10 100 Irradiation time (days)
Fig. 3. Theoretical
Enrichment
where b is the fraction of captures which lead to the metastable state of 242Am. We assumed b ¼ 0:175 for both the bare and the Cd-lined sample [4]. Therefore, under the above specified neutron flux the mean reaction rates per atom are those listed in Table 2; the log–log plots of N 42m ðtÞ=N 41 ð0Þ, i.e. the theoretical curves of growth obtained with Eq. (1) as a function of irradiation time, are reported as continuous lines in Fig. 2. At first glance it appears surprising that the theoretical enrichments expected with Cd-lined samples are lower than those with bare ones. This is true for relatively short irradiation times and/or low neutron fluxes as it can be deduced from Eq. (1). Figs. 3 and 4 show the predicted long time irradiation behaviour of 242mAm production in low ð2:6 1013 cm2 s1 Þ and high ð2:6 1014 cm2 s1 Þ neutron fluxes, respectively. In the case of the lower neutron flux the maximum expected enrichment after 1000 days of irradiation is 1.1% for Cd-lined and 1.9% for bare sample. In the higher flux the situation is strongly different: the enrichment of Cd-lined sample could reach 5% after 500 irradiation days and the value of about 9% after 1000 days is still not asymptotic. In Ref. [4] an asymptotic enrichment, although unpractical, of 17.8% was found.
Enrichment
484
1
Fig. 4. Theoretical
10 100 Irradiation time (days) 242m
1000
Am enrichment in high neutron flux.
figures corresponding to the achieved enrichments are as well reported. Measured values are also shown as dots in the log–log plot of Fig. 2. As it can be seen, the results with bare samples are in very good agreement with the expected values. On the contrary there are some discrepancies between the experimental and expected enrichment factor for Cd-lined samples. This could be ascribed to several reasons, since the reaction rates under Cd depend strongly on parameters like Cd thickness and uniformity; reactor neutron spectrum in the low epithermal energy region. As stated above, the curve was calculated for a nominal TRIGA reactor spectra and a Cd liner with uniform thickness of 1 mm.
ARTICLE IN PRESS P. Benetti et al. / Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485 Table 3 242m Am mass as measured after irradiation Sample
Bare
Irradiation time 30 h
119 h
Am Am Enrichment
ð5:56 0:08Þ 108 g ð3:85 0:06Þ 104 g 144 4 ppm
ð1:93 0:05Þ 107 g ð2:99 0:71Þ 104 g 645 170 ppm
242m
ð1:16 0:27Þ 108 g ð3:80 0:09Þ 104 g 30:5 7:8 ppm
ð5:37 0:16Þ 108 g ð4:60 0:46Þ 104 g 117 15 ppm
242m 241
Cd-lined
Am Am Enrichment 241
485
The obtained results clearly indicate that when 242mAm has to be used at an isotopic abundance larger than 10% it is necessary to apply an enrichment process, as stressed also by other authors [8]. This is not an easy task when an appreciable quantity of 242mAm has to be produced. In the past, Calutrons were used to enrich small quantities of 242m Am [9], however this is an inefficient technique. Other systems should be investigated, as Ion Cyclotron Resonance (ICR) plasma heating or Laser assisted isotope separation. References
7. Conclusions The aim of this work is to prove both the feasibility of production of 242mAm starting from 241Am samples under controlled irradiation conditions and the validation of a production growth model. According to the obtained results, these tasks were experimentally fulfilled using bare samples. In case of irradiations carried out in filtered low energy neutron conditions, the discrepancy is however within a factor of two.
[1] Y. Ronen, et al., Nucl. Technol. 129 (3) (2000) 407. [2] C. Rubbia, Fission fragments-heating for space propulsion, CERN SL-Note-2000-036, April 2000. [3] Y. Ronen, et al., Nucl. Instr. and Meth. A 531 (2004) 639. [4] A. Cesana, et al., Nucl. Technol. 148 (1) (2004) 97. [5] A. Dodaro, et al., ENEA Technical Note RAD-CAT(03)02, October 2003. [6] A. Festinesi, Present and future activities of TRIGA RC-1 reactor, Proceeding of TOC-19 Ninth European Triga Users’ Conference, ENEA Casaccia, Italy, 7–9 October 1986, II-16. [7] R.B. Firestone (Ed.), Table of Isotopes, eighth ed., Wiley, New York, 1996. [8] Y. Ronen, G. Raitses, Nucl. Instr. and Meth. A 522 (2004) 558. [9] L.O. Love, et al., Nucl. Instr. and Meth. 38 (1965) 148.
Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485 www.elsevier.com/locate/nima
Production of
242m
Am
P. Benettia, A. Cesanab, A. Dodaroc, S. Mongellib, G.L. Rasellia,, M. Terranib, F. Troianid a
Istituto Nazionale di Fisica Nucleare (INFN) Sezione di Pavia and Universita` di Pavia, via Bassi 6, I-27100 Pavia, Italy b Dipartimento di Ingegneria Nucleare, Politecnico di Milano, Via Ponzio 34/3, I-20133 Milano, Italy c Radioactive Waste Management Department, ENEA, Italy d Ente per le Nuove tecnologie, l’Energia e l’Ambiente (ENEA), Italy Received 2 December 2005; received in revised form 13 March 2006; accepted 5 April 2006 Available online 2 May 2006
Abstract The isotope 242mAm is an interesting nuclide as it shows one of the known highest neutron fission cross-section. This could be useful in some special applications, like nuclear reactors for space propulsion, electricity production by direct energy conversion and small-core nuclear reactors. However its availability, in particular at high isotopic abundance, is at present a serious drawback. We have experimentally tested a growth model for the production of 242mAm by means of 241 Amðn; gÞ 242m Am reaction in thermal nuclear reactors. The obtained results are found to be in agreement with the model itself. r 2006 Elsevier B.V. All rights reserved. PACS: 27.90.+b; 28.41.Bm; 28.80+s Keywords:
242m
Am; Nuclear fuel; Enrichment
1. Introduction Several properties of 242mAm, e.g. thermal neutron cross-section, half life and number of prompt neutrons per fission, make this nuclide an ideal nuclear fuel. Its use was suggested not only for small-core nuclear reactors [1], but also as fuel for reactors designed for space propulsion [2] and for electricity production by direct energy conversion [3]. Perhaps the only true disadvantage from this point of view is that this nuclide is difficult to obtain in amounts useful for the above mentioned purposes. It is present in the waste of LWR nuclear reactors, at a level of some tenth per cent of the total americium, a concentration too low for the above mentioned applications. 242m Am can be on purpose obtained by neutron capture reaction from 241Am which is present in large quantity in nuclear fuel waste. Although the process is not straightforward, neutron irradiation of waste leads, on one hand, to its conversion into a valuable material and, on the other Corresponding author. Tel.: +39 0382 987 410; fax: +39 0382 423241.
E-mail address: [email protected] (G.L. Raselli). 0168-9002/$ - see front matter r 2006 Elsevier B.V. All rights reserved. doi:10.1016/j.nima.2006.04.029
hand, to the disposal of an isotope which is one of the most radiotoxic nuclides. In a previous work we considered theoretically the production of 242mAm by irradiation in a thermal nuclear reactor [4]. Now we report the experimental results of an irradiation test carried out to check our computations and to validate the available crosssection libraries. 2. Sample preparation and irradiation We prepared four samples of 241Am. Americium nitrate was extracted from U/Pu mixed oxides according to the process described in Ref. [5]. A drop of solution was deposited on a nuclear grade aluminium substrate and then dried. The resulting 241Am mass of each sample was of the order of 0.3 mg; a more precise value was determined a posteriori by gamma-ray counting. The 242mAm content before the neutron irradiation was also measured by gamma-ray counting and found to be less than 1.5 ppm. Two samples were enveloped in a cadmium foil, 1 mm thick, in order to suppress the thermal neutron flux, which would be very effective in fissioning the 242mAm just produced [4].
ARTICLE IN PRESS P. Benetti et al. / Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485
Finally, all samples were grouped in two pairs, each including a Cd-lined unit and a bare one. Both pairs were placed in the central channel of the TRIGA nuclear reactor operating at ENEA [6] (Casaccia, Italy) in an integral neutron flux of 2:6 1013 cm2 s1 . The irradiation times were 30 hours and 119 hours for each pair, respectively.
3. Activity measurement After irradiation, the 241Am and 242mAm masses were determined measuring the sample activities by means of high resolution gamma-ray counting. We used a coaxial intrinsic Ge detector (r ¼ 2:85 cm, h ¼ 4:71 cm) with a nominal efficiency, at 1.33 MeV, of 25% with respect to a 300 300 NaI. The source to detector distance was 10 cm and a Cd filter 6 mm thick was used in order to shield the detector against the strong 60 keV gamma-line from 241Am. For the determination of the produced 242mAm mass we measured the intensities of the lines at 984.5 keV, 1025.9 keV and 1028.5 keV emitted in the decay of 238Np coming from 242mAm alpha decay; for the 241Am mass we used the lines at 322.5, 332.4, 335.4, 376.6 and 722.0 keV emitted in the decay of 241Am [7]. In order to reduce the effect of fission and/or activation products, the samples were counted one year after the irradiation. Despite this long cooling time, several spurious nuclides (e.g. 137Cs, 60Co, 110mAg, 54Fe) survived with an activity much larger than that of interest. The lines used for the analysis, listed above, were free from any interference apart from a strong continuous (Compton) background which imposed very long counting times, about 5 105 s for each sample, in order to get a statistical precision better than 10%. The detector absolute efficiency as a function of energy in the above experimental conditions was evaluated using 152 Eu, 60Co, 131Ba and 110mAg calibrated sources; data are reported in Fig. 1.
Table 1 Quantities used for mass evaluation. 432 y and 152 y, respectively Energy (keV)
984.45 1025.87 1028.54 322.52 332.35 335.37 376.65 722.01
Branching [7]
0.278a 0.096a 0.203a 1:518 106 1:490 106 4:960 106 1:383 106 1:960 106
241
Am and
483
242m
Am half lives are
Efficiency Bare sample
Cd-lined sample
0.00128 0.00125 0.00124 0.00253 0.00253 0.00253 0.00245 0.00159
0.00122 0.00118 0.00118 0.00224 0.00224 0.00224 0.00220 0.00148
a
Values refer to 238Np decay. The branching value adopted for alpha decay from 242mAm to 238Np is 0.00459.
4. Determination of the masses The quantities relevant for the determination of the masses are listed in Table 1. In this kind of measurements the gamma-ray self-absorption is a source of systematic errors which must be taken into account. Since in case of alleged uniform samples the maximum average thickness deduced a posteriori was about 1 mg cm2 , self-absorption factor less than 1% was expected. Furthermore, we did not found any significant variation of 241Am mass when using in the calculation both low and high energy gamma-rays; hence we considered the samples uniform and the gamma self-absorption disregarding. In conclusion the actual 241Am mass in the samples were evaluated as follows: 0:299 0:071 mg and 0:385 0:006 mg in bare samples; 0:380 0:009 mg and 0:460 0:046 mg in Cd-lined samples.
5. Expected results 0.005
In a previous paper we investigated the 242mAm production through 241 Amðn; gÞ242m Am reaction [4]. The ratio between the number of 242mAm and 241Am atoms after an irradiation time t with the initial condition N 42m ð0Þ ¼ 0 is
0.004 Bare targets
Efficiency
0.003
0.002
N 42m ðtÞ R41 b ¼ ð1 eðl42m R41 Þt Þ N 41 ð0Þ eR41 t l42m R41
Cd-lined targets
(1)
where R41 is the 241Am capture rate per atom, l is the Am depletion rate, i.e. the sum of decay constant (1:56 1010 s1 [7]) and reaction rates per atom for fission and neutron absorption respectively:
0.001 0.0009 0.0008 0.0007 0.0006
242m
0.0005 0.1
1 Energy (MeV)
Fig. 1. Detector absolute efficiency versus energy.
l42m ¼ l42m þ R42m;f þ R42m;a ( 8:59 108 s1 Bare ¼ 1:39 109 s1 Cdlined
ARTICLE IN PRESS P. Benetti et al. / Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485 10-1 Bare targets Cd-lined targets 10-2
6. Experimental results and discussion
10-3
10-4
10-5 1
10-2
10-4
R41 ðs1 Þ
R42m;f ðs1 Þ
R42m;a ðs1 Þ
Bare Cd-lined (thickness ¼ 1 mm)
8:32 109 7:64 1010
7:12 108 1:08 109
1:46 108 1:69 1010
10-3 Enrichment
Bare targets
10-4
10-5
10-6 10
Cd-lined targets
100 Irradiation time (hours)
Fig. 2. Theoretical (continuous lines) atomic ratio expected at different irradiation times. The experimental data (dots) are discussed in Section 6.
Am enrichment in low neutron flux.
10-1
Table 2 Mean reaction rates per atom for Sample
242m
Bare targets Cd-lined targets
10-3
Am production and depletion
1000
100
The measurements carried out on the irradiated samples give the results shown in Table 3, where the calculated 242m
10 100 Irradiation time (days)
Fig. 3. Theoretical
Enrichment
where b is the fraction of captures which lead to the metastable state of 242Am. We assumed b ¼ 0:175 for both the bare and the Cd-lined sample [4]. Therefore, under the above specified neutron flux the mean reaction rates per atom are those listed in Table 2; the log–log plots of N 42m ðtÞ=N 41 ð0Þ, i.e. the theoretical curves of growth obtained with Eq. (1) as a function of irradiation time, are reported as continuous lines in Fig. 2. At first glance it appears surprising that the theoretical enrichments expected with Cd-lined samples are lower than those with bare ones. This is true for relatively short irradiation times and/or low neutron fluxes as it can be deduced from Eq. (1). Figs. 3 and 4 show the predicted long time irradiation behaviour of 242mAm production in low ð2:6 1013 cm2 s1 Þ and high ð2:6 1014 cm2 s1 Þ neutron fluxes, respectively. In the case of the lower neutron flux the maximum expected enrichment after 1000 days of irradiation is 1.1% for Cd-lined and 1.9% for bare sample. In the higher flux the situation is strongly different: the enrichment of Cd-lined sample could reach 5% after 500 irradiation days and the value of about 9% after 1000 days is still not asymptotic. In Ref. [4] an asymptotic enrichment, although unpractical, of 17.8% was found.
Enrichment
484
1
Fig. 4. Theoretical
10 100 Irradiation time (days) 242m
1000
Am enrichment in high neutron flux.
figures corresponding to the achieved enrichments are as well reported. Measured values are also shown as dots in the log–log plot of Fig. 2. As it can be seen, the results with bare samples are in very good agreement with the expected values. On the contrary there are some discrepancies between the experimental and expected enrichment factor for Cd-lined samples. This could be ascribed to several reasons, since the reaction rates under Cd depend strongly on parameters like Cd thickness and uniformity; reactor neutron spectrum in the low epithermal energy region. As stated above, the curve was calculated for a nominal TRIGA reactor spectra and a Cd liner with uniform thickness of 1 mm.
ARTICLE IN PRESS P. Benetti et al. / Nuclear Instruments and Methods in Physics Research A 564 (2006) 482–485 Table 3 242m Am mass as measured after irradiation Sample
Bare
Irradiation time 30 h
119 h
Am Am Enrichment
ð5:56 0:08Þ 108 g ð3:85 0:06Þ 104 g 144 4 ppm
ð1:93 0:05Þ 107 g ð2:99 0:71Þ 104 g 645 170 ppm
242m
ð1:16 0:27Þ 108 g ð3:80 0:09Þ 104 g 30:5 7:8 ppm
ð5:37 0:16Þ 108 g ð4:60 0:46Þ 104 g 117 15 ppm
242m 241
Cd-lined
Am Am Enrichment 241
485
The obtained results clearly indicate that when 242mAm has to be used at an isotopic abundance larger than 10% it is necessary to apply an enrichment process, as stressed also by other authors [8]. This is not an easy task when an appreciable quantity of 242mAm has to be produced. In the past, Calutrons were used to enrich small quantities of 242m Am [9], however this is an inefficient technique. Other systems should be investigated, as Ion Cyclotron Resonance (ICR) plasma heating or Laser assisted isotope separation. References
7. Conclusions The aim of this work is to prove both the feasibility of production of 242mAm starting from 241Am samples under controlled irradiation conditions and the validation of a production growth model. According to the obtained results, these tasks were experimentally fulfilled using bare samples. In case of irradiations carried out in filtered low energy neutron conditions, the discrepancy is however within a factor of two.
[1] Y. Ronen, et al., Nucl. Technol. 129 (3) (2000) 407. [2] C. Rubbia, Fission fragments-heating for space propulsion, CERN SL-Note-2000-036, April 2000. [3] Y. Ronen, et al., Nucl. Instr. and Meth. A 531 (2004) 639. [4] A. Cesana, et al., Nucl. Technol. 148 (1) (2004) 97. [5] A. Dodaro, et al., ENEA Technical Note RAD-CAT(03)02, October 2003. [6] A. Festinesi, Present and future activities of TRIGA RC-1 reactor, Proceeding of TOC-19 Ninth European Triga Users’ Conference, ENEA Casaccia, Italy, 7–9 October 1986, II-16. [7] R.B. Firestone (Ed.), Table of Isotopes, eighth ed., Wiley, New York, 1996. [8] Y. Ronen, G. Raitses, Nucl. Instr. and Meth. A 522 (2004) 558. [9] L.O. Love, et al., Nucl. Instr. and Meth. 38 (1965) 148.