Transcript
1 92. URANIUM In addition to three natural isotopes of uranium, ROSFOND includes data for uranium-233, uranium-236 and two much shorter-lived isotopes, uranium-232 and uranium. Uranium-232. Radioactive. (T 1/2 = 68.9 d). The decay chain of uranium-232 leads to the formation of thallium-208, which emits hard gamma radiation (2.7 MeV) during beta decay, which significantly complicates the radiation situation during operations with spent fuel. The current libraries contain the following data estimates for uranium-232. FUND-2.2 estimate T.Ohsawa, T.Nakagawa, ENDF / B-VII.b2- estimate M. Chadwick, P. Young, 2005 JENDL-3.3 estimate T. Ohsawa, T Nakagawa, 1987 JEFF-3.1 estimate T.Mutsunobu, T.Kawano, Comparison of Resonance Integrals and Thermal Cross-Sections. Source σ c (eV) RI c σ f (eV) RI f ENDF / B-VII.b JENDL JEFF Muhabhab ± ± ± 30 Large discrepancies in the estimates of resonance capture integrals are due to the lack of direct experimental data. Conclusion Despite the later date of the assessment from ENDF / B-VII.b2, its advantages over the assessment from JEFF-3.1, if any, are not obvious. In particular, in JEFF-3.1, Derrien's 1994 estimate is used in the resonance region, while in ENDF / B-VII.b2, the resonance parameters of Muhabhab are used, estimated a decade earlier. It is recommended that ROSFOND accept the estimate from JEFF-3.1. The spectra of 8-groups of delayed neutrons are replaced by the corresponding spectra for uranium-235. Group outputs are, of course, aligned with JEFF-3.1. The file also includes data on fission product yields from ENDF / B-VII.b2 1 (other libraries do not contain data on fission product yields for uranium-232). In the future, it is desirable to perform new assessment neutron data. The author of the conclusion is Nikolaev M.N. Contents of the ROSFOND file for 92- U-232 Replace !! MF = 1 General and specific information about nuclide 1 T.R.England, B.F.Rider, ENDF-349,
2 MT = 451 heading section MT = 452 total average number of fission neutrons MT = 455 delayed fission neutrons MT = 456 average prompt fission neutrons MF = 2 Resonance parameters MT = 151 section of resonance parameters MF = 3 Neutron cross sections MT = 1 total cross section MT = 2 elastic scattering MT = 4 total cross section for inelastic scattering MT = 16 reaction (n, 2n) 92- U-231 MT = 17 reaction (n, 3n) 92- U-230 MT = 18 all fission processes MT = inelastic scattering with excitation of discrete levels MT = 91 inelastic scattering with excitation of a continuum of levels MT = 102 radiative capture: reaction (n, gamma) 92- U-233 MT = 251 average cosine of the elastic scattering angle in lab. coordinate system MF = 4 Angular distributions of secondary neutrons MT = 2 elastic scattering MT = 16 reaction (n, 2n) 92- U-231 MT = 17 reaction (n, 3n) 92- U-230 MT = 18 all fission processes MT = inelastic scattering with excitation of discrete levels MT = 91 inelastic scattering with excitation of a continuum of levels MF = 5 Energy distributions of secondary neutrons 2
3 MT = 16 reaction (n, 2n) 92- U-231 MT = 17 reaction (n, 3n) 92- U-230 MT = 18 MT = 91 all fission processes inelastic scattering with excitation of a continuum of levels Uranium-233 Radioactive. (T 1/2 = 1.592 * 10 5 years). Alpha decays to thorium-229 (T 1/2 = 7880 years). It is a promising nuclear fuel (the basis of the uranium-thorium fuel cycle). Modern libraries contain the following data estimates for uranium-233. FUND-2.2 and BROND-2 assessment by Sukhoruchkin and Klepatsky, ENDF / B-VII.b2- assessment by Young, Chadwick, Talou, Leal, Derrien, JENDL-3.3 and JEFF-3.1 assessment by T. Mutsunobu, T. Kawano, In addition, the recent (2005) assessment of V. Maslov is available. 1. The region of thermal neutrons and the region of allowed resonances. Table 1 shows the estimated thermal cross sections and resonance integrals of capture and fission, as well as the number of prompt fission neutrons in comparison with the estimates of the experimental data by Muhabhab and Telier, as well as with the consensus estimate of the thermal cross sections performed by the international standards group in 2005.2. the assessment took into account any differences in the reference values used to obtain the final results. Table 1. Thermal sections and resonance integrals. Source σ с (RI c σ f (ev) RI f ν t ev) FUND ENDF / B-VII.b JENDL Maslov Muhabhab ± ± ± ± 0.004 Tellier ± ± ± ± 17 Standards ± ± ± the estimated data on the cross sections and resonance integrals do not go beyond the estimated errors of the set of experimental data. Descriptions of the region of allowed resonances differ significantly. According to the estimates of Sukhoruchkin and Klepatsky, this region extends to 100 eV, contains 178 resonances, the energy of the latter eV. In the future, this estimate will not be viewed as clearly outdated. 2 Data provided by members of the international group from Russia V. Pronyaev, S. Badikov and E. Gai 3
4 In the estimate of Mitsunobi and Kavanaugh, the boundary of the region of allowed resonances is -150 eV. The parameters of 190 resonances with the maximum energy eV are given. In the estimate adopted in ENDF / B-VII.b2, the boundary of the region of allowed resonances is 600 eV; in this region the parameters of 738 resonances are given. In addition, the parameters of 16 bound states and 16 resonances lying above this region are set. This estimate was also accepted by Maslov. The estimation of the resonance parameters was carried out taking into account new measurements of the total cross section and the fission cross section, carried out with a very high resolution at the ORELA accelerator in using famous program SUMMY, which describes the set of experimental data by the least squares method based on the R-matrix formalism 3. Figure 1 shows the increasing sum of the number of resonances, and Fig. 2 is the growing sum of the reduced neutron widths. Thin lines show linear approximations to the initial portions (up to 400 eV) of these curves. Fig. 2, we can conclude that there is practically no missing resonances in the region under consideration. Figure 2 shows that in the eV interval there is a shortage of reduced neutron widths, and then, above 500 eV, the same rate of increase in the sum of widths remains. The lack of resonances with large widths, of course, is not evidence of missing levels, but raises doubts about the correctness of determining the resonance parameters in the indicated interval. Despite this, the estimate of the resonance parameters from ENDF / B-VII.b2 is undoubtedly the most complete and reliable, and this estimate should be accepted by ROSFOND. Number of resonances Energy, eV ENDF / B-V II Fig.1. The growing sum of the number of resonances 3 LC Leal, H. Derrien, JA Harvey, KH Guber, NM Larson and RR Spencer, R-Matrix Resonance Analysis and Statistical Properties of the Resonance Parameters of U-233 in the Neutron Energy Range from Thermal to 600 ev , ORNL / TM-2000/372, March
5 C ummah<Гn0>"Energy, eV Fig. 2. Sum of reduced neutron widths. 2. Region of unresolved resonances. ENDF / B-VII In ENDF / B-VII.b2, the region of unresolved resonances extends up to 40 keV; d-waves; the file of average resonance parameters is recommended only to take into account the resonant self-shielding of cross sections, the mean cross sections themselves are specified in the file MF = 3. In JENDL-3.3 (and therefore in JEFF-3.1), the region of unresolved resonances extends only up to 30 keV; the parameters are set only s- and p-waves, but these parameters are recommended for calculating not only the self-shielding factors, but also the mean cross sections.In Maslov's estimate, the region of unresolved resonances extends to the threshold of inelastic scattering keV. This is an obvious advantage of Maslov's estimate, however, it is necessary to consider how the calculated or given average cross sections agree with the available experimental data. full cross section are compared with experimental data. In JENDL-3.3, experimentally installed gross U-233 Total URR + Fast Region Cross section, barn ENDF / B JENDL = JEFF MASLOV Fulwood57 Stupegia62 Pattenden E + 02 1.E + 03 1.E + 04 1.E + 05 Energy, ev Fig. 3. Total cross section in the region of unresolved resonances 5
6, the structure of the total cross section is reproduced by varying the average parameters of the distances between resonances and neutron widths (for all values of J and parity). Maslov does not introduce these artificial variations, and therefore he did not show any structure of the middle section. In general, the average cross section in this estimate is approximately by barn (~ 5%) lower than in the two previous ones, which, however, does not go beyond the scatter of the experimental data. Let us now consider the data on partial cross sections. In fig. 4 compares the estimated fission cross sections Cross section, barn U-233 Fission URR ENDF / B JENDL = JEFF MASLOV Guber2001 Nizamuddin E + 02 Energy, ev 1.E + 03 Fig. 4a. Fission cross section in the region of unresolved resonances 15.0 Cross section, barn U-233 Fission URR JENDL = JEFF Guber2001 Nizamuddin74 ENDF / B MASLOV E + 03 Energy, ev 1.E + 04 Fig.4b. Fission cross section in the region of unresolved resonances 5.0 U-233 Fission URR + Fast Region Cross section, barn JENDL = JEFF Guber2001 Nizamuddin74 MASLOV ENDF / B E + 04 Energy, ev 1.E + 05 Fig.4c. Fission cross section in the region of unresolved resonances 6
7 The presentation of data in the cited works is excessively detailed: the scatter of points does not reflect either the detailed resonance structure, for this the resolution is insufficient, nor the gross structure. In fig. 4d, the estimated data are compared with experimental data in a narrow range from 600 to 800 eV. The experimental data were averaged over subintervals, and the averaging results were presented as histograms. As can be seen, the gross structure of the fission cross section displayed in the ENDF / B-VII.b2 and JENDL-3.3 estimates only qualitatively reflects the results of measurements that do not agree with each other in details. This casts doubt on the expediency of describing the structure of the fission cross section in this energy range Cross section, barn ENDF / B JENDL = JEFF MASLOV 5.0 Guber2001 Nizamuddin74 Guber2001 Nizamuddin E + 02 7.E + 02 8.E + 02 Energy, ev Fig.4d. Fission cross section in the region of unresolved resonances In Fig. 5, the estimates of the capture cross section are compared with the Weston data, which are the only ones available in EXFOR in the region of unresolved resonances. The estimate adopted in ENDF / B-VII.b2 clearly overestimates the capture cross section. In the description of the file, there are no references to any additional experimental data in this area. In connection with all the above, it seems appropriate to include in ROSFOND the Maslov estimate of the data in the unresolved resonance region U-233 Capture URR + Fast Region ENDF / B JENDL = JEFF MASLOV Cross section, barn Weston E E E E + 03 Energy, ev Fig. 5. Capture cross section in the region of unresolved resonances 7
8 3. Sections outside the resonance region In Fig. 6. Estimates of the total cross section are compared with the available experimental data. The discrepancies between the estimates are significantly less than the scatter of the experimental data, so we can state that they are all equally good. Cross section, barn ENDF / B MASLOV Green73a Poenitz83 Poenitz78 JENDL = JEFF Foster Jr71 Green73b Poenitz E E E E E + 06 Energy, ev 10.0 Fig. 6a. Full section. 9.0 Cross section, barn ENDF / B JENDL = JEFF 5.0 MASLOV Green73a Foster Jr71 Green73b 4.0 Poenitz83 1.E + 06 1.E + 07 Energy, ev Fig. 6b. Full section. Figure 7. the estimates of the fission cross section are compared with the experimental data. The situation here is not so good: the scatter of experimental data 8
9 Cross section, barn JENDL = JEFF Tovesson2004c Guber2001 Shcherbakov2001 MASLOV ENDF / B Meadows74 Poenitz E + 05 1.E + 06 1.E + 07 Energy, ev Fig. 7a. Division 9
10 Cross section, barn JENDL = JEFF Tovesson2004c Guber2001 Shcherbakov2001 MASLOV ENDF / B Meadows74 Poenitz E + 05 1.E + 06 1.E + 07 Energy, ev 2.8 Fig. 7b. Cross section division Cross section, barn JENDL = JEFF MASLOV Shcherbakov2001 ENDF / B Pankratov63 Medous Zasadny-84 Arlt-81 Alkhazov-83 Adamov E E E E E E + 07 Energy, ev Fig. 7c. Fission section. Much exceeds the errors attributed to them. As a result, the discrepancy between the estimated data and the experimental data reaches in the vicinity of 1 MeV and 8 MeV ± 5%. Below 175 keV, Maslov's estimate agrees better than others with experimental data; above, the ENDF / B-VII.b2 estimate has an advantage. Note, by the way, that when performing this estimate, the results of numerous measurements of the ratios of the fission cross sections of uranium-233 and uranium-235 were normalized to the standard uranium-235 fission cross section adopted in 2005 (and included in ROSFOND). Figure 8. the results of the estimates are compared with the only experimental data of Hopkins. The ENDF / B-VII.b2 data goes directly to the experimental points; the other two estimates differ from them by an amount of the order of the error. Experimental data on inelastic neutron scattering on uranium-233 are absent. Figure 9 compares the results of the discussed evaluations. Near the threshold, the differences between them are very large. The minimum in the total cross section for inelastic scattering in the ENDF / B-VII.b2 estimate lies at 700 keV, i.e. just at the threshold of inelastic scattering with excitation of a continuous spectrum of levels adopted in this estimate. In two other estimates, this threshold is 100 keV lower. To clarify the situation, Fig. 8 shows the total cross section for inelastic scattering from the uranium-233 file from ENDF / B-VI. It 10
11 is significantly lower than modern estimates, but as in them, no peak is observed at the threshold. 1.E + 00 Cross section, barn 1.E-01 1.E-02 ENDF / B JENDL = JEFF MASLOV Hopkins62 1.E-03 1.E + 04 1.E + 05 1.E + 06 1.E +07 Energy, ev Fig. 8. Gripping section 2.0 U-233 Inelastic 1.5 Cross section, barn E E E E + 07 Energy, ev Fig. 9. Total cross section for inelastic scattering Cross section, barn ENDF / B-VII MT = 3 ENDF / B-VII MT = 2 JENDL-3.3 MT = 2 Maslov MT = 2 Maslov MT = 3 U-235 MT = EEE E + 07 Energy, ev Fig. 10. Cross sections for elastic scattering (MT = 2) and total cross section for inelastic interactions (MT = 3) 11
12 In fig. 10 shows the estimated cross sections for elastic scattering and the total cross section for inelastic interactions 4. It can be seen that the anomaly in the cross section for inelastic scattering is reflected in the behavior of the total cross section for inelastic interactions, which differs significantly from Maslov's estimate. The presence of this anomaly, which does not occur for uranium-235 (the cross section of inelastic interactions for which is also shown for comparison in Fig. 10), raises doubts about the correctness of the estimate adopted in ENDF / B-VII.b2. Figure 11 shows data on the cross sections for the reactions (n, 2n) and (n, 3n). Cross section, barn ENDF / B (n2n) JENDL (n2n) MASLOV (n2n) ENDF / B (n3n) JENDL (n3n) MASLOV (n3n) E E E E + 07 Energy, ev Fig. 11. Cross sections for reactions (n, 2n) and (n, 2n). There are no differential experimental data for these reactions. The discrepancies in the estimates above 16 MeV are large. Indirectly in favor of the ENDF / B-VII.b2 estimate is the fact that it was carried out up to 30 MeV, where the role of the (n, xn) reactions is very significant and, undoubtedly, the calculation of their cross sections demanded increased attention from the evaluators. Reaction (n.4n) about 19 MeV. Its cross section even at 20 MeV is fractions of a millibar. When neutrons interact with uranium-233 at all energies, reactions (n, p) and (n, α) are possible. Because of the high Coulomb barrier, the cross sections of these reactions are small: even at 20 MeV, the cross section of the first of them, according to EAF-2003, is 70 mbar; the second is 5 millibarn. Nevertheless, it seems appropriate to include the cross sections for these reactions in ROSFOND. Summing up the above, we can conclude that the ROSFOND should accept the neutron cross sections estimated by Maslov, which, as a rule, being close to the estimate from ENDF / B-VII.b2, do not have anomalously high inelastic scattering cross sections in the region below 700 keV. 4. The numbers of secondary neutrons and their energy-angular distributions 4.1. Number of fission neutrons The estimated numbers of neutrons from fission of uranium-233 by thermal neutrons are given in Table 1. The value adopted in ENDF / B-VII.b2 exceeds the recommendation of the standards group (based on a joint assessment of all data dependent on ν p (233 U)) by three standard deviations assigned to this value. 4 Section МТ = 3 in JENDL-3.3 is not specified and it is not easy to obtain it, since the components are specified on different energy grids. For the same reason, the Maslovsky cross section MT = 3 is reduced only up to the reaction threshold (n, 2n). 12
13 This difference is exactly equal to the contribution of delayed neutrons adopted in this estimate: ν d = Thus, when evaluating the data for ENDF / B-VII.b2, the value recommended by the international standards group as ν t was considered as ν p. The JENDL-3.3 score is below the recommended value by 2.6 standard deviations. Maslov's estimate is also lower, but only by 1 standard deviation. It seems expedient to accept in ROSFOND the value recommended by the international standards group, i.e. ν t = The number of delayed neutrons estimated by ENDF / B-VII.b2 at low energies is; according to the JENDL estimate and practically the same according to Maslov. If we take ν d = 0.0068, then for ν p we obtain a "round" number. 12 shows the energy dependences of ν p according to different estimates in comparison with experimental data. All presented experimental data are renormalized either by ν р (252 Cf) = 3.7606, or by ν р (233 U; 0.0253 eV) = 2.490, depending on the used monitor NUbar ENDF / B JENDL 2.5 MASLOV Smirenkin-58 Nurpeisov-73 Nurpeisov- 75 Gwin-86 Kolosov-72 Standard EEEEEEEEEE E + 06 Energy, ev Fig. 12a. The number of prompt fission neutrons. The broken course of ν p with energy, adopted by Maslov, is not justified by experimental data. In general, up to 1.5 MeV, the ν p accepted in this estimate seems to be underestimated. At higher energies, the data are shown in Fig. 12b NUbar 4.0 ENDF / B JENDL 3.5 MASLOV Smirenkin Nurpeisov-73 Nurpeisov Gwin-86 Kolosov E E E E E E E E E E E E + 07 Energy, ev Fig. 12b. The number of prompt fission neutrons. 13
14 In this area, the ENDF / B-VII.b2 score appears to be the best. It is quite possible to accept it even at lower energies if we replace the value of ν p in the thermal region by (see Fig. 12a). In fig. 13 shows the estimated energy dependences of ν d. For comparison, those are given for uranium-235 and plutonium-239. The comparison shows that the energy dependence ν d adopted in ENDF / B-VII.b2 is erroneous. There are no physical grounds for this behavior. On the contrary, the decrease in ν d with energy, which manifests itself in all other estimates, is explained by the appearance of additional chances of fission. In ROSFOND, it is advisable to accept the energy dependence of ν d from JENDL-3.3, renormalizing it to the accepted value of ν d in the thermal region NUbar ENDF / B JENDL-3.3 Maslov U-235-ROSFOND Pu-239-ROSFOND EEEEEEEEEE E + 07 Energy, ev Fig. 13. Energy dependence of the delayed neutron yield 4.2. Fission neutron spectra. The spectra of prompt fission neutrons in the estimates under consideration are described in substantially different ways. In ENDF / B-VII.b2, these spectra are given by the Watt form with parameters a (e) and b (e) depending on the energy of neutrons E causing fission: 2exp (-ab / 4) χ (E) = exp (E / a) sh be πa 3 b The nature of this dependence can be seen from Fig. 14, which shows the dependence of the average energy of fission neutrons< E >= a (3/2 + ab / 4) as a function of E. The heading section states that the fission neutron spectra are assumed in accordance with the JENDL-3.3 assessment. This, obviously, is not entirely true, since in the JENDL-3.3 estimate, the prompt fission spectra are defined differently, namely, by functions specified at 164 points at each of the 7 initial energies. The fission spectra are determined in a similar way in Maslov's estimate, but the spectra are set at 326 points at each of the 22 initial energies in the range up to 20 MeV. fourteen
15 Average fission neutron energy 2.40 ENDF / B-VII,
16 1. The number of delayed neutrons emitted during fission by thermal neutrons is assumed to be equal (in JEFF-3.1 it is; in ENDF / B-VII.b, by Maslov). The energy dependence of this number is the same as in the JEFF-3.1 assessment (see Fig. 13). 2. The spectra of groups of delayed neutrons are taken the same as for uranium-235 (see n below) and for all other fissile nuclei. However, the outputs of each of the 8 groups are assumed to be the same as in JEFF-3.1, i.e. Based on the recommendations of the work Spectra and angular distributions of scattered neutrons and neutrons of reactions (n, xn) Figure 15 compares the estimated values of the first three moments of angular distributions of elastically scattered neutrons. The estimates are very close to each other. All of them were obtained by calculation. EXFORe contains the results of only one unpublished work by Haoaut-82, in which the angular distributions of neutrons with energies of 0.7 and 1.5 MeV were measured. At these energies, it is very difficult to distinguish elastically scattered neutrons from those inelastically scattered at low-lying levels. In the brief description given in EXFORe, the procedure for separating these processes is not described; it is only said that the correction for inelastic scattering introduced by the author was from 5 to 35% at both 0.7 MeV and 1.5 MeV. Since there are no discrepancies in the estimates at the indicated energies, and the experiment does not differ in high reliability, a rather laborious comparison with it was considered superfluous. In ROSFOND it is advisable to include the estimate from ENDF / B-VII.b2, which, as a rule, occupies an intermediate position Angular momentum value ENDF / B-VII 0.1 JENFF-3.1 Maslov E E E E + 07 Energy, eV Fig. 15. Angular moments of the distribution of elastically scattered neutrons: solid curves for the 1st moment (mean cosine of the scattering angle), dashed 2nd moment, dashed curves for the 3rd moment. 6 Spriggs, Campbel and Piksaikin, Prg Nucl Eng 41.223 (2002) 16
17 As for the spectra of inelastically scattered neutrons, below the level continuum excitation threshold, they are determined by the completeness of taking into account the excited levels of the target nucleus. In this respect, Maslov's estimate has a definite advantage over JENDL-3.3: it takes into account all the levels indicated in the PCNUDAT 2 database, while JENDL-3.3 does not describe the excitation of some levels with energies from 400 to 600 keV. In both estimates, the excitation of the level continuum is described starting from 600 keV, i.e. immediately following the region of discrete levels. We do not discuss the estimate adopted in ENDF / B-VII.b2 here because of the doubts it generates about the correctness of the description of the energy behavior of the total cross section for inelastic scattering (see Section 3 above). Spectra of neutrons scattered with the excitation of a continuum of levels Figure 16 shows the spectra of neutrons that underwent inelastic scattering with excitation of a continuum of levels of the target nucleus. Data are given for initial energies of 6 MeV, 10 MeV, and 14 MeV. At 6 MeV, i.e. below the reaction threshold (n, n f), the Maslov spectrum is much harder than the others: obviously, the fraction of pre-equilibrium neutrons emitted in it is higher. At 10 MeV, the estimates of neutron spectra differ significantly. In the spectrum adopted in JENDL-3.3, neutrons with energies below 3.7 MeV are absent at all; it is assumed that the emission of such slow neutrons is always followed by fission. In the ENDF / B-VII.b2 estimate, a tail of relatively slow neutrons is present, and in the Maslov estimate, this tail also exhibits a maximum in the region of the order of 1 MeV. At 14 MeV, the spectrum of JENDL-3.3 contains no neutrons with energies below 5 MeV, but the probability of emission of neutrons with energies of 6-8 MeV is significantly higher than in the other two estimates. The ENDF / B-VII.b2 and Maslovsky spectra above 7 MeV are close, but the Maslovsky spectrum has a long tail of slow neutrons. For some reason, after the emission of slow neutrons, neither the (n, 2n) reaction nor fission occurs. Probability / MeV 9.0EEEEEEEEEEEEEEEEEE E + 00 Spectra (n, n ") 0.0EEEEEEE E + 07 Energy, eV ENDF / B-VII; 6 MeV ENDF / B-VII; 10 MeV ENDF / B-VII; 14 MeV JENDL-3.3 ; 6 MeV JENDL-3.3; 10 MeV JENDL-3.3; 14 MeV Maslov; 6 MeV Maslov; 10 MeV Maslov; 14 MeV Fig. 16. Comparison of the spectra of neutrons inelastically scattered with the excitation of a continuum of levels.
18 In fig. 17 compares the estimates of the neutron spectra of the reaction (n, 2n) for two initial energies of 10 and 14 MeV. The differences in estimates are very large, especially at 14 MeV. The discrepancies indicate an unfavorable state of affairs with the assessment of the spectra, and, therefore, the cross sections of the processes occurring through different channels and in different ways(pre-equilibrium emission of neutrons and ordinary evaporation, fission after the emission of one or two neutrons in one way or another). Since there are no significant differences in the estimates of the total fission cross section, there is a compensation for the differences in the estimates of the contributions of different reaction mechanisms. Spectra (n, 2n) Probability / MeV 1.0E E E E E E E E E E-07 ENDF / B-VII; 10 MeV ENDF / B-VII; 14 MeV JENDL-3.3; 10 MeV JENDL-3.3; 14 MeV Maslov; 10 MeV Maslov; 14 MeV 0.0E E E E E E E E E E + 06 Energy, eV Fig. 17. Comparison of the neutron spectra from the reaction (n, 2n). It can be seen from what has been considered that the assessment of the spectra of continual reactions in ENDF / B-VII.b2 is in a sense intermediate and this gives rise to the temptation to choose it for ROSFOND. However, further validation of a composite file, in which cross sections are taken from one estimate and spectra from another, can cause problems. Since it was decided to take the cross sections from, the spectra should also be taken in accordance with this estimate. Note that the data on the spectra in ENDF / B-VII.b2 are presented (in contrast to the other two) in the file format MF = 6, i.e. spectra are given taking into account correlations between energy and scattering angle. This correlation, however, is described oversimplified according to the semiempirical Kalbach-Mann taxonomy. In addition to the neutron spectra, the spectra of recoil nuclei are also described (for which there is no practical reason), but the spectra of photons emitted in continuous processes are not described. This is one more indication of the unfavorableness of the assessment, which should be eliminated in the future, when the assessment is revised. 5. Data on the production of photons in neutron reactions Neither the Maslov estimate nor the JENDL-3.3 estimate provides data on the production of photons. JEFF-3.1 includes photon production data from ENDF / B-VI (estimated by Stuart and Weston 1978). In ENDF / B-VII.b2 with revised data on gamma radiation during radiative capture. So 18
In this way, there is practically no choice of estimates. Consider what the available estimates are based on. Total inelastic scattering: MT = 4. Since, in the estimate by Stuart and Weston, the excitation of only the first four levels of the target nucleus was individually taken into account, transitions only between these four levels are described in the photon spectrum. The spectrum of photons formed upon excitation of the continuum is described by a continuous spectrum of photons, which is assumed to be the same as for plutonium. Above 1.09 MeV, the multiplicity for MT = 4 is assumed to be zero. The possibility of a more correct description of the photon spectra, which opened up in connection with the explicit description of a much larger number of levels (28 in ENDF / B-VII.b2, 25 in Maslov, 25 in JENDL-3.3) has nowhere been realized. Fission photons: multiplicity up to 1.09 MeV corresponds to the Hoffmanov estimate 8; the spectra themselves are taken the same as for plutonium. Above 1.09 MeV, the multiplicity is assumed to be zero. The multiplicity of emission of photons during capture below 1.09 MeV is arbitrarily assumed to be equal. The spectrum is assumed to be the same as for plutonium-239 with correction for the difference in reaction energies. Above 1.09 MeV, the cross section for photon production in inelastic interactions (file MF = 13) and the normalized spectrum (file MF = 15) are the same as for plutonium In ENDF / B-VII.b2, the multiplicity of emission of photons during capture and their spectra are calculated from the GNASH program. All other data is assumed to be the same as described above, i.e. from ENDF / B-VI.7. ROSFOND should include data on photon production from ENDF / B-VII.b2. With further revisions of the file and, especially, in the case of a decision to include the file MF = 6, a more correct calculation of the photons produced in neutron reactions should be carried out. Conclusion Based on the above, it seems expedient to create a combined file for ROSFONDA as follows. 1. Files MF = 2 and MF = 3 are taken from Maslov's estimate. In the region of allowed resonances, they coincide, as noted. 2. Take the energy dependence of fission neutrons in accordance with ENDF / B-VII.b2, replacing the value at thermal energy by i.e. so that the total number of fission neutrons coincides with the value recommended by the standards group. and (n, alfa). 4. The number of delayed fission neutrons at the thermal point is assumed to be equal, and its energy dependence in accordance with the JEFF estimate. Accept also the 8-group description of delayed neutrons from JEFF. JEFF ENDF / B-VI. 7, MAT = D. C. Hoffmann and M. M. Hjffmann, Ann. Rev. Nucl. Sci. 24, 151 (1974) 19
20 6. Take the angular distributions of elastically scattered neutrons in accordance with the ENDF / B-VII.b2 estimate, the rest of the angular distributions in accordance with Maslov's estimate. 7. The spectra of prompt fission neutrons and the continual spectra of other reactions should be taken in accordance with Maslov's estimate. 8. Include data on fission product yields in accordance with R. Mills' estimate (JEFF). 9. Accept data on the production of photons in neutron reactions in accordance with ENDF / B-VII.b2. The author of the recommendation is Nikolaev M.N. File content 20
21 92.3. Uranium-234 Content in the natural mixture% Radioactive. (T 1/2 = 2.455 * 10 5 years). Alpha decays to thorium-230 (T 1/2 = 7.54 * 10 4 years). The current libraries contain the following data estimates for uranium-233. FUND-2.2 estimate by T.Ohsawa, M.Inoue, T.Nfkagawa, 1987 ENDF / B-VII.b2 - estimate by Young, Chadwick, JENDL-3.3 estimate by T. Watanabe, 1987 JEFF-3.1 estimate by Maslov, In the estimates adopted in ENDF / B-VII. b2 and in JEFF-3.1 the boundary of the region of allowed resonances, which contains 118 resonances and one bound state, is 1500 eV. The positions of the resonances are exactly the same. The resonance widths, however, differ. In ENDF / B-VII.b2 they correspond to the data of Muhabhab-84; Maslov uses a later estimate from JENDL-3.2. In fig. 1 shows the increasing sum of the number of resonances, in Fig. 2 is the sum of the reduced neutron widths. From the graphs it can be concluded that above 900 eV, some of the resonances are missed, but the missed resonances have small widths and their omission should not significantly affect the calculated cross sections. Number of resonances Energy, eV Fig. 1. The growing sum of the number of resonances Sum<Гn0>"ENDF / B-VII Maslov Energy, eV Fig. 2. Sum of reduced neutron widths 21
22 From fig. 2 that, in Maslov's estimate, the neutron widths are taken to be smaller than in ENDF / B-VII.b2 (by about 12%). Radiation widths, on the contrary, are larger, on average by 45%. The dividing widths are practically the same. Both estimates contain regions of unresolved resonances described by the parameters of the s, p, and d waves. In Maslov's estimate, these parameters vary greatly with energy, describing the gross-structure of the cross sections. The result can be seen from Fig. 3 and 4, which compare the capture and fission cross sections above the region of allowed resonances. 1.00E E + 00 Maslov, capture ENDF / B-VII, capture Muradyan-99 Cross-section, barn 1.00E E E E E E E E E E + 07 Energy, eV Fig. 3. Capture section 1.00E E + 00 Section, barn 1.00E E-02 James-77 Medous-78 Maslov, division 1.00E-03 ENDF / B-VII, division 1.00E E E E E + 07 Energy, eV Fig. 4. Fission section. The increased capture cross section in Maslov's assessment is justified by the only available result of Muradyan. The subthreshold division structure reflected in Maslov's assessment reflects the results of James. Conclusion It is recommended that ROSFOND accept Maslov's estimate from JEFF-3.1. The spectra of 8 groups of delayed neutrons should be taken the same as those of uranium-235. Outputs 22
The 23 fission products of uranium-234 are contained in ENDF / B-VI (England and Reeder 1989) and in JEFF-3.1 (Mills, 2005). It is natural to accept the last estimate. The cross sections of the main reactions in the integral spectra are given in the following table Total Elastic Inelastic (n, 2n) (n, f) (n, γ) ev Resonance integral Fission spectrum 235 U MeV Author of the conclusion Nikolaev M.N. Contents of the file ROSFOND for 92- U-234 Remake !! MF = 1 General and specific information on the nuclide MT = 451 heading section MT = 452 total average number of fission neutrons MT = 458 energy release during fission MF = 2 Resonance parameters MT = 151 section of resonance parameters MF = 3 Neutron cross sections MT = 1 total cross section MT = 2 elastic scattering MT = 4 total cross section for inelastic scattering MT = 16 reaction (n, 2n) 92- U-233 MT = 17 reaction (n, 3n) 92- U-232 MT = 18 all fission processes MT = 19 fission ( first chance) MT = 20 fission (second chance) - reaction (n, nf) - U- MT = 21 fission (third chance) - reaction (n, 2nf) - U- MT = inelastic scattering with excitation of discrete levels MT = 91 inelastic scattering with excitation of a continuum of levels MT = 102 radiative capture: reaction (n, gamma) 92- U-235 MF = 4 Angular distributions of secondary neutrons MT = 2 elastic scattering MT = 16 reaction (n, 2n) 92- U-233 MT = 17 reaction (n, 3n) 92- U-232 MT = 18 all fission processes MT = 20 fission (second chance) - reaction (n, nf) - U- MT = 21 fission (tert chance) - reaction (n, 2nf) - U- MT = inelastic scattering with excitation of discrete levels 23
24 MT = 91 inelastic scattering with excitation of a continuum of levels MF = 5 Energy distributions of secondary neutrons MT = 16 reaction (n, 2n) 92- U-233 MT = 17 reaction (n, 3n) 92- U-232 MT = 18 all processes divisions MT = 19 division (first chance) MT = 20 division (second chance) - reaction (n, nf) - U- MT = 21 division (third chance) - reaction (n, 2nf) - U- MT = 91 inelastic scattering with excitation of a continuum of levels MT = 455 fractions of groups and delayed neutron spectra MF = 8 Yields and decay characteristics of the resulting radionuclides MT = 16 reaction (n, 2n) 92- U-233 MT = 17 reaction (n, 3n) 92- U-232 MT = 102 radiation capture: reaction (n, gamma) 92- U-235 MT = 457 radioactive decay data 24
25 92.4. URANUM General characteristics 1.1. Z = A = ± Aw = ± Content in the natural mixture: 0.72 at%; weight% 1.5. List of neutron reactions 9 MT Reaction Q, MeV E threshold, MeV Nucleus-product *) 234 U 16 (n, 2n) (n, 3n) U 37 (n, 4n) U 19 (n, f 1) FP + n + γ 20 (n, nf 2) FP + n + γ 21 (n, 2nf 3) FP + n + γ 38 (n, 3nf 4) FP + n + γ 102 (n, γ) U 103 (n, p ) Pa 107 (n, α) Th 1.6. Radioactivity: Half-life: 7.038 * 10 8 years. Alpha decay probability: Spontaneous fission probability: 2 * 10-8 Decay energy Q α = 4.678 MeV; Q sf = Resonant region: (MF = 2) 2.1. Region of allowed resonances General characteristics of the region of allowed resonances 9 In the energy range under consideration, other reactions with the emission of charged particles are possible - (n, d), (n, t), (n, 3 He), etc. - including exoenergetic ones, - (n, 2α), (n, nα), - the cross sections of which, however, are very small and are not given in the file of the estimated data. 25
26 Spin and parity of the target nucleus: 7/2 - Scattering radius: R = 0.9602 * cm does not depend on energy. It is used only for calculating the potential barrier permeabilities and scattering phases. Resonant formula: Reich-Mura. The calculation of the scattering anisotropy from the resonance parameters is not envisaged The number of orbital moments is one (namely, l = 0, i.e., only s-resonances are considered) The number of systems of resonances with different spins J: two (J = 3 and J = 4) Boundaries of the region of allowed resonances : from 10-5 eV to 2250 eV The number of considered resonances is 3193; of these, 14 are below the binding energy of a neutron and 9 are above the boundary of the region of allowed resonances. The number of resonances with J = 3 is 1449; of which 1433 are in the region from 0 to 2250 eV. The number of resonances with J = 4 is 1744; of which 1732 are in the range from 0 to 2250 eV. Details of the estimate Below is a translation of the description of the estimate of the resonance parameters given in the header section of the data file for uranium-235 from the ENDF / B-VI revision 5. This estimate was carried out at the Oak Ridge Laboratory. Lealom et al. In 1997, adopted in all evaluated neutron data libraries for uranium-235 from ENDF / B-VI (Rev.5). It is also included in the ENDF / B-VII.b2 library. The resonance parameters were estimated by the least squares method taking into account the results of both differential measurements of neutron cross sections and integral experiments. The thermal cross sections (fission, capture and elastic scattering) and the Westcott g-factors from the ENDF / B-6 10 neutron standards file, as well as the K1 factor estimated by Hardy 11 were used as input parameters. only based on the results of differential experiments, and then taking into account the integral data, are compared with the input data of the SAMMY program. The value of ν obtained as a result of fitting to the listed parameters turned out to be ± Poenitz, G.M. Hale et al., "The ENDF / B-6 Neutron Cross Section Measurements Standards," National Institute of Standards and Technology report NISTIR (1993) 11 J. Hardy, Brookhaven National Laboratory, report BNL-NCS (1979) Sec. B.1. 26
27 Table 1. Thermal parameters. Parameter Input value Diff fit only. data Fission cross section ± Capture cross section 98.96 ± Scattering cross section 15.46 ± g f ± g a ± g γ K ± Diff. and integra. data Table 2. Calculated and experimental values of integrals of the fission cross section (barn * eV) Energy range, eV Calculation based on res. Experimental data from to parameters Shark88 Weston84 Weston Table 3. Calculated and experimental values of integrals of the capture cross section (barn * eV) Energy range, eV Calculation based on res. Experimental data from to parameters desaussure67 Perez Resonance integrals of fission and capture calculated from the estimated resonance parameters are equal, respectively, barn and barn, which leads to 27
28 to an alpha value of 0.509, which is in excellent agreement with the data of integral experiments. When evaluating the resonance parameters, the data of the following differential experiments were taken into account. 1. Experiments by Harvey88 on transmission at the ORELA accelerator on an 18-meter flight path with a sample atomic / barn thick cooled to 77K (from 0.4 to 68 eV). 2. Experiments by Harvey88 on transmission at the ORELA accelerator on an 80-meter flight path with a sample atomic / barn thick cooled to 77K (from 4 to 2250 eV). 3. Harvey88 transmission experiments on the ORELA accelerator on an 80-meter flight path with a sample atomic / barn thick cooled to 77K (from 4 to 2250 eV). 4. Measurements of the Schark88 fission cross section on the RPI accelerator on a flight path of 8.4 m (from 0.02 to 20 eV). 5. Measurements of the fission and capture cross section desaussure67 on the ORELA accelerator on a flight base of 25.2 m (from 0.02 to 2250 eV). 6. Measurements of the fission and capture cross section of Perez73 on the ORELA accelerator on a flight base of 39 m (from 0.01 to 100 eV). 7. Measurements of the Gwin84 fission cross section at the ORELA accelerator on a flight path of 25.6 m (from 0.01 to 20 eV). 8. Spencer84 transmission experiments at the ORELA accelerator on an 18-meter flight path with a sample of atoms / barn thick (from 0.01 to 1.0 eV). 9. Measurements of the Wagemans88 fission cross section at the GELINA accelerator on an 18-meter flight base (from up to 1.0 eV) 10. Measurements of the Gwin96 absorption and fission cross sections at the ORELA accelerator (from 0.01 to 4 eV). 11. Measurements of the Weston84 fission cross section at the ORELA accelerator on an 18.9-meter flight base (from 14 to 2250 eV). 12. Measurements of η Wartena87 on an 8-meter span (from up to 1.0 eV). 13. Measurements of the η Weigmann90 value on a mechanical breaker (from up to 0.15 eV) 14. Measurements of the Weston92 fission cross section at the ORELA accelerator on an 86.5-meter flight base (from 100 to 2000 eV). 15. Measurements of the Moxon92 fission cross section at the ORELA accelerator (from 0.01 to 50 eV) References to the used experimental works. Index Reference Harvey88 J.A. Harvey, N.W. Hill, F.G. Perey et al., Nuclear Data for Science and Technology, Proc. Int. Conf. May 30-June 3, 1988, Mito, Japan. (Saikon Publishing, 1988) p. 115 Schark88 R.A. Schrack, "Measurement of the 235U (n, f) Reaction from Thermal to 1 kev," Nuclear Data for Science and Technology, Proc. Int. Conf. May 30 - June 3, Mito, Japan (Saikon Publishing, 1988) p. 101 desaussure67 G. de Saussure, R. Gwin, L.W. Weston, and R.W. Ingle, "Simultaneous Measurements of the Neutron Fission and Capture Cross Section for 235U for Incident Neutron Energy from 0. 04 ev to 3 kev, "Oak Ridge National Laboratory report ORNL / TM-1804 (1967) Perez73 R.B. Perez, G. de Saussure, and E.G. Silver, Nucl. Sci. Eng. 52, 46 (1973) 28
29 Gwin84 R. Gwin, R.R. Spencer, R.W. Ingle, J.H. Todd, and S.W. Scoles, Nuc.Sci.Eng. 88, 37 (1984) Spencer84 R.R. Spencer, J.A. Harvey, N.W. Hill, and L. Weston, Nucl. Sci. Eng. 96, 318 (1987) Wagemans88 C. Wagemans, P. Schillebeeckx, A.J. Deruyter, and R. Barthelemy, "Subthermal Fission Cross Section Measurements for 233U and 239Pu," Nuclear Data for Science and Technology, Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon Publishing, 1988) p. 91 Gwin96 R. Gwin, To be published in Nuclear Science Engineering Weston84 L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 88, 567 (1984) Wartena87 J.A. Wartena, H. Weigmann, and C. Burkholz, report IAEA Tecdoc 491 (1987) p. 123 Weigmann90 H. Weigmann, P. Geltenbort, B. Keck, K. Shrenckenbach, and J.A. Wartena, The Physics of Reactors, Proc. Int. Conf., Marseille, 1990, Vol. 1 (1990) p. 133 Weston92 L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 111, 415 (1992) Moxon92 M.C. Moxon, J.A. Harvey, and N.W. Hill, private communication, Oak Ridge National Laboratory (1992) Discussion of the results of estimating the parameters of allowed resonances Note, first of all, that in 1985 the same group of evaluators, based on the same experimental data, using the same SAMMY program, estimated the parameters of allowed resonances. resonances of uranium-235 in the same energy region 12. However, at that time, due to limited computer capabilities, the energy region under consideration had to be divided into 5 intervals. The results of the evaluation were accepted into the ENDF / B-VI library. 2, to the FOND-2 library and to many other libraries of evaluated data. In fig. 1 compares the multi-group cross-sections calculated on the basis of the 1985 and 1997 estimates. The diagrams show deviations of cross-sections calculated according to ENDF / B-VI (Rev.5) from those calculated according to ENDF / B-VI (Rev.2) in percentage ENDF / B-VI (Rev.5 / Rev.2) capture, % fission,% alfa,% ENDF / B-VI (Rev.5 / Rev.2) capture,% fission,% alfa,% Discrepancy,% Discrepancy,%, 5 5.5 10.5 15.5 Energy, eV Fig.1a Energy, eV Fig.1b 12 NMLarson, ORNL / TM-9719 / R1, (1985) 29
30 Discrepancy,% ENDF / B-VI (Rev.5 / Rev.2) fission,% capture.% Alfa,% Energy, EE ENDF / B-VI (Rev.5 / Rev.2) fission,% capture.% alfa,% Energy, eV Discrepancies,% Fig.1c Fig.1 d. As can be seen, the overestimation effect turned out to be very significant: the capture cross section and its relation to the fission cross section increased significantly. It must be said that it was this increase that sharply reduced the calculated and experimental discrepancies in the criticality of aqueous solutions of highly enriched uranium, reducing them to an insignificant level. The reason for such a large change in the estimated data was not explained by the authors of the assessment. In the header section of the data file from ENDF / B-VI (Rev.2), it is noted that not all resonances are allowed above 110 eV. In a similar section from ENDF / B-VI (Rev.5) and more later versions library ENDF / B, such a clause is not contained (see section above). Therefore, it is of interest to consider how complete the set of resonances contained in the last estimate is. In fig. 2 shows the energy dependence of the level density with J = 3 and J = 4. The level density is expressed in the number of resonances per 100 eV Number of resonances per 100 eV N (J = 3) N (J = 4) Energy, eV Fig. 2 Energy dependence of the level density As can be seen, with an increase in energy up to 1000 eV The "observed" level density decreases monotonically, decreasing by half. This is followed by a jump upward by about one and a half times, followed by a monotonic decline again to about the previous level of 2000 eV. At this energy, the level density again increases abruptly to almost the initial value, followed by another drop, this time quite 30
98. CALIFORNIUM The main interest in neutron cross sections for californium isotopes was associated with the production of 5 Cf as a compact neutron source used in various fields. In this case, the original product
53. Iodine Remark on the assessment of data quality for fission fragments Taking into account that heavy isotopes of iodine are important fission products, we will make general remarks on the priorities for data quality. Most
32. GERMANIUM Natural germanium contains 5 isotopes: 70 Ge, 72 Ge, 73 Ge, 73 Ge and 76 Ge (the latter is weakly radioactive). In addition, there are three more long-lived radioisotopes: 78 Ge, 79 Ge and 71 Ge. For stable
12. MAGNESIUM Magnesium does not have long-lived radioactive isotopes. For three stable isotopes there are estimates by V. Hatchya and T. Asoni (1987), adopted in FOND-2.2 from JENDL-3.2. At 21, Shibata contributed to these estimates
45 RODIUM 45.1 Rhodium-99 Radioactive (T 1/2 = 16.1 days). Capturing an orbiting electron turns into stable ruthenium-99. In reactors, it can be formed in trace amounts due to the reaction 102Pd
14. SILICON General remarks. Natural silicon contains three stable isotopes in the following atomic concentrations: 28 Si 92.23%; 29 Si 4.67%; 30 Si - 3.10%. In addition, there is a beta-active isotope
37 RUBID 37.1. Rubidium-83 Radioactive (T 1/2 = 86.2 days). Capturing an orbiting electron turns into stable krypton-83. Possible reactions of 85 Rb (n, 3n) formation; 85 Rb (n, 2n) 84 Rb (n, 2n); 84
55. CESIUM Consideration of the state of affairs on neutron data for all cesium isotopes was carried out by VG Pronyaev. He also gave recommendations on the inclusion of the evaluated data files in ROSFOND. Footnotes
35. BROMINE 35.1. Bromine-79 Content in the natural mixture is 50.69%. Output at dividing 235 U 2.5 * 10-7; when fission 239 Pu 8.6 * 10-4. In modern libraries of evaluated data, two evaluations are used:
30. ZINK FOND-2.2 contains a data file for natural zinc (Nikolaev, Zabrodskaya, 1989) for calculating neutron transport. Data for all stable isotopes (Nikolaev, 1989) and Grudzevich data,
18. ARGON FOND-2.2 contained neutron cross sections for stable and radioactive isotopes of argon from EAF-3, as well as a complete dataset for natural argon (Howerton, 1983, from ENDL-84).
33. ARSENIC 33.1. Arsenic-71 is radioactive (T 1/2 = 65.28 h.) Capturing an orbital electron, it turns into germanium-71, which decays in the same way (T 1/2 = 11.43 days) into stable gallium-71. In reactors
51. Antimony Consideration of the state of affairs on the basis of neutron data for all isotopes of antimony was carried out by VG Pronyaev. He also gave recommendations on the inclusion of the evaluated data files in ROSFOND. Footnotes
49. INDIUM 49.1. Indium-111 Radioactive (T 1/2 = 2.8047 days). Undergoing capture of an orbital electron, it turns into stable cadmium-111. In reactors can be formed in trace amounts due to
50. TIN Possessing the magic number of protons (50), tin has the largest number of stable isotopes (10). Difficulties in the model description of cross sections at energies below several MeV are due to the low density
20. CALCIUM IN FUND-2.2 the complete data set is contained only for natural calcium. For stable and radioactive isotopes, the estimates of neutron cross sections from eaf-3 are accepted. ENDF / B-VII contains data only
5. FILE 5. ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS 1 5.1. GENERAL DESCRIPTION File 5 contains data for the energy distributions of secondary neutrons presented in the form of distributions of normalized
9. POTASSIUM IN FUND-2.2 the complete data file is contained only for natural potassium (H. Nakamura, 987). For stable and long-lived isotopes, the EAF-3 estimate is accepted. ENDF / B-VII contains data for natural
9. FLUORINE Fluorine does not have long-lived radioactive isotopes. ROSFOND includes data for the only stable isotope 19 F. 9.1. Fluorine-19 The -VIIb2, JEFF-3.1 and FOND-2.2 libraries use the estimate
79. GOLD 79.1. Gold-194 Radioactive (T 1/2 = 38.0 h.). Decays by capturing an orbiting electron into stable platinum-194. Possible ways formation in the reactor - ternary reaction 197 Au (n, 2n)
75. RENIUM 77.0 General Notes This section describes the isotopes of rhenium: two stable and seven radioactive isotopes with a half-life of more than a day. 75.1. Rhenium-182. Radioactive Experiencing the capture of an orbital
52. TELLURUS 52.1. Tellurium-118 Half-life: (6 ± 2) days. Decay modes: e - 100%. Ground state spin: 0+. JEFF-3.1 / A = EAF-2003 incomplete 2003 file estimate for activation library based
16. SULFUR The ROSFOND presents data for all 4 stable isotopes of sulfur and for radioactive sulfur-35 16.1. Sulfur-32 Content in the natural mixture 92% is the main isotope. In all modern libraries
71. LUTHESIUM 71.1. Lutetium-169 Radioactive (T 1/2 = 1.42 days). Experiencing the capture of an orbital electron, it turns into ytterbium-169, which, in turn, is converted in the same way (T 1/2 = 32.026 days)
80. MERCURY 80.0. General remarks In the FOND-2.2 library, all neutron data for 13 stable and long-lived mercury isotopes were taken mainly from the EAF-3 library. Complete neutron data files
76. OSMIUM ROSFOND should have provided complete sets of neutron data for 7 stable isotopes of osmium and data on the cross sections of neutron reactions for 5 long-lived radioactive isotopes. Unfortunately,
Half-life: (2.43 ± 0.05) days. Decay modes: e - 100%. Ground state spin: 0+. 56.BARIUS 56.1. Barium-128 JEFF-3.1 / A incomplete 2003 file estimate for activation library based on
34. SELENIUM 34.1. Selenium-72 Radioactive (T 1/2 = 8.4 days) Experiencing the capture of an orbital electron turns into arsenic-72, and the latter, emitting a positron (T 1/2 = 26 hours) into germanium-72. In insignificant quantities can
67. HOLMIUM Natural holmium contains only one isotope, 165 No. In addition, there is one very long-lived neutron-deficient isotope - 165 Ho (4570 years) and one neutron-abundant isotope - 165 Ho (26.8 hours),
4. BERYLLIUM The ROSFOND library contains data for three isotopes of beryllium: radioactive 7 Be (53.29 days), stable 9 Be and radioactive 10 Be. 4.1. Beryllium-7 is Radioactive. T 1/2 = 53.12 d. Capture
91. PROTACTINIUM Protactinium possesses five long-lived isotopes, data for which should be submitted to the ROSFOND library. 91.1. Protactinium-229 Radioactive (T 1/2 = 1.5 days). Experiencing grip
82. LEAD ROSFOND includes data for all 4 stable and 4 long-lived radioactive lead isotopes. 82.1. Lead-202 is radioactive. (T 1/2 = 5.25 * 10 4 years). By capturing an orbital electron
48. CADMIUM 48.0. General remarks For the ROSFOND library, it was required to select neutron data for 8 stable and 4 long-lived cadmium isotopes. Consider the results of data reassessment activities
1 3. FILE 3. CROSS-SECTIONS OF REACTIONS 3.1. GENERAL DESCRIPTION File 3 gives the cross-sections and derivatives of quantities in the form of a function of the energy E, where E is the energy of the incident particle (in eV) in the laboratory system. They represent
68. ERBIUM Natural erbium contains six isotopes. Table 1 lists the contribution of each isotope to the natural mixture. Table 1 Composition of natural erbium,% Isotope% Er-162 0.139 Er-164 1.601 Er-166 33.503
70. ITTERBIUM Natural ytterbium has 7 stable isotopes: 168 Yb, 170 Yb, 171 Yb, 172 Yb, 173 Yb, 174 Yb, 176 Yb and three rather long-lived radioactive isotopes: 166 Yb, 169 Yb, 175 Yb. Neither of
5. BOR 5.1. Bor-10 Content in natural mixture: 19.8 ± 0.3%. Ground state spin: 3+. 1. Reaction files 10 B (n, α) (MT = 107) and 10 B (n, αγ 1) (MT = 801) are used as standards when measuring
27. COBALT FOUNDATION-2.2 contains an estimate by T.Aoki, T.Asami, 1982. For radionuclides, the EAF-3 estimate is accepted. In -VII the estimate of A. Smith, G. DeSaussure, 1989 is accepted. In -3.3 is the estimate of T. Watanabe, 1994. In JEFF-3.1
88. RADIUS 88.0. General remarks Element 88 was discovered by the Curies in 1898 in a mineral known as uranium tar, blende resin, and pitchblende. Already during this very first work, it became clear
62.SAMARIUM There are 11 stable and long-lived isotopes of samarium, of which 7 have survived in nature. Two radioactive isotopes (151 Sm and 153 Sm) are formed as a result of the fission of heavy nuclei. As
23. VANADIUM Natural vanadium contains two isotopes V-5 (a weakly radioactive isotope with a content of 25%) and V-51. Thus, natural vanadium consists almost entirely of one isotope. Two more radioisotopes
69.THULLIUM Tullium has only one stable isotope - 169 Tm and 6 radioactive ones with a half-life of more than a day: 3 neutron-deficient (165 Tm, 167 Tm, 168 Tm) and three neutron-rich (170 Tm,
72. HAFNIUS 72.0. General remarks Hafnium has 6 stable isotopes: 174 Hf, 176 Hf, 177 Hf, 178 Hf, 179 Hf, 180 Hf. Two of them have long-lived isomers (and the latter). These are 178 Hf n (T1 / 2 = 31g.) And 179
93. NEPTUNIUM There are three natural radioactive families of thorium-232, uranium-235 and uranium-238, and one artificial radioactive series, the family of neptunium-237. In addition to "artificiality", this family is distinguished by
1 4. FILE 4. ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS 4.1. GENERAL DESCRIPTION File 4 contains representations of the angular distributions of secondary neutrons. It is only used for neutron reactions, reaction
DISPROSIUM.0 General remarks For the ROSFOND library, it was required to select neutron data for 10 stable and long-lived isotopes of dysprosium. It also seemed appropriate to include data for
3. The Hauser-Feshbach theory. Following Hauser and Feshbach, we express the cross sections of the compound processes in terms of the average values of the widths. We will proceed from the Breit-Wigner formalism. For an element of the S-matrix in the presence of a direct
95. AMERITIUS 95.0. General remarks The classical scheme for obtaining americium looks like this: 239 94 Pu + 1 0n (γ) 240 94Pu + 1 0n (γ, β) 241 95Am. Americium is a silvery white metal, ductile and malleable.
6. CARBON General remarks. Natural carbon contains two stable isotopes in the following atomic concentrations: 12 C 98.89%; 13 C 1.11%. There is also a very long-lived (T 1/2 = 5730 y) isotope 14C,
2. HELIUM 4 Not. The ROSFOND library contains data for two isotopes of helium, 3 He and 2.1. Helium-3 1. General remarks Modern libraries contain three independent estimates of neutron data for helium-3,
54. XENON 54.0 General remarks 14 stable and long-lived isotopes and isomers of samarium are known, of which 9 have survived in nature. Of the remaining five, four are long-lived isomers. Very
64. GADOLINIUM 64.0 General remarks For the ROSFOND library it was required to select neutron data for 12 stable and long-lived isotopes of gadolinium. Data for all of these isotopes are contained in the library
77. IRIDIUM 77.0 General Notes This section describes: two stable and seven radioactive isotopes of iridium with a half-life of more than a day. 77.1. Iridium-188. Radioactive. Experiencing the capture of the orbital
7.NITROGEN The data for two stable nitrogen isotopes are entered into ROSFOND: N-14 (99.634%) and N-15 (0.366%). Nitrogen has no long-lived radioactive isotopes. In the process of analyzing neutron data, we used
1 12. FILE 12. MULTIPLIQUES OF FORMATION OF PHOTONS AND PROBABILITIES OF TRANSITIONS File 12 can be used to represent the energy dependences of the cross sections for the production of photons either in terms of multiplicities,
Neutron nuclear reactions Neutron nuclear reactions A nuclear reaction is the process and result of the interaction of nuclei with various nuclear particles (alpha, beta particles, protons, neutrons, gamma quanta
36 CRYPTON 36.1. Krypton-78 Content in the natural mixture 0.35%. 1982 assessment by a team of experts for ENDF / B-V. fission products. assessment for the international library of product data
73. TANTALUM ROSFOND should provide neutron data for 2 natural and 4 long-lived radioactive isotopes of tantalum. Of the two natural isotopes of tantalum, only 181 Ta is stable.
89 ACTINIUM 89.0. General remarks There is only one reason why element 89 of anemones is of interest to many today. This element, like lanthanum, turned out to be the ancestor of a large family of elements, in
13. ALUMINUM Natural aluminum contains one isotope, 27 Al. There is also a long-lived isotope 26 Al, for which data should also be presented in the ROSFOND library. 13.1. Aluminum-26 Radioactive.
The element, named after one of the main Scandinavian gods, can save humanity from the energy crisis that awaits us in the near future.
In 1815, the famous Swedish chemist Jens Jakob Berzelius announced the discovery of a new element, which he named thorium in honor of Thor, the god of thunder and the son of the supreme Scandinavian god Odin. However, in 1825 it was discovered that this discovery was a mistake. Nevertheless, the name came in handy - Berzelius gave it to a new element, which he discovered in 1828 in one of the Norwegian minerals (now this mineral is called thorite). This element may have a great future, where it will be able to play a role in nuclear power that is not inferior in importance to the main nuclear fuel - uranium.
Distant relatives of the bomb
Nuclear power, on which so many hopes are now pinned, is a side branch of military programs, the main goals of which were the creation of nuclear weapons (and, a little later, reactors for submarines). As a nuclear material for making bombs, one could choose from three possible options: uranium-235, plutonium-239, or uranium-233.
Uranium-235 is found in natural uranium in a very small amount- only 0.7% (the remaining 99.3% is isotope 238), and it needs to be isolated, and this is an expensive and complicated process. Plutonium-239 does not exist in nature, it must be produced by irradiating uranium-238 with neutrons in a reactor, and then separating it from the irradiated uranium. In the same way, uranium-233 can be obtained by irradiating thorium-232 with neutrons.
The first two methods were implemented in the 1940s, but physicists decided not to bother with the third. The fact is that in the process of irradiation of thorium-232, in addition to useful uranium-233, a harmful impurity is also formed - uranium-232 with a half-life of 74 years, the decay chain of which leads to the appearance of thallium-208. This isotope emits high-energy (hard) gamma rays, which require thick lead plates to protect against. In addition, hard gamma radiation disables the electronic control circuits, which are indispensable in the design of the weapon.
Thorium cycle
Nevertheless, thorium was not entirely forgotten. Back in the 1940s, Enrico Fermi proposed producing plutonium in fast reactors (this is more efficient than thermal reactors), which led to the creation of the EBR-1 and EBR-2 reactors. In these reactors, uranium-235 or plutonium-239 is the source of neutrons that convert uranium-238 to plutonium-239. In this case, more plutonium can be formed than is “burned” (by 1.3–1.4 times), therefore such reactors are called “breeders”.
Ideal ecosystem
In the 1960s, it was planned to close the nuclear cycle for uranium and plutonium, using about 50% of nuclear power plants in thermal reactors and 50% in fast reactors. But the development of fast reactors has caused difficulties, so that currently only one such reactor is in operation - BN-600 at the Beloyarsk NPP (and another one, BN-800, has been built). Therefore, a balanced system can be created from thorium thermal reactors and about 10% of fast reactors, which will fill the missing fuel for thermal reactors.Another scientific group, led by Eugene Wigner, proposed its own design for a breeder reactor, but not on fast, but on thermal neutrons, with thorium-232 as the irradiated material. At the same time, the reproduction rate decreased, but the design was safer. However, there was one problem. The thorium fuel cycle looks like this. By absorbing a neutron, thorium-232 turns into thorium-233, which quickly turns into protactinium-233, and it already spontaneously decays into uranium-233 with a half-life of 27 days. And during this month protactinium will absorb neutrons, interfering with the production process. To solve this problem, it would be nice to remove protactinium from the reactor, but how can this be done? After all, constant loading and unloading of fuel reduces the operating efficiency to almost zero. Wigner proposed a very ingenious solution - a reactor with liquid fuel in the form of an aqueous solution of uranium salts. In 1952, at Oak Ridge National Laboratory, under the direction of Wigner's student, Alvin Weinberg, a prototype of such a reactor was built - Homogeneous Reactor Experiment(HRE-1). And soon an even more interesting concept appeared, ideally suited for working with thorium: it is a molten salt reactor, Molten-Salt Reactor Experiment... Fuel in the form of uranium fluoride was dissolved in a melt of lithium, beryllium and zirconium fluorides. MSRE operated from 1965 to 1969, and although thorium was not used there, the concept itself turned out to be quite workable: the use of liquid fuel increases the production efficiency and allows the removal of harmful fission products from the core.
Path of Least Resistance
Nevertheless, molten salt reactors (ZhSR) did not become widespread, since conventional thermal reactors using uranium turned out to be cheaper. The world nuclear power industry took the simplest and cheapest path, taking as a basis the proven pressurized water-cooled reactors (VVER), the descendants of those that were designed for submarines, as well as boiling water-cooled reactors. Graphite-moderated reactors such as the RBMK are another branch of the family tree - they descend from reactors used to produce plutonium. “The main fuel for these reactors is uranium-235, but its reserves, although quite significant, are nevertheless limited,” Stanislav Subbotin, head of the strategic systems research department at the Kurchatov Institute Research Center, explains to Popular Mechanics. - This issue began to be considered back in the 1960s, and then the planned solution to this problem was considered to be the introduction of waste uranium-238 into the nuclear fuel cycle, the reserves of which are almost 200 times larger. For this, it was planned to build many fast neutron reactors, which would produce plutonium with a breeding ratio of 1.3–1.4, so that the excess could be used to power thermal reactors. The BN-600 fast reactor was launched at the Beloyarsk NPP, although not in breeder mode. Another BN-800 was recently built there. But to build an effective ecosystem of nuclear power such reactors need about 50%. "
Mighty thorium
This is where thorium comes in. “Thorium is often called an alternative to uranium-235, but this is completely wrong,” says Stanislav Subbotin. - Thorium itself, like uranium-238, is not nuclear fuel at all. However, by placing it in a neutron field in the most ordinary pressurized water reactor, you can get an excellent fuel - uranium-233, which can then be used for the same reactor. That is, no alterations, no major changes to the existing infrastructure are needed. Another plus of thorium is its abundance in nature: its reserves are at least three times higher than those of uranium. In addition, there is no need for isotope separation, since only thorium-232 is found along with rare-earth elements during associated mining. Again, when uranium is mined, the surrounding area is contaminated with relatively long-lived (half-life 3.8 days) radon-222 (in the series of thorium, radon-220 is short-lived, 55 seconds, and does not have time to spread). In addition, thorium has excellent thermomechanical properties: it is refractory, less prone to cracking, and releases less radioactive gases when the fuel element cladding is damaged. The production of uranium-233 from thorium in thermal reactors is about three times more efficient than plutonium from uranium-235, so that the presence of at least half of such reactors in the nuclear power ecosystem will close the cycle for uranium and plutonium. True, fast reactors will still be needed, since the breeding ratio of thorium reactors does not exceed unity. "
However, thorium also has one rather serious drawback. Under neutron irradiation of thorium, uranium-233 becomes contaminated with uranium-232, which undergoes a decay chain leading to the hard gamma-emitting isotope thallium-208. “This greatly complicates the work on fuel reprocessing,” explains Stanislav Subbotin. “But on the other hand, it makes it easier to detect such material, reducing the risk of theft. In addition, in a closed nuclear cycle and in automated fuel handling, this does not really matter. "
Thermonuclear ignition
Experiments on the use of thorium fuel rods in thermal reactors are being carried out in Russia and other countries - Norway, China, India, and the USA. “Now is the time to return to the idea of molten salt reactors,” says Stanislav Subbotin. - The chemistry of fluorides and fluoride melts is well studied thanks to the production of aluminum. For thorium, molten salt reactors are much more efficient than conventional pressurized water reactors, since they allow flexible loading and removal of fission products from the reactor core. Moreover, they can be used to implement hybrid approaches, using thermonuclear installations, at least the same tokamaks, not nuclear fuel as a neutron source. In addition, the molten salt reactor allows solving the problem with minor actinides - long-lived isotopes of americium, curium and neptunium (which are formed in irradiated fuel), “afterburning” them in a scavenger reactor. So, in the future, we will not be able to do without thorium in the nuclear power industry for several decades ”.
Plan:
- Introduction
- 1 Formation and decay
- 2 Receiving
- 3 Application Notes (edit)
Introduction
Uranium-232(eng. uranium-232) is a radioactive nuclide of the chemical element uranium with atomic number 92 and mass number 232. Due to its long decay chain and higher specific energy release than most other isotopes, uranium-232 is a promising nuclide for use in radioisotope energy sources.
The activity of one gram of this nuclide is approximately 827.38 GBq.
1. Formation and decay
Uranium-232 is formed as a result of the following decays:
- β + -decay of 232 Np nuclide (half-life is 14.7 (3) min):
- β - -decay of 232 Pa nuclide (half-life is 1.31 (2) days):
- α-decay of nuclide 236 Pu (half-life is 2.858 (8) years):
The decay of uranium-232 occurs in the following directions:
- α-decay in 228 Th (100% probability, decay energy 5 413.63 (9) keV):
the energies of the emitted alpha particles are 5,263.36 keV (in 31.55% of cases) and 5,320.12 keV (in 68.15% of cases).
- Spontaneous division (probability less than 1 × 10 −12%);
- Cluster decay with the formation of the 28 Mg nuclide (the decay probability is less than 5 × 10 −12%):
- Cluster decay with the formation of the 24 Ne nuclide (decay probability 8.9 (7) × 10 −10%):
2. Receiving
Uranium-232 is formed as a by-product in the production of uranium-233 by neutron bombardment of thorium-232. Along with the reaction of formation of uranium-233, the following side reactions occur in the irradiated thorium fuel:
Since the effective cross section for (n, 2n) reactions for thermal neutrons is small, the yield of uranium-232 depends on the presence of a significant amount of fast neutrons (with an energy of at least 6 MeV).
If thorium-230 nuclide is present in thorium fuel in significant quantities, then the formation of uranium-232 is supplemented by the following reaction proceeding with thermal neutrons:
Since the presence of uranium-232 in the irradiated fuel makes it difficult to work safely with it (see the Application section), to reduce the formation of uranium-232 it is necessary to use thorium fuel with a minimum concentration of thorium-230.
3. Application
Uranium-232 is the ancestor of a long decay chain, which includes nuclides emitting hard gamma quanta:
232 U (α; 68.9 years) 228 Th (α; 1.9 years) 224 Ra (α; 3.6 days; emits a γ-quantum 0.24 MeV in 4.10% of decays) 220 Rn (α ; 56 s; γ 0.55 MeV, 0.114%) 216 Po (α; 0.15 s) 212 Pb (β−; 10.64 h) 212 Bi (α; 61 s; γ 0.73 MeV, 6, 67%; γ 1.62 MeV, 1.47%) 208 Tl (β−; 3 min; γ 2.6 MeV, 99.16%; γ 0.58 MeV, 84.5%) 208 Pb (stable)
The rapid sequence of decays starting with radium-224 is accompanied by a significant amount of gamma radiation, with about 85% of all gamma-ray energy generated by the decay of thallium-208, which emits predominantly 2.6 MeV gamma quanta. This feature leads to the fact that the presence of uranium-232 as an impurity to uranium-233 is extremely undesirable, making it difficult to work safely with it.
On the other hand, the high specific energy release makes this nuclide extremely promising for use in radioisotope energy sources.
Notes (edit)
- 1 2 3 4 5 G. Audi, A.H. Wapstra, and C. Thibault (2003). “The AME2003 atomic mass evaluation (II). Tables, graphs, and references. - www.nndc.bnl.gov/amdc/masstables/Ame2003/Ame2003b.pdf ". Nuclear Physics A 729 : 337-676. DOI: 10.1016 / j.nuclphysa.2003.11.003 - dx.doi.org/10.1016/j.nuclphysa.2003.11.003.
- 1 2 3 4 5 6 7 8 9 G. Audi, O. Bersillon, J. Blachot and A. H. Wapstra (2003). "The NUBASE evaluation of nuclear and decay properties - www.nndc.bnl.gov/amdc/nubase/Nubase2003.pdf". Nuclear Physics A 729 : 3-128. DOI: 10.1016 / j.nuclphysa.2003.11.001 - dx.doi.org/10.1016/j.nuclphysa.2003.11.001.
- 232 U properties on the IAEA (International Atomic Energy Agency) website - www-nds.iaea.org/relnsd/tablenucsENSDF.jsp?query=3447
- 1 2 Carey sublette Nuclear Weapons Frequently Asked Questions - nuclearweaponarchive.org/Nwfaq/Nfaq6.html. nuclearweaponarchive.org.
- Nuclide table on the IAEA website - www-nds.iaea.org/relnsd/vchart/index.html
In 1815, the famous Swedish chemist Jens Jakob Berzelius announced the discovery of a new element, which he named thorium in honor of Thor, the god of thunder and the son of the supreme Scandinavian god Odin. However, in 1825 it was discovered that this discovery was a mistake. Nevertheless, the name came in handy - Berzelius gave it to a new element, which he discovered in 1828 in one of the Norwegian minerals (now this mineral is called thorite). This element may have a great future, where it will be able to play a role in nuclear power that is not inferior in importance to the main nuclear fuel - uranium.
| Advantages and disadvantages | |
| + | Thorium on Earth is several times more than uranium |
| + | No need to separate isotopes |
| + | Radioactive contamination from thorium mining is significantly less (due to shorter-lived radon) |
| + | Existing thermal reactors can be used |
| + | Thorium has better thermomechanical properties than uranium |
| + | Thorium is less toxic than uranium |
| + | When using thorium, minor actinides (long-lived radioactive isotopes) are not formed |
| - | Irradiation of thorium produces gamma-emitting isotopes, which creates difficulties in fuel reprocessing |
Distant relatives of the bomb
Nuclear power, on which so many hopes are now pinned, is a side branch of military programs, the main goals of which were the creation of nuclear weapons (and, a little later, reactors for submarines). As a nuclear material for making bombs, one could choose from three possible options: uranium-235, plutonium-239, or uranium-233.
This is how the thorium nuclear cycle looks like, illustrating the transformation of thorium into a highly efficient nuclear fuel - uranium-233.
Uranium-235 is contained in natural uranium in a very small amount - only 0.7% (the remaining 99.3% is isotope 238), and it must be isolated, and this is an expensive and complex process. Plutonium-239 does not exist in nature, it must be produced by irradiating uranium-238 with neutrons in a reactor, and then separating it from the irradiated uranium. In the same way, uranium-233 can be obtained by irradiating thorium-232 with neutrons.
In the 1960s, it was planned to close the nuclear cycle for uranium and plutonium, using about 50% of nuclear power plants in thermal reactors and 50% in fast reactors. But the development of fast reactors has caused difficulties, so that currently only one such reactor is in operation - BN-600 at the Beloyarsk NPP (and another one, BN-800, has been built). Therefore, a balanced system can be created from thorium thermal reactors and about 10% of fast reactors, which will fill the missing fuel for thermal reactors.
The first two methods were implemented in the 1940s, but physicists decided not to bother with the third. The fact is that in the process of thorium-232 irradiation, in addition to useful uranium-233, a harmful impurity is also formed - uranium-232 with a half-life of 74 years, the decay chain of which leads to the appearance of thallium-208. This isotope emits high-energy (hard) gamma rays, which require thick lead plates to protect against. In addition, hard gamma radiation disables the electronic control circuits, which are indispensable in the design of the weapon.
Thorium cycle
Nevertheless, thorium was not entirely forgotten. Back in the 1940s, Enrico Fermi proposed producing plutonium in fast reactors (this is more efficient than thermal reactors), which led to the creation of the EBR-1 and EBR-2 reactors. In these reactors, uranium-235 or plutonium-239 is the source of neutrons that convert uranium-238 to plutonium-239. In this case, more plutonium can be formed than is “burnt” (1.3-1.4 times), therefore such reactors are called “breeders”.
Another scientific group, led by Eugene Wigner, proposed its own design for a breeder reactor, but not on fast, but on thermal neutrons, with thorium-232 as the irradiated material. At the same time, the reproduction rate decreased, but the design was safer. However, there was one problem. The thorium fuel cycle looks like this. By absorbing a neutron, thorium-232 turns into thorium-233, which quickly turns into protactinium-233, and it already spontaneously decays into uranium-233 with a half-life of 27 days. And during this month protactinium will absorb neutrons, interfering with the production process. To solve this problem, it would be nice to remove protactinium from the reactor, but how can this be done? After all, constant loading and unloading of fuel reduces the operating efficiency to almost zero. Wigner proposed a very ingenious solution - a reactor with liquid fuel in the form of an aqueous solution of uranium salts. In 1952, a prototype of such a reactor, the Homogeneous Reactor Experiment (HRE-1), was built at Oak Ridge National Laboratory under the direction of Wigner's student, Alvin Weinberg. And soon an even more interesting concept emerged, ideally suited to working with thorium: the Molten-Salt Reactor Experiment. Fuel in the form of uranium fluoride was dissolved in a melt of lithium, beryllium and zirconium fluorides. MSRE operated from 1965 to 1969, and although thorium was not used there, the concept itself turned out to be quite workable: the use of liquid fuel increases the production efficiency and allows the removal of harmful fission products from the core.
A liquid salt reactor allows for much more flexible control of the fuel cycle than conventional thermal power plants, and uses fuel with the highest efficiency, removing harmful fission products from the core and adding new fuel as needed.
Path of Least Resistance
Nevertheless, molten salt reactors (ZhSR) did not become widespread, since conventional thermal reactors using uranium turned out to be cheaper. The world nuclear power industry took the simplest and cheapest path, taking as a basis the proven pressurized water-cooled reactors (VVER), the descendants of those that were designed for submarines, as well as boiling water-cooled reactors. Graphite-moderated reactors such as the RBMK are another branch of the family tree - they descend from reactors used to produce plutonium. “The main fuel for these reactors is uranium-235, but its reserves, although quite significant, are nevertheless limited,” explains Stanislav Subbotin, head of the strategic systems research department at the Kurchatov Institute Research Center, to Popular Mechanics. - This issue began to be considered back in the 1960s, and then the planned solution to this problem was considered to be the introduction of waste uranium-238 into the nuclear fuel cycle, the reserves of which are almost 200 times larger. For this, it was planned to build many fast neutron reactors, which would produce plutonium with a breeding ratio of 1.3-1.4, so that the excess could be used to power thermal reactors. The BN-600 fast reactor was launched at the Beloyarsk NPP, although not in breeder mode. Another BN-800 was recently built there. But to build an effective ecosystem of nuclear power such reactors need about 50%. "
All radioactive isotopes that occur naturally in natural conditions belong to one of three families (radioactive series). Each such row is a chain of nuclei linked by sequential radioactive decay. The ancestors of the radioactive series are the long-lived isotopes uranium-238 (half-life 4.47 billion years), uranium-235 (704 million years), and thorium-232 (14.1 billion years). The chains end in stable lead isotopes. There is another series, starting with neptunium-237, but its half-life is too short - only 2.14 million years, so it does not occur in nature.
Mighty thorium
This is where thorium comes in. “Thorium is often called an alternative to uranium-235, but this is completely wrong,” says Stanislav Subbotin. - Thorium itself, like uranium-238, is not nuclear fuel at all. However, by placing it in the neutron field in the most ordinary pressurized water reactor, you can get an excellent fuel - uranium-233, which can then be used for the same reactor. That is, no alterations, no major changes to the existing infrastructure are needed. Another plus of thorium is its abundance in nature: its reserves are at least three times higher than those of uranium. In addition, there is no need for isotope separation, since only thorium-232 is found along with rare-earth elements during associated mining. Again, when uranium is mined, the surrounding area is contaminated with relatively long-lived (half-life 3.8 days) radon-222 (in the series of thorium, radon-220 is short-lived, 55 seconds, and does not have time to spread). In addition, thorium has excellent thermomechanical properties: it is refractory, less prone to cracking, and releases less radioactive gases when the fuel element cladding is damaged. The production of uranium-233 from thorium in thermal reactors is about three times more efficient than plutonium from uranium-235, so that the presence of at least half of such reactors in the nuclear power ecosystem will close the cycle for uranium and plutonium. True, fast reactors will still be needed, since the breeding ratio of thorium reactors does not exceed unity. "
The production of 1 GW during the year requires: 250 tons of natural uranium (contain 1.75 tons of uranium-235), it is required to extract 215 tons of depleted uranium (including 0.6 tons of uranium-235) go to dumps; 35 tons of enriched uranium (of which 1.15 tons of uranium-235) are loaded into the reactor; spent fuel contains 33.4 tons of uranium-238, 0.3 tons of uranium-235, 0.3 tons of plutonium-239, 1 ton of fission products. 1 ton of thorium-232, when loaded into a molten salt reactor, is completely converted into 1 ton of uranium-233; 1 ton of decay products, of which 83% are short-lived isotopes (decay to stable ones in about ten years).
However, thorium also has one rather serious drawback. Under neutron irradiation of thorium, uranium-233 becomes contaminated with uranium-232, which undergoes a decay chain leading to the hard gamma-emitting isotope thallium-208. “This greatly complicates the work on fuel reprocessing,” explains Stanislav Subbotin. “But on the other hand, it makes it easier to detect such material, reducing the risk of theft. In addition, in a closed nuclear cycle and in automated fuel handling, this does not really matter. "
Thermonuclear ignition
Experiments on the use of thorium fuel rods in thermal reactors are being carried out in Russia and other countries - Norway, China, India, and the USA. “Now is the time to return to the idea of molten salt reactors,” says Stanislav Subbotin. - The chemistry of fluorides and fluoride melts is well studied thanks to the production of aluminum. For thorium, molten salt reactors are much more efficient than conventional pressurized water reactors, since they allow flexible loading and removal of fission products from the reactor core. Moreover, they can be used to implement hybrid approaches, using thermonuclear installations, at least the same tokamaks, not nuclear fuel as a neutron source. In addition, a molten salt reactor can solve the problem with minor actinides - long-lived isotopes of americium, curium and neptunium (which are formed in irradiated fuel), "afterburning" them in a scavenger reactor. So, in the future, we cannot do without thorium in the nuclear power industry for several decades ”.
