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Isotope-production cross sections for p-, d-, and C-induced spallation reactions on Nb93 at 113 MeV/nucleon were measured using the inverse-kinematics method employing secondary targets of CH2, CD2, and C. The measured cross sections for Mo90, Nb90, Y86,88 produced by p-induced reactions were found to be consistent with those measured by the conventional activation method. We performed benchmark tests of the reaction models INCL-4.6, JQMD, and JQMD-2.0 implemented in the Particle and Heavy Ion Transport code System (PHITS) and of the nuclear data libraries JENDL-4.0/HE, TENDL-2017, and ENDF/B-VIII.0. The model calculations also showed generally good agreement with the measured isotope-production cross sections for p-, d-, and C-induced reactions. It also turns out that, among the three nuclear data libraries, JENDL-4.0/HE provides the best agreement with the measured data for the p-induced reactions. We compared the present Nb93 data with the Zr93 data, that were measured previously by the same inverse kinematics method (Kawase et al., Prog. Theor. Exp. Phys. 2017, 093D03 (2017)2050-391110.1093/ptep/ptx110), with particular attention to the effect of neutron-shell closure on isotope production in p- and d-induced spallation reactions. The isotopic distributions of the measured production cross sections in the Zr93 data showed noticeable jumps at neutron number N=50 in the isotopic chains of ΔZ=0 and -1, whereas no such jump appeared in isotopic chain of ΔZ=0 in the Nb93 data. From INCL-4.6 + GEM calculations, we found that the jump formed in the evaporation process is smeared out by the intranuclear cascade component in Nb91 produced by the Nb93(p,p2n) and (d,d2n) reactions on Nb93. Moreover, for Nb93, the distribution of the element-production cross sections as a function of the change in proton number ΔZ is shifted to smaller ΔZ than for Zr93, because the excited Nb prefragments generated by the cascade process are more likely to emit protons than the excited Zr prefragments, due to the smaller proton-separation energies of the Nb isotopes.

The Sodium-cooled Fast Reactor (SFR) is one of the most promising concepts suggested for Generation-IV nuclear reactor systems. Some SFRs adopt Steam Generator (SG) as their heat exchange system between sodium and water. Sodium-water reaction occurs in the tube failure accident of a SG. The tube failure may propagate to adjacent tubes resulting in a large scale tube failure by this reaction. In an advanced loop-type SFR design promoted by Japan Atomic Energy Agency (JAEA), a straight double-walled tube SG is adopted to prevent this sodium-water reaction [1], [2]. The double-walled tube is expected to prevent water leakage by acting as double wall boundary and mitigate consequences of the sodium-water reaction. It is expected for the outer tubes to practically behave as waste resistant for the adjacent tubes to mitigate sodium water reaction consequences. Mitigation is expected in Design Extension Conditions (DECs) such as the loss of the mitigation function which might lead an initial water leakage to large scale tube failure. In addition to the prevention of the initial leakage, the initial water leakage rate is practically suppressed because of the narrow gap between the inner and the outer tube. In this paper, tube failure propagation has been calculated to assess property protection performance on outer tubes. The evaluation results showed that the total leakage rate is limited to one double-ended guillotine scale hence the double-walled tube SG has the property protection performance. By additional calculations assuming the loss of the mitigation function, a sever event in DECs is cleared. These calculations suggest that increase of the reliability of water blowdown system and enhancement of the pressure release system are effective for the boundary integrity between primary and secondary cooling systems. There is an issue to be addressed to adopt the concept described above, that is, the decrease of temperature difference between exchange tubes especially for structural integrity of the straight double-walled tube SG for its thermal contact resistance between double-walled tubes and its lack of bending part to release thermal stress. The dispersion of thermal contact resistance between tubes causes temperature difference there due to their heat transfer rate difference. To suppress this dispersion, the oxidized scale is reduced on the interface between the inner and the outer tubes by applying heat treatment using hydrogen furnace for the tube element production. Then, thermal contact resistance of the double-walled tube is successfully reduced at laboratory scale. Thus, these results suggest that the double-walled tube SG may suppress water leakage rate and sodium-water reaction consequences in DECs. Furthermore, temperature difference between exchange tubes due to oxidized scale on the interface between the inner and the outer tubes can be reduced at laboratory scale. Hereafter, the specifications of the double-walled tube SG will be determined including tolerance reinforcement of sodium boundary.