Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

Cross Section Measurement for 199Hg(n,n′)199m Hg Reaction from 0.78 to 6.3 MeV

Cross sections for the ¹⁹⁹Hg(n, n′)^199m Hg reaction have been measured at 10 energy points from 0.78 to 6.3 MeV by the activation method. Monoenergetic neutrons below 2 MeV were produced by the ⁷Li (p, n)⁷Be reaction and those above 2 MeV were produced by the D(d, n)³He reaction using a 5.5 MV Van de Graaff accelerator. The neutron flux was determined with a proton recoil telescope counter and In-foils. The measured cross sections were expressed by an empirical formula. The fission spectrum averaged cross section calculated with this formula is 238.3 mb for the Watt-type fission spectrum, and is about 14% smaller than that recently measured by three of the authors.     

Measurement of 76Se and 78Se (γ, n) Cross Sections

The (γ, n) cross sections of Se isotopes (⁷⁶Se,⁷⁸Se) were measured to supply fundamental data for estimating the inverse reaction cross section, i.e., the ⁷⁹Se(n, γ)⁸⁰Se cross section. The enriched samples and a reference ¹⁹⁷Au sample were irradiated with laser-Compton scattering (LCS) γ-rays. The excitation function of each (γ, n) cross section was determined for the energy range from each near neutron separation energy to the threshold energy of (γ, 2n) reaction. The energy point corresponding to each cross section was deduced using the accurately determined energy distribution of LCS γ-rays. Systematic (γ, n) cross sections for Se isotopes including ⁸⁰Se were compared with those calculated by using a statistical model calculation code TALYS.     

55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

This study evaluated methods for increasing the helium production rate in ferritic steel irradiation in a fission reactor neutron spectrum in order to increase the helium to atomic displacement ratio to values typical of fusion reactor first wall conditions. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched ⁵⁴Fe in iron in the High Flux Isotope Reactor (HFIR), ORNL. In the current work, the ferritic steel He(appm)/dpa ratio was studied in the neutron spectrum of HFIR with the ⁵⁵Fe thermal neutron helium production taken into account. A benchmark calculation for the same sample, as used in the aforementioned experiment, was then used to adjust and evaluate the ⁵⁵Fe (n, a) cross section values in TALYS-based Evaluated Nuclear Data Library (TENDL). The analysis showed that a decrease of a factor of 6700 for the TENDL ⁵⁵Fe (n, a) cross section in the intermediate and low energy regions was required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy was about 27 mb. With the adjusted ⁵⁵Fe (n, a) cross sections, calculation showed that the ⁵⁴Fe and ⁵⁵Fe isotopes could be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the ⁵⁵Fe (n, a) cross section at thermal energy.     

Measurement of Average Cross Section for 233U(n, 2n)232U Reaction

Isotopically pure ²³³U samples, with only 3 × l0^?3 ppm²³²U content, were prepared by thermal neutron irradiation of thoria and subsequent chemical processing. The ²³³U sample thus obtained was reirradiated with a fission neutron spectrum in the core of the Kyoto University Reactor (KUR), and measurements were made of the fission spectrum average cross section for the ²³³U(n, 2n) ²³²U reaction. A value of 4.08±0.30 mb was obtained for this cross section, in agreement with the renormalized value of Halperin et al. within the limits of experimental error.

In order to assess the energy dependent cross section from the value of this integral measurement, the ²³³U (n, 2n) cross section was calculated assuming a Maxwellian-type fission spectrum and adopting the energy dependent evaluated cross sections of ENDF/B-III and other authors. The values of the cross section thus determined were found to be about 32 to 91% larger than the measured cross section given above. The result of Pearlstein's calculation of the ²³³U(n, 2n) cross section by the statistical model, again assuming the Maxwellian distribution, is smaller than the measured cross section by about 19%.     

Measurement of the 93Nb(n,n')93mNb Reaction Cross Section at 14.5 and 14.9MeV

Abstract

Cross sections at 14.5 and 14.9 MeV for the low threshold energy reaction of ⁹³Nb(n, n')^93mNb were measured by the activation method at the FNS (Fusion Neutronics Source) of JAERI. Although there are multiple experimental data for this reaction cross section from the threshold to 9 MeV, only one data at 14.3 MeV has been reported by Ryves et at. The correction for low energy neutron contribution was found very sensitive on the reaction rate determination. Careful treatments were performed to arrive at final error evaluation considering neutron spectrum calculated with a Monte Carlo Code and cross section curves available. The cross sections measured in the present work are larger by more than 30% than those in both IRDF-90 and JENDL Dosimetry File, which are based on the data of Ryves et al. On the other hand, the present data are lower by 10-15% than the evaluation by Odano et al. It is highly recommended to re-evaluate the cross section by taking the present data into account.     

Measurements of activation cross sections of (n,p) and (n, α) reactions in the energy range of 3.5–5.9 MeV using a deuterium gas target

Activation cross sections of (n, p) and (n, α) reactions were measured by means of the activation method in the neutron energy range of 3.5–5.9 MeV using a deuterium gas target. The irradiated target isotopes were ²⁷Al, ^28,29Si, ⁴¹K, ⁵¹V, ⁶¹Ni, ⁶⁵Cu, ^64,67Zn, ⁶⁹Ga, ⁷⁹Br, ⁹²Mo and ⁹³Nb. The cross sections of the ²⁹Si(n, p) ²⁹Al, ⁶⁷Zn(n, p) ⁶⁷Cu, ⁶⁹Ga(n, p) ^69mZn, ⁷⁹Br(n, p) ^79mSe, and ⁶⁹Ga(n, α) ⁶⁶Cu reactions were obtained for the first time in the studied energy range. The d-D neutrons were generated by the deuterium gas target at the Van de Graaff accelerator (KN-VdG) at Nagoya University. All cross section values were determined relative to those of the ¹¹⁵In(n, n′)^115mIn reaction. The activities induced by the low-energy neutrons were corrected. For the corrections, the neutron spectra and mean neutron energies at the irradiation positions were calculated taking into account the energy loss of incident deuterons, the angular differential cross section of the d-D reaction and the solid angle subtended by the sample. The systematics of the (n, p) reactions at the neutron energy of 5.0 MeV in the mass range between 27 and 92 were proposed for the first time. This systematics can predict the cross sections within an accuracy of a factor of 1.6.     

Evaluation of Neutron Cross Sections of Plutonium-241

The neutron cross sections of ²⁴¹Pu were evaluated in the energy range between 10^?5 eV and 15MeV, and are stored in the Japanese Evaluated Nuclear Data Library Version-1 (JENDL-1). In the energy range below 100eV, the evaluated data contained in ENDE/B-IV and the resonance parameters recommended in BNL-325 were tentatively adopted. The unresolved resonance parameters were determined between 100 eV and 21.5 keV so as to reproduce the experimental data of the fission and capture cross sections. Above 21.5 keV, the fission cross section was evaluated on the basis of the experimental data, most of which were reported as the ratio to the fission cross section of ²³⁵U and then were normalized by the fission cross section of ²³⁵U adopted in JENDL-1. The capture cross section was obtained from the experimental data of a in the energy range up to 250 keV. The capture cross section above 250 keV and the elastic and inelastic scattering, (n, 2n) and (n, 3n) reaction cross sections above 21.5 keV were obtained on the basis of the theoretical calculations. The calculated cross sections are connected smoothly with those obtained from the unresolved resonance parameters at 21.5 keV. This suggests the self-consistency of the present evaluation.     

Measurements of activation cross sections for Lu(n, α)Tm,Lu(n, α)Tm and Lu(n, p)Yb reactions induced by neutrons around 14 MeV

The cross sections for the ¹⁷⁵Lu(n, α)¹⁷²Tm, ¹⁷⁶Lu(n, α)¹⁷³Tm and ¹⁷⁵Lu(n, p)^175m+gYb reactions have been measured in the neutron energy range of 13.5–14.8 MeV using the activation technique. The first data for ¹⁷⁵Lu(n, α)¹⁷²Tm reaction cross sections are presented. In our experiment, the fast neutrons were produced via the ³H(d, n)⁴He reaction on K-400 Neutron Generator at Chinese Academy of Engineering Physics (CAEP). Induced gamma activities were measured by a high-resolution (1.69 keV at 1332 keV for ⁶⁰Co) gamma-ray spectrometer with high-purity germanium (HPGe) detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. The neutron fluences were determined by the cross section of ⁹³Nb(n, 2n)^92mNb or ²⁷Al(n, α)²⁴Na reactions. The neutron energy in the measurement was by the cross section ratios of ⁹⁰Zr(n, 2n)^89m+gZr and ⁹³Nb(n, 2n)^92mNb reactions. The results were discussed and compared with experimental data found in the literature and with results of published empirical formulae.     

Development of the high-energy neutron fluence rate standard field in Japan with a peak energy of 45 MeV using the 7Li(p,n)7Be reaction at TIARA

In this study, we developed a 45 MeV neutron fluence rate standard of Japan. Quasi-monoenergetic neutrons with a peak energy of 45 MeV in the neutron standard field were produced by the ⁷Li(p,n)⁷Be reaction using a 50-MeV proton beam from an azimuthally varying field (AVF) cyclotron of the Takasaki Ion Accelerators for Advanced Radiation Application (TIARA). The neutron energy spectrum was measured using an organic liquid scintillation detector and a ⁶Li-glass scintillation detector by the time-of-flight method, and using a Bonner sphere spectrometer by the unfolding method. The absolute neutron fluence was determined using a proton recoil telescope (PRT) composed of the liquid scintillation detector and a Si(Li) detector that was newly developed in the present study. The detection efficiency of the PRT was obtained using the MCNPX code. The peak neutron production cross section for the ⁷Li(p,n)⁷Be reaction was also derived from the neutron fluence in order to confirm the neutron fluence of the TIARA high-energy neutron field. The peak neutron production cross section obtained in the present study was in good agreement with those of previous studies. The characteristics of the 45-MeV neutron field in TIARA were successfully evaluated in order to calibrate high-energy neutron detectors and high-energy neutron dosimeters.     

Isomeric ratio and cross section of the237Np(n, 2n) reaction

Conclusion It has thus been shown that on the basis of the Hauser-Feshbach theory, and taking into account the law of the conservation of moment, the isomer ratio for the reaction²³⁷Np (n, 2n) can be calculated. The indeterminateness of the modeling of the²³⁶Np level scheme, to all appearances, has little effect on the energy dependence of the isomeric ratio.The error in the calculated cross section for the (n, 2n) reaction is determined chiefly by the error in the experimental data on the isomeric ratio and on the cross section for the formation of the short-lived state. Obtaining a correct estimate of the error is made difficult by the scarcity of experimental data on the isomeric ratio.The results of this work can be useful in practical activity when combined with an estimate of the cross sections and the creation of a complete system of neutron cross sections for²³⁷Np. Theoretical estimates of the cross sections can to a significant extent compensate for the scarcity and indeterminateness of the experimental data.Translated from Atomnaya Énergiya, Vol. 63, No. 2, pp. 110–113, August, 1987.     

Measurements of Average Cross Section for 232Th(n, 2n)231Th Reaction to Neutrons with Fission-Type Reactor Spectrum and of Gamma-Ray Intensities of 231Th

The average cross section for the ²³²Th(n, 2n)²³¹Th reaction to neutrons with the energy spectrum close to that of fission neutrons was obtained in the core of the Kyoto University Reactor, KUR. The value obtained was 12.5±0.84 mb. This value agrees satisfactorily with Phillips' and with the calculated value obtained with the cross section in the U-K library and the Maxwellian fission neutron spectrum given by Leachman. A somewhat poorer agreement is seen with the calculated value obtained from Butler & Santry's cross section and Leachman's spectrum. The discrepancy amounts to 24 and to 39% respectively, for the average cross sections calculated with these two excitation functions and the fission neutron spectrum given by McElroy.

By making use of a Ge(Li) counter whose photopeak efficiency had been carefully calibrated, the absolute intensities were determined for eleven photopeaks observed on the γ-ray spectrum emitted by ²³¹Th.     

Measurements of neutron total and capture cross sections of 241Am with ANNRI at J-PARC

Neutron total and capture cross sections of ²⁴¹Am have been measured with a new data acquisition system and a new neutron transmission measurement system installed in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex. The neutron total cross sections of ²⁴¹Am were determined by using a neutron time-of-flight (TOF) method in the neutron energy region from 4 meV to 2 eV. The thermal total cross section of ²⁴¹Am was derived with an uncertainty of 2.9%. A pulse-height weighting technique was applied to determine neutron capture yields of ²⁴¹Am. The neutron capture cross sections were determined by the TOF method in the neutron energy region from the thermal to 100 eV, and the thermal capture cross section was obtained with an uncertainty of 4.1%. The evaluation data of JENDL-4.0 and JEFF-3.2 were compared with the present results.     

Maxwellian-averaged neutron-induced reaction cross sections and astrophysical reaction rates for kT = 1 keV to 1 MeV calculated from microscopic neutron cross section library JENDL-3.3

Maxwellian-averaged cross sections and astrophysical reaction rates were calculated for neutron capture, fission, (n,p), (n,α), and some threshold reactions based on a microscopic neutron cross section data library, the Japanese Evaluated Nuclear Data Library Version 3.3 (JENDL-3.3). The calculation was made for temperatures (kT) from 1 keV to 1 MeV including the most important range in astrophysical nucleosynthesis. Results are presented in the tables and graphs. The Maxwellian-averaged neutron capture cross sections are compared with recommendations of other authors. Implications of the differences between the present data and previous ones to s-process nucleosynthesis are discussed.     

The (p,n) reaction in lithium and the ground state of Be6

Using a time-of-flight method the neutron spectra in the Li⁶ + + p and Li⁷ + p reactions have been investigated at a proton energy of 9 Mev. Neutron groups have been found in the (p, n) reaction corresponding to the ground state in Be⁶ and the three lowest states of Be⁷ as well as a continuous neutron spectrum at lower energy, due to more complicated reactions. The observation of the neutron group for the Li⁶(p, n)Be⁶ reaction is the first experimental indication of the existence of the Be⁶ nucleus. The energy of the Li⁶(p, n)Be⁶ reaction is 5.2 Mev, the width of the ground state in Be⁶ is T < 0.3 Mev. The differential cross sections for neutron formation have been measured at 0, 15, 30, 60 and 120 °.     

Neutron Activation Cross Section of Molybdenum Isotopes at 14.8 MeV

The neutron activation cross sections of Mo isotopes have been measured for the 14.8 MeV neutron. The cross sections have been determined with reference to the known ²⁷A1 (n, α)²⁴Na and the ²⁷Al(n, p)²⁷Mg reactions. The cyclic activation method was employed for the γ-ray measurement of short-lived nuclei. A 55 cm³ Ge(Li) detector was used for the measurement of γ-ray spectra. Cross section data are presented for (n, 2n), (n, p) and (n, a) reactions on Mo isotopes. The cross sections of (n, np) reactions on ⁹⁸Mo are also presented. The exponential dependence on (N-Z)/A of the (n, p) reaction cross sections are discussed.     

Measurements of Average Cross Sections for Some Threshold Reactions for Neutrons with Fission-type Reactor Spectrum

The fast neutron spectrum in the core of the Kyoto University Reactor (KUR) was measured by using seven threshold reactions, a ⁶Li sandwich counter and also nuclear emulsion plates, and the results were compared with theoretical calculations by the S_n method. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons. Making use of these neutrons with fission-type spectrum, measurements were made of the average cross sections for twelve threshold reactions (⁴⁶Ti(n, p)⁴⁶Sc, ⁴⁷Ti(n, p)⁴⁷Sc, ⁴⁶Ti(n, p)⁴⁸Sc, ²⁸Si(n, p)²⁸Al, ²⁹Si(n, p)²⁹Al, ³⁰Si(n, α)²⁷Mg, ⁵¹V(n, α)⁴⁸Sc, ⁶⁴Zn(n, p)⁶⁴Cu, ⁹²Mo(n, p) ^92mNb(n, 2n)_92mNb, ₂₀₄Pb(n, n¹ )^204mPb and ²⁰⁴Pb(n, 2n)²⁰³Pb). For the purpose of comparison, the average cross sections for six among these reactions were measured also with neutrons from a fission plate. The results agreed within experimetal error with those obtained for neutrons in the KUR core. A Ge(Li) counter was mainly used for the measurement of γ-rays emitted from induced activities.     

Measurement of thermal neutron cross section and resonance integral for (n,γ) reaction in Sm

This study implies that ⁵⁵Mn(n,γ)⁵⁵Mn monitor reaction may be a convenient alternative comparator for the activation method and thus, it was used for the determination of thermal neutron cross section (TNX) and the resonance integral (RI) of the reaction ¹⁵²Sm(n,γ)¹⁵³Sm. The samples of MnO₂ and Sm₂O₃ diluted with Al₂O₃ powder were irradiated within and without a cylindrical 1 mm-Cd shield case in an isotropic neutron field obtained from the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated n-type Ge detector. The correction factors for γ-ray attenuation, thermal neutron and resonance neutron self-shielding effects and epithermal neutron spectrum shape factor (α) were taken into account in the determinations. The thermal neutron cross section for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has been determined to be 204.8 ± 7.9 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. For the TNX, most of the experimental data and evaluated one in JEFF-3.1, ENDF/B-VI, JENDL 3.3 and BROND 2.0, in general, agree well with the present result. The RI value for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has also been determined to be 3038 ± 214 b, relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction, using a 1/E^1+α epithermal neutron spectrum and assuming Cd cut-off energy of 0.55 eV. In surveying literature, the existing experimental and evaluated data for the RI values are distributed from 1715 to 3462 b. However, when the Cd cut-off energy is defined as 0.55 eV, the present RI value agrees with some previously reported RI values, 3020 ± 163 b by Simonits et al., 3141 ± 157 by Van Der Linden et al., and 2962 ± 54 b by Kafala et al., within the limits of error.     

Activation cross sections for (n, p) reactions on nickel isotopes induced by neutrons around 14 MeV

The cross sections of ⁵⁸Ni(n, p)^58(m+g)Co,⁶⁰Ni(n, p)^60mCo,⁶¹Ni(n, p)⁶¹Co and ⁶²Ni(n, p)^62mCo reactions induced by neutrons around 14 MeV were measured using activation technique and a coaxial HPGe γ-ray detector. The natural nickel foils of 99.9% purity were used as target materials. Fast neutrons were produced by the T(d, n)⁴He reaction. The neutron flux was determined using the monitor reaction ²⁷Al(n, α)²⁴Na and the neutron energies were measured with the method of cross-section ratios for ⁹⁰Zr(n, 2n)⁸⁹Zr to ⁹³Nb(n, 2n)^92mNb reactions. The results of this work are compared with the collected partial recent data published previously and the estimations obtained from the published empirical formula based on the statistical model with dependence on the Q-value and odd-even effect taken into consideration.     

Effective Cross Section of 238Samples for Analyzing Doppler Effect Measurements in Fast Critical Assembly

A new effective cross section of ²³⁸U sample has been derived to accurately analyze the Doppler effect measurements in a fast critical assembly. The neutron spectrum in the sample is determined by considering the interference effect of neutron spectrum between the sample and the surrounding fuel region, and is used to obtain the effective cross section of the sample. The new effective cross section can be calculated using the conventional self-shielding factors. The use of the present cross section has increased the Doppler reactivity worth in ZPPR-9 by about 4% compared to the result calculated by the conventional self-shielding factor method without the spectrum interference effect. The physical meaning of this increment is discussed.