Measurements of Thermal Neutron Cross Section and Resonance Integral for 237Np(n, γ)238Np Reaction

Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the ²³⁷Np(n, γ)²³⁸Np reaction was measured at a neutron energy of 0.0253 eV by the activation method. The result is 158±3 b, which is obtained relative to the reference value of 98.65±0.09 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction. Although the data in JENDL-3 is larger by about 15% than the present value, the recently revised data in JENDL-3.2 is close to the present. The ENDF/B-V, ENDF/B-VI, JEF-2 and Mughabghab's data are also larger by 7–15%. Old measurements are larger by 7–18% than the present data.

The resonance integral for the ²³⁷Np(n, γ)²³⁸Np reaction was also measured relative to the reference value of 1,550±28 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction with a 1/E standard neutron spectrum field. By defining the Cd cut-off energy as 0.5 eV for the ²³⁷Np(n, γ)²³⁸Np reaction, the present resonance integral is 652 ± 24 b, which is in good agreement with the JENDL-3, -3.2, ENDF/B-V, -VI, JEF-2 and Mughabghab's data. However, most of the old experimental data are, in general, larger by 24–38% than the present measurement.     

Application of BGO Scintillators to Absolute Measurement of Neutron Capture Cross Sections between 0.01 eV and 10 eV

Applying a total energy absorption γ-ray detector composed of 12 bricks (5 × 5 cm², 7.5 cm thick) of BGO scintillators, the absolute measurement of capture cross sections for Au and Sb has been made in an energy region between 0.01 and 10eV using the linac time-of-flight method. Incident thermal neutron flux was absolutely determined by using the BGO detection system with a Sm sample. To extend the neutron flux measurement from the thermal neutron region to higher neutron energies, the ¹⁰B(n, αγ) reaction was applied. Absolute capture yield for the relevant capture sample was obtained by the saturated capture yield at a large resonance of the sample.

Gold was selected to investigate the application of the BGO detection system to the absolute measurement of the capture cross sections, since the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section is a well known standard one. The result of the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section showed good agreement with the evaluated data in JENDL Dosimetry File and ENDF/B-VI. Then, the detection system was applied to the Sb(n, γ) cross section measurement. Antimony has a large scattering-to-capture cross section ratio comparing to that of gold. The result showed good agreement with the evaluated data in JENDL-3.2 and ENDF/B-VI.     

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.     

Measurement of neutron capture cross-section of indium in the energy region from 0.003 eV to 30 keV

The neutron capture cross-section of indium (In) has been measured in the energy region from 0.003 eV to 30 keV by the neutron time-of-flight (TOF) method with a 46-MeV electron linear accelerator (linac) of the Research Reactor Institute, Kyoto University. An assembly of Bi₄Ge₃O₁₂ (BGO) scintillators, which was composed of 12 pieces of BGO and placed at a distance of 12.7±0.02 m from the neutron source, was employed as a total energy absorption detector for the prompt capture γ-ray measurement from the sample. In order to determine the neutron flux impinging on the capture sample, a plug of ¹⁰B powder sample and the ¹⁰B(n,α) standard cross-section were used. The data measured by Haddad et al. (Haddad, E., Friesenhahn, S., Lopez, W.M., 1963. Report of Gulf Energy and Environmental Systems, p. 3874) seem to be in good agreement with the present measurement. Popov et al. obtained the poor energy resolution data in the resonance region with a lead slowing-down spectrometer and the consistent data with the present above about 300 eV. The experimental data measured by Kononov et al. (Kononov, U.N., Jurlov, B.D., Poletaev, E.D., Timokhov, V.M., Manturov, G.N., 1977. Report of Obninsk, pp. 22–29) and Gibbons et al. (Gibbons, J.H., Macklin, R.L., Miller, P.D., Neiler, J.H., 1961. Phys. Rev. 122, 182) showed good agreement with the present values in the higher energy region. However, the data measured by Block et al. (Block R.C., Kaushal, N.N., Hockenbury, R.W., 1972. Conference on New Developments in Reactor Physics and Shielding at Kiamesha Lake, Vol. 2, p. 1107) seem to be a little higher than the present measurement above 800 eV. The evaluated data in ENDF/B-VI, JENDL-3.2, and JEF-2.2 have been in general agreement with the present result in the relevant energy region, although the JENDL-3.2 are higher than the measurement, the ENDF/B-VI and the JEF-2.2 from 2 to about 10 keV. Most of the previous experimental and the evaluated thermal neutron cross-sections are generally close to the present value of 199.6±5.6 b at 0.0253 eV.     

Measurements of neutron induced fission cross-section for 242mAm from 0.003 eV to 10 keV using lead slowing-down spectrometer,thermal neutron facility and time-of-flight method

The neutron-induced fission cross-sections of ^242mAm were measured relative to that of ²³⁵U from 0.003 eV to 10 keV with a back-to-back type double fission chamber. These measurements were performed from 0.03 eV to 10 keV, at 0.025 eV and from 0.003 to 35 eV using the Kyoto University lead slowing-down spectrometer, the thermal neutron facility of Kyoto University Reactor and the time-of-flight method, respectively. The results were compared with both the evaluated data of JENDL-3.2, ENDF/B-VI and JEF-2.2, and the existing experimental data. The validity of the cross-section in the files was discussed through the comparison.     

Neutron capture cross section measurements of 151,153Eu using a pair of C6D6 detectors

We have measured the neutron capture cross sections of ¹⁵¹Eu and ¹⁵³Eu by the time-of-flight (TOF) method in the range from 0.005 eV to keV region using the Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC). We employed a pair of C₆D₆ liquid scintillators for the prompt capture γ-ray measurement. The pulse-height weighting technique was employed to obtain the capture yields from the γ-ray spectra of ^151,153Eu. The obtained thermal cross sections at 0.0253 eV are 9051 ± 683 b for ¹⁵¹Eu and 364 ± 44 b for ¹⁵³Eu, respectively. The resonance integrals have been derived as 3490 ± 162 b for ¹⁵¹Eu and 1538 ± 106 b for ¹⁵³Eu.

The obtained capture cross sections were compared with the previously reported experimental data and the evaluated data. The evaluated data in JENDL-4.0 and JEFF-3.2 show good agreement with the present experiment results of ¹⁵¹Eu, however, the evaluated data in ENDF/B-VII.1 are larger than the present experiment results of ¹⁵¹Eu about 10% to 20% in the energy region from 0.03 to 0.2 eV. For the neutron capture cross sections of ¹⁵³Eu, the evaluated data in ENDF/B-VII.1 and Widder's data are in good agreement with the present results in the energy region below 0.35 eV.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the Reaction 133Cs(n, γ)134m,134gCs

The measurements of the thermal neutron (2,200 m/s neutron) cross section (σ₀) and the resonance integral (I ₀) of the ¹³³Cs(n, γ;) reaction were performed by an activation method to obtain fundamental data for research on the transmutation of nuclear wastes. The cross sections for the formations of the isomeric state ^134mCs and the ground state ^134mCs were measured respectively by following the behavior of the γ-ray counting rate after the irradiation.

The thermal neutron capture cross sections and the resonance integrals of the ¹³³Cs(n, γ) reaction were determined to be 2.70±0.13 b and 23.2±1.8 b for the formation of the isomeric state ^134mCs, and 26.3±1.0 b and 275±16 b for the formation of the ground state of ^134gCs. The results for the reaction ¹³³Cs(n, γ)^134m+gCs were 29.0±l.0 b and 298±16 b, respectively. As for the thermal neutron capture cross section for the formation of ^134m+gCs, the evaluated value (29.00 b) of JENDL-3.2 agreed with the present result. The reported value by Baerg et al. was in good agreement with the present result within the limits of error on the thermal neutron capture cross section for ^134mCs. On the other hand, the resonance integral for ^134m+g Cs was 32% smaller than the experimental value by Steinnes et al.     

Correction of the thermal neutron capture cross section of 241Am obtained by the Westcott convention

There is large discrepancy among the reported experimental data of the thermal neutron capture cross section of ²⁴¹Am, where the activation measurements provided larger cross sections than those in the time-of-flight ones. The Westcott convention has been widely used for the derivation of the thermal neutron capture cross section in the activation measurements. We have estimated that this large discrepancy is due to the existence of the resonances below the cadmium cut-off energy (E_Cd ～ 0.5 eV). By reviewing the Westcott convention, we developed the correction method taking account of the contribution of the resonances near or below E_Cd. The correction term was evaluated using the JENDL-4.0. Application of the present method successfully improved the existing discrepancy of the thermal capture cross section of ²⁴¹Am.     

Method of Measuring Boron-10 Depletion in Control Blade

The neutron capture cross sections and capture γ-ray spectra of ^143,145,146Nd were measured in the neutron energy region of 10 to 90 keV and at 550 keV. A neutron time-of-flight method was adopted with a 1.5-ns pulsed neutron source by the ⁷Li(p, n)⁷Be reaction and with a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross sections were obtained with the error of about 5% by using the standard capture cross sections of ¹⁹⁷Au. The evaluated values of JENDL-3.2 and previous measurements were compared with the present results. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. An anomalous shoulder was observed around 2 MeV in the γ-ray spectra of ^145,146Nd, and the energy position of the shoulder was consistent with the systematics obtained in our previous work.     

Thermal neutron capture cross sections for the Sm(n,γ)Sm and Sm(n,γ)Sm reactions at 0.0536 eV energy

The neutron capture cross sections for the ¹⁵²Sm(n,γ)¹⁵³Sm and ¹⁵⁴Sm(n,γ)¹⁵⁵Sm reactions at 0.0536 eV neutron energy were measured using an activation technique based on the TRIGA Mark-II research reactor, relative to the reference reaction ¹⁹⁷Au(n,γ)¹⁹⁸Au. The activity was measured nondestructively using gamma-ray spectroscopy. Our measured values at this neutron energy are the first ones and are compared with 1/v based evaluated cross sections reported in the ENDF/B-VII and JENDL-3.3 libraries. The measured value for the ¹⁵²Sm(n,γ)¹⁵³Sm reaction is 0.28% lower than JENDL-3.3 and 0.48% higher than ENDF/B-VII. Our value for the production of ¹⁵⁵Sm is about 3% and 2.3% higher than the evaluated value with ENDF/B-VII and JENDL-3.3 at 0.0536 eV, respectively.     

Neutron total cross section measurements of polyethylene using time-of-flight method at KURNS-LINAC

ABSTRACT

The neutron total cross sections of polyethylene have been measured in the energy region from 0.001 eV to 40 keV by the time-of-flight (TOF) method using the Kyoto University Institute for Integrated Radiation and Nuclear Science – Linear Accelerator (KURNS-LINAC). A ⁶Li detector and a gas electron multiplier (GEM) detector have been used as a neutron detector, and the polyethylene plates of 0.31 and 0.20 cm thickness were employed for the neutron transmission measurement.

The present results were compared with the previous results and the evaluated data in JENDL-4.0. In the energy region below 0.01 eV, the present results are in good agreement with the data measured by Herdade et al. (1973) and by Granada et al. (1987). On the other hand, the evaluated data in JENDL-4.0 are larger than all the measured data. In the energy region from 0.035 to 0.15 eV, the data measured by Granada et al. and the evaluated data in JENDL-4.0 are up to about 4 ~ 6% larger than the present results.     

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%.     

Neutron Nuclear Data Evaluation of Cesium Isotopes for JENDL-4.0

For the development of JENDL-4.0, neutron nuclear data for fission product nuclides, ^{133,134,135,136,137}Cs, were revised in the incident neutron energy range from 1 eV to 20MeV by using a coupled-channels optical model (OM), and nuclear reaction models. The OM potential parameters were determined for stable ¹³³Cs to reproduce the experimental data of total and elastic scattering cross sections and angular distributions of elastically scattered neutrons. The present results reasonably reproduce measured data for (n; 2n), (n; p), (n; α), and capture reactions on ¹³³Cs. Important differences between the present results and JENDL-3.3 are found for the capture cross sections of ^134,137Cs. The cross section obtained for ¹³⁷Cs was smaller than that in JENDL-3.3. This result makes the transmutation of medium-lived ¹³⁷Cs increasingly difficult. The production probabilities of metastable states for ^134,138Cs via capture reactions on ^133,137Cs are compared with experimental values. The present result for ^134m Cs production is marginally consistent with measured data. However, a large discrepancy is recognized for ^138m Cs production. The γ-ray emission data were evaluated with available measurements, and newly compiled in JENDL-4.0. Maxwellian-averaged capture cross sections were calculated in the energy range from 1 to 103 keV, and are compared with other derived data.     

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.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the 237Np(n, γ)238Np Reaction

The thermal neutron capture cross section (σ₀) and the resonance integral (I ₀)of ²³⁷Np have been measured by an activation method to supply basic data for the study of transmutation of nuclear waste. The neutron irradiation of ²³⁷Np samples have been done at the Research Reactor Institute, Kyoto University (KUR). Samples of ²³⁷Np were irradiated between two Cd sheets or without a Cd sheet. Since ²³⁷Np has a strong resonance at the energy of 0.49 eV, the Cd cutoff energy was adjusted at 0.358 eV (thickness of the Cd sheets: 0.125 mm). A high purity Ge detector was employed for activity measurement. The reaction rate to produce ²³⁸Np from ²³⁷Np was analyzed by the Westcott's convention. Results obtained were 141.7±5.4 barns for σ₀ and 862±51 barns for I ₀ above 0.358 eV of ²⁷³Np. By setting the Cd cut-off energy at 0.358 eV considering the resonance at 0.49 eV, a smaller value of σ₀ was obtained in this work than the values reported by the previous authors.     

Measurement of keV-Neutron Capture Cross Section of 96Zr

The neutron capture cross section of ⁹⁶Zr at incident neutron energies from 15 to 100 keV has been measured by the time-of-flight method. Capture γ-rays were detected with an anti-Compton NaI(Tl) spectrometer, and the pulse-height weighting technique was applied to derive the neutron capture cross section. The present measurement provided the capture cross section as a function of incident neutron energy in the keV region. The results were compared with previous measurements and cross section data in the evaluated nuclear data libraries, JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and ENDF/B-VI.8. The present results revealed considerable underestimation of the evaluated cross sections in the high-energy region of 35–100 keV.     

Measurement and Analysis of241 Am Fission Rate Ratio Relative to235 U Fission Rate in Thermal Neutron Systems Using Kyoto University Critical Assembly

Integral measurements of²⁴¹ Am fission rate ratio relative to²³⁵ U fission rate were performed at Kyoto University Critical Assembly. The fission rates were measured using the back-to-back type double fission chamber at five thermal cores with different H/²³⁵ U ratio so that the neutron spectra of the cores were systematically varied. The measured fission rate ratios, normalized to number of atoms, were 0. 0144 to 0. 0522, with a typical uncertainty of 2%. The measured data were compared with the calculated results using MVP based on JENDL-3.2, which gave the averaged calculated-to-experimental ratio (C/E) of 0.88. Obtained results of C/E using ²⁴¹Am fission cross sections from JENDL-3/2, ENDF/B-VI and JEF2.2 showed that the latter two gave larger C/E values than those by JENDL-3.2 by about 2% and 7 to 9%, respectively. It has been found that the large difference between JENDL-3.2 and JEF2.2 arises mainly from the significant cross section difference at the vicinity of resonance at 0.576 eV, whereas the difference of thermal cross sections, especially in the range of 0.01 eV to 0.2 eV also has significant contribution for well-thermalized cores.     

Measurement of ke V-Neutron Capture Cross Sections and Capture Gamma-Ray Spectra of 147,148,149,150,152,154Sm

The neutron capture cross sections and capture γ-ray spectra of ^{147,148,149,150,152,154}Sm were measured in the neutron energy region of 10 to 90 keV and at 550 keV. A neutron time-of-flight method was adopted with a 1.5-ns pulsed neutron source by the ⁷Li(p, n)⁷Be reaction and with a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross sections were obtained with the error of about 5% by using the standard capture cross sections of ¹⁹⁷Au. The present results were compared with the evaluated values of JENDL-3.2 and previous measurements. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. An anomalous shoulder was clearly observed around 3 MeV in the γ-ray spectra of ^150,152,154Sm, and the energy position of the shoulder was consistent with the systematics obtained in our previous work.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the 129I(n, γ) 130I Reaction

To obtain fundamental data for research on the transmutation of nuclear wastes, the thermal neutron cross section and the resonance integral of the ¹²⁹I(n, γ)¹³⁰I reaction have been measured using an activation method. The neutron cross sections for the formation of the ground (5⁺) state and the isomeric (2⁺) state of ¹³⁰I were measured separately.

Six ¹²⁹I targets were irradiated for 10 min with thermal reactor neutrons; three of them containing 2.55- 2.61 kBq of ¹²⁹I were irradiated within a Cd capsule, and the other three targets containing 259–261 Bq of ¹²⁹I were irradiated without it. The Co/Al and Au/Al alloy wires were used to monitor the neutron flux and the fraction of the epithermal part (Westcott's epithermal index). The gamma-ray spectra from the irradiated samples were measured with a Ge detector.

The thermal neutron capture cross section (the 2,200 m/s neutron cross section) and the resonance integral of the ¹²⁹I(n, 7)¹³⁰I reaction were determined to be 12.5±0.5b and 15.6±0.7b for the formation of the ground state ^130gI(5+), 17.8±0.7b and 18.2±0.8b for the formation of the isomeric state ^130mI(2⁺), and 30.3±1.2b and 33.8±1.4b for the formation of ¹³⁰I (the sum of the 2⁺ and the 5⁺ states), respectively. The sum of the thermal neutron capture cross sections forming the 2⁺ and the 5⁺ states was 12% larger than the evaluated values of JENDL-3.2 and ENDF/B-VI and that reported by Roy et al. This discrepancy is explained by the population of the isomeric level.