Validation of Pb nuclear data by Monte Carlo analyses of sample reactivity experiments at Kyoto University Critical Assembly

Sample reactivity experiments on the uncertainty analyses of Pb nuclear data are carried out by substituting Al plates for Pb ones at the Kyoto University Critical Assembly, as part of basic research on Pb–Bi for the coolant. Numerical simulations of sample reactivity experiments are performed with the Monte Carlo calculation code MCNP6.1 together with four nuclear data libraries JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1, to examine the accuracy of cross-section uncertainties of Pb isotopes by comparing measured and calculated sample reactivities. A library update from JENDL-3.3 to JENDL-4.0 is demonstrated by the fact that the difference between Pb isotopes of the two JENDL libraries is dominant in the comparative study, through the experimental analyses of sample reactivity by the MCNP approach. In addition, JENDL-4.0 reveals a slight difference from ENDF/B-VII.0 in all Pb isotopes and ²⁷Al, and from JEFF-3.1 in ²³⁸U and ²⁷Al. Based on these results, further experiments are needed to investigate the uncertainties of Bi isotopes with the use of the Pb–Bi and Bi plates.     

Detail analysis of fusion neutronics benchmark experiment for iron

In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at Japan Atomic Energy Agency (JAEA)/Fusion Neutronics Source (FNS) was analyzed in detail with MCNP-4C and the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It was noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. Through the DORT calculations based on the iron data in ENDF/B-VII.0, it was found out that the first inelastic scattering cross-section data of ⁵⁷Fe in JENDL-3.3 caused the overestimation.     

Accuracy of criticality calculations made using current nuclear data libraries for MOX fuelled fast critical assemblies

Criticality calculations have been made for a set of ten mixed plutonium–uranium oxide (MOX) fuelled fast critical assemblies using the current nuclear data libraries, JEFF-3.1, JEFF-3.1.1, JENDL-3.3 and ENDF/B-VII.0. The results obtained using the different libraries are compared and conclusions drawn concerning the accuracy of criticality calculations made for MOX fuelled fast reactors.     

Validation of MVP-II with JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2 through HTC critical experiments

In order to validate MVP-II, Haut Taux de Combustion (HTC) experiments were analyzed using a code with relatively new nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2. The effective neutron multiplication factor k_eff values were obtained through analyses of all phases of the HTC experiments. Consequently, the k_eff biases evaluated for each nuclear data library were within 300 pcm. Additionally, microscopic production and capture reaction rates of major actinide isotopes were analyzed to substantiate differences among the libraries for a representative case of Phase 1 of the HTC experiments. The analysis showed that microscopic cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 were somewhat large compared to those of ENDF/B-VII.1 and JENDL-4.0 for the representative case of Phase 1.     

Benchmark experiment on titanium with DT neutron at JAEA/FNS

Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li₂O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of ²³⁵U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′_cont) reaction cross section data and resonance parameters were improved.     

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.     

Theoretical calculations and evaluations of n+23Na reaction

The data for neutron-induced reactions are indispensable in a lot of applications of nuclear science and technologies. All reaction cross sections, angular distributions, energy spectra, and double-differential cross sections of neutron, proton, deuteron, triton, and alpha-particle emissions are consistently calculated and analyzed for n+~(23)Na reactions at incident neutron energies below200 Me V, based on nuclear theoretical models. The calculated results are compared with the experimental data and the evaluated data in the ENDF/B-VII, JENDL-4.0,and JEFF-3.2 libraries. In most cases, the calculated results describe the corresponding experimental data well. At the resonance energy region, evaluated experimental data are adopted to fit to the resonance structures.     

Evaluation of neutron cross sections for a complete set of Dy isotopes

Neutron cross sections for a complete set of Dy isotopes, ^{156,158,160,161,162,163,164}Dy, were evaluated in the incident energy range from 10⁻⁵ eV to 20 MeV. In the low energy region, including thermal and resolved resonances, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. In the unresolved resonance region we performed additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data. In the fast neutron region, we used the nuclear reaction model code EMPIRE-2.19 with the model parameters adjusted to the experimental data. The results are compared with the available experimental data and with the existing nuclear data libraries, including ENDF/B-VI.8 and JEFF-3.1. The new evaluations are suitable for neutron transport calculations and they were adopted by the new US evaluated nuclear data library, ENDF/B-VII.0, released in December 2006.     

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.     

Calculations of Maxwellian-averaged cross sections and astrophysical reaction rates using the ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and ENDF/B-VI.8 evaluated nuclear reaction data libraries

We have calculated the Maxwellian-averaged cross sections and astrophysical reaction rates of the stellar nucleosynthesis reactions (n, γ), (n, fission), (n, p), (n, α), and (n, 2n) using the ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and ENDF/B-VI.8 evaluated nuclear reaction data libraries. These four major nuclear reaction libraries were processed under the same conditions for Maxwellian temperatures (kT) ranging from 1 keV to 1 MeV. We compare our current calculations of the s-process nucleosynthesis nuclei with previous data sets and discuss the differences between them and the implications for nuclear astrophysics.     

Resonance analysis of 151,153Eu from neutron capture cross section measurements in the energy range from 1 to 20 eV

The neutron capture cross sections of Europium-151 and Europium-153 have been measured by the time-of-flight method in the energy range from 0.005 to 100 eV using the Kyoto University Research Reactor Institute-Linear Accelerator (KURRI-LINAC). An assembly of Bismuth Germanate (BGO) scintillators was used to detect the prompt capture of γ rays. The absolute values of the neutron capture cross sections of ¹⁵¹Eu and ¹⁵³Eu were deduced by normalizing the thermal capture cross sections in JENDL-4.0 and ENDF/B-VII.1, respectively. Then, we have obtained the resonance parameters of 20 resonances in ¹⁵¹Eu and 17 resonances in ¹⁵³Eu using the code SAMMY.

For the 3.36-eV resonance of ¹⁵¹Eu, the evaluated resonance peak area in JENDL-4.0 is about 95% smaller than the present result. For the 7.00-, 7.22-, and 7.42-eV resonance; we confirmed that there are significant differences between the measured peaks and evaluated peaks in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2. For the ¹⁵³Eu, the evaluated resonance peak areas in JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2 are about 15% larger than the measured resonance peak areas at the 2.46-, 3.29-, and 3.94-eV resonances.     

Neutronics Calculations of HTTR with Several Nuclear Data Libraries

Benchmark calculations for several HTTR core states were performed with four cross-section sets which were generated from JENDL-3.3, JENDL-3.2, ENDF/B-VI.8 and JEFF-3.0 using a continuous energy Monte Carlo code MVP. The core states were a critical approach in which an annular core was formed at room temperature and solid cores at room temperature and at full power operation. Study of k_eff discrepancies caused by difference of the nuclear data libraries and identification of nuclides which have large effects on the k_eff discrepancies were carried out. Comparison of the respective k_eff from calculations and experiments was also carried out. As the results, for each of the HTTR core states, JENDL-3.3 yields a k_eff agreeing with the experiments within 1.5%Δk, JENDL-3.2 yields k_eff agreement within 1.7%Δk, and ENDF/B-VI.8 and JEFF-3.0 yield k_eff agreement within 1.8%Δk. There is little k_eff discrepancy between ENDF/B-VI.8 and JEFF-3.0. The k_eff between JENDL-3.3 and JENDL-3.2 is caused by difference of ²³⁵U data and has temperature dependency. The k_eff discrepancy between JENDL-3.3 and ENDF/B-VI.8 or JEFF-3.0 is mainly caused by difference in graphite data.     

Activation cross-section for reactions induced by 14 MeV neutrons on natural silver

Cross-sections for (n, 2n), (n, p), and (n, α) reactions have been measured on silver isotopes at the neutron energies from 13.5 to 14.8 MeV using the activation technique in combination with high-resolution γ-ray spectroscopy. Corrections were made for the literature cross-sections of ¹⁰⁹Ag(n, 2n) ^108mAg reaction with incorrect half-life of product ^108mAg. Neutrons were produced via the ³H(d, n)⁴He reaction using solid TiT. The neutron fluences were determined using the monitor reaction ²⁷Al(n, α)²⁴Na. The neutron energy in this measurement was determined by cross-section ratios for the ⁹⁰Zr(n, 2n) ^89m+gZr and ⁹³Nb(n, 2n)^92mNb reactions. Data are reported for the following reactions: ¹⁰⁹Ag(n, 2n)^108mAg, ¹⁰⁷Ag(n, 2n)^106mAg, ¹⁰⁹Ag(n, p)^109m+gPd, and ¹⁰⁹Ag(n, α)^106mRh. The cross-sections were discussed and compared with experimental data found in the literature, and with the comprehensive evaluation data in ENDF/B-VII, JENDL-3.3, and JEFF-3.1/A libraries.     

Calculation and evaluations for n+Cu reactions

All cross sections of neutron induced reactions, angular distributions, energy spectra and double differential cross sections are consistently calculated and analyzed for n+^63,65,nat.Cu reactions at incident neutron energies below 200 MeV based on the nuclear theoretical models. The optical model, preequilibrium and equilibrium reaction theories, the distorted wave Born approximation theory are used. Theoretical calculated results are compared with existing experimental data and the evaluated results in ENDF/B-VII and JENDL-3 libraries. The optical model potential parameters are obtained according to the experimental data of total, nonelastic scattering cross sections and elastic scattering angular distributions.     

Benchmark experiment of large-angle scattering reaction cross section of iron at 14 MeV using two shadow bars – Comparison of experimental results with ENDF/B-VIII –

ABSTRACT

It is important to perform neutron transport simulations with accurate nuclear data in the neutronics design of a fusion reactor. However, absolute values of large-angle scattering cross sections vary among nuclear data libraries even for well-examined nuclide of iron. Benchmark experiments focusing on large-angle scattering cross sections were thus performed to confirm the correctness of nuclear data libraries. The series benchmark experiments were performed at a DT neutron source facility, OKTAVIAN of Osaka University, Japan, by the unique experimental system established by the authors’ group, which can extract only the contribution of large-angle scattering reactions. This system consists of two shadow bars, target plate (iron), and neutron detector (niobium). Two types of shadow bars were used and four irradiations were conducted for one experiment, so that contribution of room-return neutrons was effectively removed and only large-angle scattering neutrons were extracted from the measured four Nb reaction rates. The obtained experimental results were compared with calculations for five nuclear data libraries including JENDL-4.0, JEFF.-3.3, FENDL-3.1, ENDF/B- VII, and recently released ENDF/B-VIII. It was found from the comparison that ENDF/B-VIII showed the best result, though ENDF/B-VII showed overestimation and others are in large underestimation at 14 MeV.     

板状Bi样品泄漏中子和γ谱实验测量与模拟计算

通过飞行时间法,测量了氘氘脉冲中子与不同厚度209Bi样品作用后61°和119°方向的泄漏中子飞行时间谱和泄漏γ能谱,样品尺寸分别为30 cm×30 cm×5 cm、30 cm×30 cm×10 cm和30 cm×30 cm×15 cm。采用BC501A液体闪烁体探测器测量0.8～3.2 MeV能区的泄漏中子飞行时间谱,钾冰晶石探测器（CLYC）测量0.2～0.8 MeV的泄漏中子飞行时间谱和泄漏γ能谱。用MCNP-4C程序对泄漏中子飞行时间谱和泄漏γ能谱进行了模拟计算,²⁰⁹Bi的评价中子核数据分别采用了CENDL-3.1库、ENDF/B-Ⅷ.0库、JENDL-4.0库以及JEFF-3.3库中的数据,模拟结果分别与实验结果进行比较分析,研究结果表明,泄漏中子谱CENDL-3.1库的模拟结果在119°方向弹性峰位置有较严重的低估现象,JENDL-4.0库在1.5 MeV附近（第二非弹能区）有一定高估,而在低能区有明显低估;泄漏γ能谱JENDL-4.0库和JEFF-3.3库的模拟结果与实验结果偏差明显,而CENDL-3.1库符合较好。     

Problems of lead nuclear data in fusion blanket design

In an irradiation experiment using a LiAl/Pb assembly, we found out that the neutron flux inside the assembly calculated with JENDL-3.3 underestimates an experimental value in the 10–16 MeV region by around 30% and that in the 0.5–5 MeV region by around 15%, while the calculated flux with JEFF-3.1 overestimates the measurement in the 5–10 MeV region by around 20%. In order to reveal a reason of the discrepancy, problems of the nuclear data libraries for lead were investigated. As a result, the following problems of the evaluated libraries were pointed out: the cross-sections of the (n,2n) reaction in JENDL-3.3 for lead isotopes are too large and cause a significant underestimation of the neutron flux above 10 MeV, which appeared in the analysis of the above experiment. Inelastic scattering data for ²⁰⁸Pb in JENDL-3.3 reproduce previous experimental double-differential cross-section data most well. However, those for the other lead isotopes have some problems and cause a large underestimation of the neutron flux from 0.5 to 5 MeV. The reason of the overestimation in the energy region of 5–10 MeV with JEFF-3.1 is still unclear.     

Calculation and analysis of n+Zr reactions in the En ⩽ 250 MeV energy range

All of reaction cross sections, angular distributions, energy spectra, γ-ray production cross sections, and the double differential cross section for neutron, proton, deuteron, triton, helium and alpha emission are calculated and analyzed for n+^{90,91,92,94,96,nat}Zr at incident neutron energies from 0.1 to 250 MeV. The optical model, intranuclear cascade model, the unified Hauser–Feshbach theory and the exciton model which included the improved Iwamoto–Harada model are used. Theoretical calculated results are compared with existing experimental data and other evaluated data from ENDF/B-VI.8, ENDF/B-VII.0 and JENDL-3.3. The optical model potential parameters are obtained according to the experimental data of total, nonelastic cross sections and elastic scattering angular distributions.     

Analysis of Measured Isotopic Compositions of High-Burnup BWR MOX Fuel

Analysis of measured isotopic compositions of four high-burnup BWR MOX fuel samples was performed by using a general-purpose neutronic calculation code SRAC and a continuous-energy Monte Carlo burnup code MVP-BURN. The initial Pu fissile content of the samples was 5.52 wt%, and the burnups ranged from 50 to 80 GWd/t. It is confirmed that a geometrical model including the effect of UO₂ assemblies adjacent to the MOX assembly is necessary in the burnup calculations to obtain accurate calculated isotopic compositions. The calculated results of MVP-BURN with JENDL-3.3 taking such effect into account show more accurate results for major actinides (U, Pu, and Am isotopes) and most fission products than those of infinite assembly calculations. The paper also shows the results calculated using SRAC with JENDL-3.3, ENDF/B-VII, and JEFF-3.1.     

Benchmarking of evaluated nuclear data for uranium by a 14.8 MeV neutron leakage spectra experiment with slab sample

An experimental system for benchmarking of evaluated nuclear data by a 14.8 MeV neutron leakage spectra measurement with slab sample has been setup at China Institute of Atomic Energy (CIAE). The first measurement of the neutron leakage spectra from a 10 × 10 × 5 cm pure uranium slab sample at 45° and 135° was reported. The experimental results were compared with the calculated ones by MCNP-4C simulation, using the evaluated data of uranium from the CENDL3.0, ENDF/B-VII and JENDL3.3 libraries. This work shows that the apparatus and the data processing procedures work well. And more experiments will be done in the future for benchmarking of the evaluated data, especially for the CENDL data.