Table of Content

    20 May 2022, Volume 56 Issue 5
    Nuclear Data Evaluation and Chinese Evaluated Nuclear Data Library
    GE Zhigang, XU Ruirui, LIU Ping
    2022, 56(5):  783-797.  DOI: 10.7538/yzk.2022.youxian.0221
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    Nuclear data are the essential data for nuclear physics scientific research, the development of nuclear energy as well as the applications of nuclear technology, and it is an important bridge connecting fundamental nuclear physics research with nuclear engineering and nuclear technology application. Nuclear data play important roles in national defense, nuclear security, national economic development and basic science research etc. A nuclear data research system which including nuclear data measurement, nuclear data evaluation, model calculation, nuclear data library development and nuclear data integral benchmarking, has been established in China since 1960’s. A lot of highreliability nuclear data, highlevel methodologies and libraries have been achieved with the effect of the national nuclear research network since China Nuclear Data Center established in 1975. All the output feed the needs of nuclear energy and nuclear technology applications, and nuclear data research also prompted the development of nuclear physics study in China. The nuclear data evaluation is an important part of nuclear data research, and it involves experimental data evaluation, nuclear model calculation, recommendation, complicated physics checking of complete set of data, evaluated data file integral verification and validation, etc. This paper briefly introduces the main research procedure of nuclear data evaluation, corresponding theoretical models and calculation codes, and the method of the establishment and validation testing for the latest version of Chinese Evaluated Nuclear Data Library (CENDL32). Some notable advances in CENDL32 are noted, such as new evaluation of the nn and np scattering cross section, model dependent covariance data for main reaction cross sections of some fission product nuclides and the updated evaluation of nuclear reaction data of several key nuclides, such as 7Li, 233,235U, 56Fe, 240Pu, etc. The evaluated nuclear data integral verification and validation is the essential links in the application of nuclear data to nuclear engineering. In order to verify the physical rationality, systematic comparisons between CENDL-3.2 and other major evaluated libraries (e.g. ENDF, JENDL, BROND, JEFF and TENDL) have been implemented using a suite of criticality benchmarks from the International Criticality Safety Benchmark Evaluation Project, and compared with available experimental data. This paper also briefly introduces relevant benchmarking testing and application results based on the CENDL32, and the χ2 value obtained implies that CENDL32 has a potential remarkable improvement of predictions for 235U and Pu systems. The CENDL32 library also has good performances on the reactor physics simulation of PWR, VENUS3 shielding benchmark by ARES transport code, HTR10 benchmark, etc. In general, integral validation performance of the CENDL32 library is improved relative to the previous CENDL31 library.
    Measurement of Post-neutron Mass Distribution in Thermal-neutron-induced Fission of 239Pu
    LIU Chao, LIU Shilong, YANG Yi, ZHAO Kunling, CHEN Yongjing, SU Yang, ZHANG Kai, FENG Jing
    2022, 56(5):  798-804.  DOI: 10.7538/yzk.2022.youxian.0054
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    Nuclear fission is one of the most complex physical processes. Up to now, there is still a lack of theory that can well describe the prefission and postfission processes. Mass yields in the neutroninduced fission of 239Pu provide fundamental and significant data support for nuclear energy utilization, but the independent fission yields are scarce and far from precise. In addition, the current evaluation work for fission yields partially refers to the experimental data based on radiochemical methods and mass spectrometric, it will make sense to compare the evaluation data with the data obtained by direct systematic measurement. In order to determine independent mass yield distributions precisely, the fission fragments identification spectrometer (FFIS) device was developed. The axial grid ionization chamber and timing detectors based on microchannel plates were designed and tested for determining the kinetic energy and timeofflight of fission fragments, respectively. Specifically, the intrinsic energy resolution for the ionization chamber was measured to be about 0.5% for 80 MeV 63Cu particles, and the time resolution was about 157 ps (FWHM) determined by 241Am source. By directly coincident measurement of energy and velocity of the fission fragments on their flight path, the postneutronemission masses can be obtained according to the classical formula for kinetic energy. The mass yield distribution in the thermalneutroninduced fission of 239Pu was measured at the inhospital neutron irradiator (IHNI1). With the flight trajectory of 10045 cm, the relative uncertainty of the flight path length is better than 01%, and the time resolution is about 02%. When using the Ev method to measure mass distribution, two vital issues need to be resolved. For one thing, the nonlinear energy response of the detector affects the accuracy of the mass calculation. In this work, energy losses in the carbon foil of the stop timing detector and the silicon nitride window of the ionization chamber were corrected eventbyevent based on Geant4 calculation. For another thing, the scattering of heavy ions or the inhomogeneous of the detector materials may make the recorded energy to be underestimated. These abnormal data points are hard to eliminate in the valley region, which makes the mass yields of symmetric fission fragments being overestimated, and improving statistics may relatively reduce this kind of effect. Taking into account the correction uncertainty, the energy resolution for the light fission fragment is better than 07%, and that is approximately 1% for heavy fragments peak. Thus, the mass resolution is estimated to be 1 amu at 99 amu, and 15 amu at 138 amu.
    Radiative Capture Cross-section Measurement and Resonance Parameter Analysis for 169Tm Based on White Neutron Source
    REN Jie, RUAN Xichao, WANG Jincheng, BAO Jie, LUAN Guangyuan, ZHANG Qiwei, HUANG Hanxiong, NIE Yangbo
    2022, 56(5):  805-815.  DOI: 10.7538/yzk.2022.youxian.0125
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    Radiative neutron capture cross-section and the resonance parameter are very important nuclear data in the field of the R&D of accelerator driven systems, the transmutation of nuclear waste and nuclear astrophysics. In the recent years, radiative neutron capture crosssections of important nuclides were mostly measured based on the white neutron source facilities, such as the CERN n_TOF, GELINA, LANSCE and so on. The back streaming white neutron beam line (Backn) of China Spallation Neutron Source (CSNS) is the first white neutron facility in China, which can provide pulsed neutron beam with 25 Hz frequency in the energy region from thermal neutron to hundreds MeV. The main application of Backn is to perform nuclear data measurement, such as crosssection, reaction yield, nuclear structure, and so on. The C6D6 liquid scintillation detector is widely used for radiative neutron capture crosssection measurements on timeofflight facilities, because of its low neutron sensitivity and good time response. A C6D6 detection system with four detectors was installed at Backn to perform radiative neutron capture crosssection measurement, especially in the resonance energy region. In this work, the measurement of 169Tm(n,γ)170Tm reaction was performed with the C6D6 detection system at the Backn. The stable isotope of thulium, 169Tm, is one of the fission product poisons, which makes the accurate neutron radiative capture crosssection of 169Tm be of major significance for fission and fusion reactor design. Besides, the radioactivity induced by the reactions of 169Tm(n,γ)170Tm and 170Tm(n,γ) 171Tm makes the 169Tm to be an ideal spectrumsensitive activation detector for the neutron intensity diagnosis, in which the precise neutron capture crosssection is of critical importance. In the present measurement, the pulse height weighting technique was used to make detection efficiency be proportional to the excitation energy of the compound nucleus. The black filter method was used to determine the experimental background due to scattered neutron and inbeam gamma ray. The saturated resonance method was used to normalize the capture yield. The Rmatrix code SAMMY was used to fit the capture yield of 169Tm to obtain the resonance parameters, such as resonance energy, neutron width, radiative capture width, and so on. Then the radiative capture crosssection was calculated with the measured resonance parameters and ReichMoore approximation. For most energy region between 1 eV and 100 eV, the present measurement result is consistent with the recommended value of ENDF/BⅧ.0, but there were still some discrepancies between the measured crosssection and the evaluated data. According to the current results, the measurement and data analysis methods used in this work are suitable for the measurement of radiative capture crosssection and resonance parameter. However, in order to perform high precision measurement of radiative capture crosssection at Backn, the neutron flux and the energy resolution function still need to be determined more accurately.
    Measurement of 209Bi Neutron Inelastic Scattering Cross Section with Prompt γ-Ray Method
    SUN Qi, WANG Zhaohui, ZHANG Qiwei, HUANG Hanxiong, REN Jie, RUAN Xichao, LIU Shilong, BAO Jie, LUAN Guangyuan, DING Yanyan, CHEN Xiongjun, NIE Yangbo, LIU Chao, ZHAO Qi, WANG Jincheng, HE Guozhu, DU Shubin
    2022, 56(5):  816-824.  DOI: 10.7538/yzk.2022.youxian.0189
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    Bismuth is widely used in advanced nuclear reactors such as lead-bismuth eutectic (LBE) fast reactors, space reactors and so on. Its neutron nuclear data, especially the inelastic scattering cross sections have significant impact on safety and economics of these nuclear facilities. A facility for measurement of neutron reaction cross sections using prompt γ ray method based on the HI13 tandem accelerator was established at China Institute of Atomic Energy (CIAE). Neutrons were produced using a D2 gas target bombarded with deuterons at 65, 80 and 95 MeV. The neutron target was positioned in the center of a shielding in size of 2 m×2 m×2 m composed of concrete, iron, lead and borated polyethylene. Neutrons were extracted at 0° with respect to the deuteron beam through a collimator made of copper, iron, polyethylene and lead. Neutron inelastic scattering cross sections at 90, 105 and 120 MeV were measured experimentally for 209Bi with a 209Bi sample in size of 50 mm×4 mm. A natural titanium sample in size of 50 mm×1 mm was used as the reference sample. By measuring production cross sections of the 983.5 keV γ rays produced when neutrons scattered inelastically off 48Ti nuclei, a normalization factor can be determined. 4 Clover detectors were used to measure produced γ rays. They were placed at 30°, 70°, 110° and 150° with respect to the deuteron beam. Lead shielding in thickness of approximately 2 cm was used for detector head to shield scattered γ rays. Timeofflight method was used to determine neutron energy, thus neutrons produced by DD breakup could be discriminated with neutrons produced by DD fusion. The energy and absolute efficiency were calibrated using 152Eu, 60Co, 22Na and 133Ba standard γ sources with known radioactive activities. A 508 cm×508 cm liquid scintillation detector was positioned at the end of neutron beam line to monitor the neutron fluxes passing through the samples. Two specific measurement angles at 110° and 150° were selected to obtain angleintegrated γ production cross sections by performing a simple linear summation of the partial cross sections measured at these angles. Talys 195 code was used to calculate ratios of γ production cross sections and total inelastic scattering cross sections. The experiment results were compared with other experiment data, data retrieved from ENDF/BⅧ0, JEFF33, JENDL40, ROSFOND2010 and CENDL31 evaluated nuclear data libraries as well as calculation by Talys 195 code with the default parameters. The results show that the tendency of these 3 measured energy points is similar to these results. For cross sections measured at 90 and 105 MeV, the results are closer to data calculated with Talys 195 code. For cross section measured at 120 MeV, it fits the ROSFOND2010 evaluated data better.
    Cross Section Measurement Method for (n,α) Reaction of Gas Sample Based on Combination of Experiment and Simulation
    HU Yiwei, CUI Zengqi, LIU Jie, BAI Haofan, XIA Cong, JIANG Haoyu, CHEN Jinxiang, ZHANG Guohui
    2022, 56(5):  825-834.  DOI: 10.7538/yzk.2022.youxian.0110
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    A method for measuring (n,α) reaction cross sections of the gas sample induced by fast neutrons using a gridded ionization chamber (GIC) was established. There are several advantages using the GIC to measure gas samples. With this method, the (n,α) reaction cross sections of inert gases which are hard to exist as a solid can be measured. And the (n,α) reaction cross section for α particle emitting in all directions can be directly measured by using the gas sample. Furthermore, compared with using solid samples, using gas samples have far more nuclei in the measurement samples. Therefore, it is feasible to use neutron source with a better energy resolution to measure the resonance cross sections with gas sample. However, there are many problems to be solved when measuring (n,α) reaction cross section of gas sample by using the GIC. On the one hand, the nuclear reaction event will be affected by the wall effect, when the nuclear reaction is generated near the electrodes of the GIC or at the edge of the electric field. The wall effect will cause the measured event to locate outside the event area which cannot be counted. Not only is the number of nuclei of the gas samples involved in the (n,α) reaction difficult to determine, but the number of events, which are affected by the background or located below the threshold, is also difficult to count. On the other hand, the neutron fluence involved in the measurement of (n,α) reaction cross section of gas sample is difficult to measure. Aiming at solving the problems of the determination of the sample nuclei number, the measurement of neutron fluence, and the selection of events, which are all difficult for the gas sample, this method included three steps: First, simulation calculation was performed to select a suitable effective volume of the gas samples, so that the event is free of background and easy to be counted. Second, the neutron collimator and the timing information of cathode and anode signals were used to select and count the events generated in the effective volume of the gas sample. Third, combining the fission fragment measurement of the 238U(n,f) reaction and the simulation based on SuperMC, the neutron fluence through the effective volume was obtained. Measurements of relative cross sections of the 14N(n,α0) reaction using gas sample were performed at En=471, 487, 500, 512, 529, and 545 MeV. The results are in good agreement with those of ENDF/BⅧ0 which verified the present measurement method.
    Measurement of 78Kr(n,2n)77Kr Reaction Cross Section Induced by 14.8 MeV Neutron
    LIANG Jianfeng, XIE Feng, RUAN Xichao, BAO Jie, CHEN Xiongjun, SHI Quanlin, LI Xuesong, YU Gongshuo, KANG Tai
    2022, 56(5):  835-842.  DOI: 10.7538/yzk.2021.youxian.1030
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    The measurement of (n,2n) reaction cross section induced by fast neutron plays an important role in nuclear reaction mechanism research and nuclear technology applications. Experiments have been carried out over the last several decades. However, most experiments focused on solid samples including 89Y, 151Eu, 203Tl, 232Tl and so on, and there are few reported results about gaseous samples such as xenon and krypton. The 78Kr(n,2n)77Kr reaction cross section was poorly studied, and this work aimed to measure the cross section induced by 148 MeV neutron. The measurement was performed by the activation method, and the experiment was carried out at the Cockcroft Walton Accelerator of China Institute of Atomic Energy (CIAE). A cell made of PMMA was used as the gas container, and the 78Kr gas (enriched to 995% abundance) was injected into the cell using a syringe. The mass of 78Kr gas sample was determined by weighing method. Quasi monoenergetic neutron beams with a yield of about 3×1010 s-1 were produced via the T(d, n)4He reaction (Q=17.6 MeV). The ion beam current was about 250 μA, and the average energy of D+ particles was about 300 keV. The neutron flux determination was accomplished by attaching two 93Nb foils (same diameter as the cell and thickness of 50 μm) with purity of 9999% to the upstream and downstream faces of the cell. The sample was placed at 0° angel relative to the incident D+ beam direction and centered about the TTi target at a distance of about 10 cm, where the neutron flux was 2×107 cm-2·s-1. After irradiation, the activities of 77Kr and 92Nbm were measured by offline gammaray spectrometric technique using a precalibrated HPGe detector on the surface of the detector end cap due to their low activity. To obtain detection efficiencies, another irradiation experiment was carried out individually to produce 77Kr and 92Nbm with high activities, which were used to measure the count rate ratio of the characteristic γray at a distance of 25 cm and on the surface. The efficiencies at the position of 25 cm were obtained by interpolating the efficiency curve which was pre-calibrated using various standard γ sources. Besides, Monte Carlo codes were written to calculate correction factors caused by geometry difference and neutron flux difference. As a result, the cross section was determined to be (464±19) mb, and was compared with the literature data and evaluations. The result of this work is much higher than that of previously reported. The result is helpful to the analysis and evaluation of 78Kr(n,2n)77Kr reaction cross section.
    Development of Radioactivity Measurement Device of 196Au and 44Scm under Strong Interference
    JIANG Wengang, XIE Feng, BAI Tao, MA Yue, LIANG Jianfeng, ZHANG Xiaolin, XU Jiang, HE Xiaobing, SHI Quanlin
    2022, 56(5):  843-848.  DOI: 10.7538/yzk.2021.youxian.1057
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    In the spectral mean cross section measurements of 197Au(n,2n)196Au and 45Sc(n,2n)44Scm, the activation products 196Au and 44Scm have low activity and shorter halflife, but the activation products 198Au and 46Sc from neutron capture reactions of 197Au(n,g) and 45Sc(n,g) have strong activity and longer halflife, which results in poor signaltonoise ratio of gammaray characteristic peaks, and increases the measuring uncertainty of 196Au and 44Scm activity. In order to accurately measure 196Au and 44Scm, a radioactive activity measuring device based on anticoincidence principle was developed. The main techniques used were as follows: Firstly, liquid flash measurement system has been used to detect the betaray emitted by the decay of 198Au and 46Sc, basing on the anticoincidence principle, the betaray would be used to suppress Compton background caused by gammarays emitted by the two nuclides above; Secondly, a portable antiCompton spectrometer has been constructed using a ring sodium iodide crystal, and the Compton background caused by the decay of 198Au and 46Sc has been further suppressed by the anti-Compton principle. Thirdly, a thin layer of lead shield has been placed between the radioactive source distributed evenly in the liquid scintillation and a wellshaped high purity germanium (HPGe) crystal to absorb the Xrays emitted by the decay of the 196Au, as results, the coincidence between the Xrays and the characteristic gammarays emitted by 196Au in the HPGe crystal reduces significantly, and the detection efficiency of this measurement device for 196Au increases too. By using the digital spectrometer, the liquid flash measurement system and the antiCompton spectrometer integrated into one, the anticoincidence effect could be superimposed and good experimental results were obtained. For 198Au and 46Sc, the anticoincidence inhibition ratio is 871% and 869% respectively. The detection efficiency for 44Scm is 313%. After increasing the shielding layer, the detection efficiency for 196Au increases greatly, from 5.5% when the shielding layer is not available up to 11.9%. Under the condition of strong isotope interference (198Au or 46Sc of 02 MBq), the minimum detectable activity (MDA) of 24 h detecting to 196Au is less than 07 Bq, and to 44Scm is less than 03 Bq. Applying the double anticoincidence measuring device established in this paper to the activity measurement of 196Au and 44Scm could effectively reduce the uncertainty of 197Au(n,2n)196Au and 45Sc(n,2n)44Scm average crosssection measurement.
    Benchmark Experiment on Slab Nb with D-T Neutron for Validation of Evaluated Nuclear Data
    ZHAO Qi, NIE Yangbo, DING Yanyan, REN Jie, RUAN Xichao, HU Zhijie, XU Kuozhi
    2022, 56(5):  849-859.  DOI: 10.7538/yzk.2022.youxian.0020
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    In the fusion reactor, the first wall, blankets and superconducting magnets are directly exposed to 14.5 MeV neutrons produced by fusion reaction. Some of the characteristics of Nb metal such as high melting point, good thermal conductivity and low neutron capture cross section make Nb often used in these parts of reactor. Therefore, the quality of the evaluated nuclear data of Nb is required more accurate to ensure the safety and economy in the process of the fusion reactor working. Benchmark experiments are an important way to test and evaluate the reliability of data. A benchmark experiment on Nb was carried out with 145 MeV DT neutron source in China Institute of Atomic Energy. Slab Nb with thickness of 5 cm, 10 cm and 15 cm was selected as samples. For each thickness of Nb, the leakage neutrons from 08 MeV to 16 MeV were measured using the timeofflight method in the 60° and 120° directions. The Monte Carlo neutron transport code MCNP4C was used to simulate the leakage neutron spectra under the same experimental conditions, the Nb data was retrieved from CENDL-3.1, ENDF/B-Ⅷ.0 and JENDL-4.0 evaluation data libraries. The whole time of flight spectrum was divided into the parts contributed by four reaction channels including (n, el), (n, inl)D, (n, inl)C, and (n, 2n). The simulation integral value of each part was compared with the experimental integral value, and the C/E was obtained as the standard to verify the reliability and accuracy of the relevant data. As the results of the benchmark experiment on Nb, it can be found that: Except a little overpredict at the simulation result of elastic scattering at 120°, the Nb nuclear data of the JENDL40 data library is in good agreement with experiment results. One of the problems of CENDL31 library are that it gives too large discrete inelastic scattering cross section. The other problem may be that the secondary neutron energy spectrum given by continuous level of inelastic scattering and (n, 2n) reaction is soft. The simulation results of the ENDF/B-Ⅷ.0 data library make a huge difference. It can be seen from the comparison results that the cross section of the energy spectrum in the discrete inelastic scattering part is high and the energy spectrum given in the continuum inelastic scattering part is also problematic. The discrepancies of elastic scattering peak between 60° and 120° are considered as caused by the wrong angular distribution, especially the underestimate at 60°. So the ENDF/BⅧ.0 library has a lot to improve. In a conclusion, these three libraries are all need to be improved to get more accurate data.
    Calculation of Neutron Induced Nuclear Reaction on Deuteron with Faddeev Equation
    LI Yan, PANG Danyang, CHEN Wendi, TAO Xi, XU Ruirui, GE Zhigang
    2022, 56(5):  860-868.  DOI: 10.7538/yzk.2022.youxian.0108
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    Neutron induced nuclear reaction on light nuclei is an important subject in nuclear data research. The reaction of the n+d three nucleons system is also an important platform for examining the nucleonnucleon interaction theories. The progress on calculations of neutron induced nuclear reactions on deuterons with the FaddeevAGS equations was introduced in this paper. The equations were solved with the wave packet continuum discretization method in momentum space. The calculated results include the angular distributions of elastic scattering, doubledifferential cross sections of the emitting neutrons and protons from breakup reactions, excitation functions of the total elastic scattering cross sections, and the excitation function of the total breakup reaction cross sections. The elastic scattering cross sections were calculated with the separable Paris nucleonnucleon interaction for incident energy below 20 MeV and with the Nijmegen potential for incident energy above 20 MeV. A comparison between results with the Paris potential, which consists of both s and pwave nucleonnucleon (NN) interactions, and those with the swave Yamaguchi potential for the n+d elastic scattering suggest that NN interactions that are beyond the swave are necessary to describe the experimental data, especially at higher incident energy. Due to practical reasons, calculations of the breakup cross sections, namely, the double differential cross sections of the emitting neutrons and protons are limited with a separable swave Yamaguchitype nucleonnucleon interaction, which limited the calculation of the double differential cross sections to below 20 MeV. Fair agreements are found between the results in this paper and the experimental data and with the evaluated data in CENDL32, ENDF/BⅧ.0, JENDL5, and JEFF33.
    Microscopic Study on Induced Fission Dynamics of 258Fm within Covariant Density Functional Theory
    CHEN Shengyuan, LI Zeyu, CHEN Yongjing, LI Zhipan
    2022, 56(5):  869-878.  DOI: 10.7538/yzk.2022.youxian.0151
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    Nuclear fission is an import object in nuclear physics and engineering due to its crucial role in the military, energy application and industry, agriculture, medical and other fields. In recent years, nuclear fission attracts more attention due to the following reasons: Nuclear fission plays a crucial role in determining the stability of superheavy elements, and is one of the primary mechanisms that terminate nucleosynthesis and determine the abundance of the synthesized nuclides; It is also an important mechanism used to produce shortlived exotic nuclides and fission of exotic nuclei far from stability also provides a new and effective way to study exotic nuclei. Generally, heavy nucleus contains hundreds of nucleons which are coupling with each other by nucleonnucleon manybody interaction. Due to the complication of lowenergy heavy nuclear fission, it is out of reach for ab initio calculation or interacting shellmodel. Modern microscopic approaches are based on the framework of nuclear energy density functionals(NEDFs) which consider the Pauli exclusion principle and nucleon-nucleon interaction simultaneously and include the shell effect and quantum manybody effect selfconsistent. In our previous work, the timedependent generator coordinator method plus Gaussian overlap approximation (TDGCM+GOA) has been implemented based on the covariant density functional theory (CDFT) and applied to describe the fission dynamics of 226Th where three peaks were observed in the fragment yield distribution. In this work, the dynamics of lowenergy induced fission of 258Fm was analyzed by using the TDGCM+GOA based on CDFT, mainly focusing on the potential energy surface(PES), total kinetic energies (TKE) of the fragments, and fragment mass yields. A remarkable symmetric fission valley is found in the potential energy surface, and thus both the TKE distribution and fragment mass yields present a single symmetric peak structure. The scission lines show an obvious separation between the number of the nucleons in the neck Qn=4,3,2, and an almost coincidence between Qn=2,1. The evolution of wave function probability distribution with time was presented, and the results show that most of the wave function flow through the scission line along the symmetric fission path, while a negligible part of the wave function flow through the scission line along the cluster emission path. The TKE distribution and fragments mass yields distribution were analyzed with the variance of the number of nucleons in the neck. As the number of nucleons in the neck decreases from 4 to 1, the peak of TKE distribution becomes narrow and the maximum of mass yield increases from 988% to 1028%. In addition, the influence of initial excitation energy on the fragments mass distribution was analyzed, and the results show that the peak of mass yield becomes lower, i.e., from 988% to 855%, as the excitation energy of the initial state increases from 83 MeV to 173 MeV.
    Nuclear Fission in Low Excitation Energy within Macroscopic-microscopic Model and Langevin Approach
    LIU Lile, CHEN Yongjing, WU Xizhen, LI Zhuxia, GE Zhigang, SHEN Caiwan, SU Yang, HUANG Xiaolong, SHU Nengchuan
    2022, 56(5):  879-887.  DOI: 10.7538/yzk.2022.youxian.0127
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    The threedimensional Langevin model was used to study the fragment mass distribution and total kinetic energy (TKE) distribution and the scission configuration in lowenergy nuclear fission, in which the potential energy surface was calculated by using the macroscopicmicroscopic model based on the twocenter shell model and the finite range liquid drop model, and the inertia tensor and the friction tensor were obtained within the WernerWheeler method and the wallandwindow model, respectively. Taking the case of 14 MeV n+235U fission as an example, the influence of the elongation deformation space on the fission fragment mass distribution and the TKE distribution was investigated, and the correlation between the elongation and the mass asymmetry at the scission point was also studied. It is found that the elongation deformation space has a significant influence on the symmetric fission channel, especially for the calculation of the TKE around the symmetric mass region, which is because that the symmetric channel corresponds to the super-long deformation for the light actinide nuclei such as U, Np, Pu and so on, so that the less elongation deformation space could block the larger elongated nuclear shapes along the Langevin trajectories. In addition, the influence of the shell damping parameter on the fission fragment mass distribution and TKE distribution and the scission configuration was studied. It is found that the shell damping parameter has a larger influence on the fragment mass distribution, and the ratio of the peak height and the valley increases with the larger shell damping parameter, due to the stronger shell effect. However, it has little influence on the TKE distribution which indicates that the strength of the shell effect has little influence on the nuclear elongation. In the last, the preneutron and postneutron fragment mass distributions for the 14 MeV n+233,235U fission were calculated, and the results agree well with the evaluated data from ENDF/BⅧ0, which shows that the present model has the power of calculating the fission fragment mass distribution quantitatively.
    Macro-microscopic Calculation of Five-dimensional Potential Energy Surface for 234U Fission
    ZHU Xin, WANG Zhiming, LUO Changkai, NI Lei, ZHONG Chunlai, FAN Tieshuan
    2022, 56(5):  888-895.  DOI: 10.7538/yzk.2022.youxian.0250
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    A key question about the fission process of a heavy nucleus is to probe the nuclear potential energy landscape and its evolution from the single groundstate compound nucleus over the top of the fission barrier and further to the scission point, finally terminating in the formation of fission fragments. In this work, a macromicroscopic model was established to calculate the fission properties for uranium elements. The total potential energy as a function of deformation parameters was divided into a smoothly varying macroscopic part and a microscopic part representing quantum effect correction. The nuclear shape was described by the SwiateckiNix threequadraticsurface (3QS) parametrization, which could accurately represent the nuclear shape evolution all the way from the ground state to the scission configuration. The macroscopic part of the nuclear potential energy was calculated by the LSD (LublinStrasbourg drop) model. In the microscopic part, the foldedYukawa potential was used as the independent particle potential, whereas the shell correction method of Strutinsky and the smooth BCS pairing model were used to calculate the deformation energy of the compound nucleus. For the 234U compound nucleus, a potential energy surface (PES) with 5 906 250 lattice points was calculated and analyzed in a fivedimensional deformation space given by the 3QS parametrization. The watershed algorithm was used to search the fission paths on the fivedimensional PES for 234U. For different nuclear shapes, there are two well separated fission paths, asymmetric and symmetric, which share the same inner barrier and deviate at the point of second minimum, and finally end at two different points of the PES. During reaching scission points, the asymmetric fission pass will cross a new barrier with lower height than the outer barrier, but the height of the new barrier needs to be crossed is higher than the outer barrier for the symmetric fission path. The heights of fission potential barrier and the nuclear shapes at special positions, such as the saddle point and the scission point, were given for 234U. The comparison of our results, the inner barrier height and outer barrier height, with available experimental data and other’s theoretical results confirms the reliability of our calculations. The calculated results of this work, especially the outer barrier in the large deformation area, are in good agreement with the experimental data of RIPL library and Moller’s calculation, which means that the 3QS parametrization might be closer to the real nuclear shape than the generalized Lawrence shape description specially in the large deformation region.
    Theoretical Calculation for Photonuclear Reaction of 142-146,148,150Nd
    JIN Yongli, TIAN Yuan, TAO Xi, WANG Jimin, XU Ruirui, LIU Ping, GE Zhigang
    2022, 56(5):  896-904.  DOI: 10.7538/yzk.2022.youxian.0213
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    Photonuclear reaction data can describe the physical process of the interaction between photons and nuclei as well as the emission of particles. It has important applications in nuclear reactor physics, accelerators, radiation shielding, activation analysis, nuclear waste transmutation, and nuclear synthesis of celestial bodies. The natural stable isotopes of neodymium (Nd), 142Nd, 143Nd, 144Nd, 145Nd, 146Nd, 148Nd and 150Nd, are important fission products which are necessary in the activation analysis and reactor physics. In order to provide more complete and better quality data to the nuclear application, the Chinese theoretical photonuclear reaction program MENDG is developed by China Nuclear Data Center and Nankai University, which contains the optical model, HauserFeshbach model and Exciton models, and can deal with the complicated particle emission reactions below the photon induced energy 200 MeV. Base on MENDG, the nuclear data of photon nuclear reactions of 142146,148,150Nd in the energy range below 200 MeV were systematically calculated in this work. It should be noticed before the theoretical calculations that, various existing experimental data in EXFOR for photonuclear reaction including (γ, xn), (γ, sn), (γ, in) (i=1,4) and so on were collected and evaluated, some discrepancies and selfinconsistencies among different data were removed, and the present theoretical calculations were guided under the well evaluated measurements. Photon absorption cross sections for the photon incident energy below 200 MeV were firstly determined with the classical photon strength functions and the quasideuteron model. Eight formula with Lorentz functions including the standard Lorentzian (SLO) model, the enhanced generalized Lorentzian (EGLO) model, the Hybrid  (GH) model, the generalized Fermiliquid (GFL) model, and the three kinds of modified Lorentzian (MLO1, MLO2, MLO3) models were all utilized to describe the giant dipole resonance below about 40 MeV, and SLO model was validated as the suitable one in this case with the minimum Chi square for experimental data and calculations. Then, the theoretical calculations of photonuclear reaction of 142146,148,150Nd were carried out by MENDG. The competing reactions especially for the multiple neutron emissions were globally obtained by the six optimized GilbertCameron level density parameters, the theoretical selfconsistency for 142146,148,150Nd were satisfied as much as possible. The partial photoneutron cross sections were specially discussed systematically, and satisfying theoretical descriptions were observed in the global case. Moreover, to improve the theoretical results, the local parameters were also adjusted for each isotope and better fitness with the experimental data can be observed comparing to the global ones. As a result, all the theoretical data are good in physics and are included in the photonuclear data sublibrary of CENDL32.
    Reaction Cross Section and Covariance Evaluation of 56Fe(n,p)56Mn below 35 MeV
    LI Xiaojun, LAN Changlin, ZHANG Yue, YANG XianlinZHANG Zhi, SUN Xiaodong
    2022, 56(5):  905-917.  DOI: 10.7538/yzk.2021.youxian.0838
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    Neutron induced reaction cross sections play an important role in nuclear science and technology research, such as national defense, nuclear energy construction and development, nuclear medicine, radiation protection and nuclear safety. With the rapid development of nuclear technology, there are increasing demand for the variety and accuracy of neutron data. As an important structure material nuclide, the accuracy of neutron induced reaction cross section on 56Fe is directly related to the design and operation of reactors, spent fuel disposal and miniaturization design, and also of great significance to nuclear physics basic research. The 56Fe(n,p)56Mn reaction is usually used as the standard to monitor the neutron field flux in the experiment. It greatly influences the results of nuclear data measurement, and also has guiding significance for upon carry the determination of the parameters in the nuclear reaction model. Due to the obvious differences existing in the previous experimental data and inconsistency between the evaluation databases, the experimental measurement data evaluation and covariance analysis were carried out for the 56Fe(n,p)56Mn reaction cross section in this paper. Firstly, the existing experimental measurement data of 56Fe(n,p)56Mn reaction cross section in EXFOR and literature were downloaded and systematically collected, then summarized, contrasted, and classified in terms of neutron source, measurement method, detector kinds, etc. Secondly, the data with low reliability were discarded with checking the original papers, and the uncertainty mentioned in these papers was also analyzed. Additionally, previous experimental results which using the imprecise values were adjusted and modified with the more accurate data such as recently ratios, supervised reaction standard cross sections, and isotopic abundances. The data with significant divergence near 14 MeV were normalized, which made it more concentrated and less divergent. The polynomial curve fitting was applied to obtain the excitation curve of 56Fe(n,p)56Mn reaction between 29535 MeV which was derived from the corrected and normalized results. Following that, the correlation covariance matrix was calculated for the suggested experimental data after evaluation. Meanwhile, the energy density, optical potential parameters, energy level density parameters, and other parameters of 56Fe(n,p)56Mn reaction were adjusted within the physical model calculated by the nuclear reaction simulation program TALYS. Finally, the reaction cross section of 56Fe(n,p)56Mn was computed theoretically, compared with experimental data and evaluation databases. This paper broadened the existing evaluation method of neutron activation reaction cross section experimental data, and also improved the accuracy of the assessment data for neutroninduced 56Fe(n,p)56Mn reaction cross section measured data below 35 MeV. The recommended assessment value and its covariance of the 56Fe(n,p)56Mn reaction excitation function were given based on the evaluted results of experimental data. Excitation function input file for 56Fe(n,p)56Mn reaction derived by adjusting the TALYS parameters agree well with the experimental and evaluation data. A set of standard assessment techniques were constructed, which serves as a guide for theoretical calculation and database construction of related nuclear data.
    Optimization Method Applied to Theoretical Calculation of Fission Nuclear Reaction
    TIAN Yuan, XU Ruirui, ZHANG Yue, TAO Xi, WANG Jimin, JIN Yongli, ZHANG Zhi, SUN Xiaodong, GE Zhigang
    2022, 56(5):  918-926.  DOI: 10.7538/yzk.2021.youxian.0923
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    Nuclear data are fundamental data for nuclear energy, nuclear engineering, nuclear technology applications and nuclear science research, and have always received national attention. In particular, the importance of neutroninduced fission nuclear reaction data is unquestionable and the theoretical evaluation of nuclear data is required for a long period of time. The reaction mechanism of neutron induced fission reaction is very complicated and needs to be described by several different nuclear reaction models, such as optical potential model and compound nuclei model and so on. Nowadays, the most famous international nuclear reaction program systems are TALYS and EMPIRE. In China, the program FUNF is a fission nuclear reaction program system with our own independent intellectual property rights, which was developed by Prof. Zhang Jingshang of the China Nuclear Data Center. Each of these program systems need more than 40 parameters to be adjusted. Therefore, it is necessary to optimize a large number of parameters at the same time in order to correctly describe the nuclear fission reaction data. In this work, the FUNF program system was used to calculate the fission reaction data of neutron induced 238U. The fast neutron reaction data of fissile materials with incident energy from about 1 keV up to 20 MeV were calculated by FUNF code. It consists of the spherical optical model, the HauserFeshbach model, and the angular momentum dependent exciton model to describe the emission from compound nucleus to the discrete levels of the residual nuclei in preequilibrium process, meanwhile equilibrium reaction process was described by the HauserFeshbach model with width fluctuation correction. Based on the FUNF nuclear fission reaction model, the parameters of the FUNF program system was adjusted by the advanced optimization method MINUIT. MINUIT is a numerical minimization software library in the Fortron by CERN staff Fred James in the 1970 s. Then, combined with the parallel calculation method of MPI (message passing interface), preliminary calculations for the neutron induced fission reaction of 238U were performed. The results show that the FUNF parameters obtained by the MINUIT optimization method can well describe the fission reaction data of 238U. Next, the procedure will continue to be optimized to further improve the efficiency of the tuning parameters and the accuracy of the parameters. And the application to the relevant fission nuclear reactions will be developed to improve the accuracy of our fission nuclear reaction data.
    Fission Yield Evaluation Method Based upon Zp Model and Uncertainty Analysis in Burnup Calculation
    SHU Nengchuan, ZHU Shuyu, ZU Tiejun, JIN Yongli, LU Zerun, CAO Chaowei, SU Yang, LIU Lile, CHEN Yongjing, GE Zhigang, LIU Ping, HUANG Xiaolong, XU Ruirui
    2022, 56(5):  927-936.  DOI: 10.7538/yzk.2022.youxian.0188
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    Nuclear fission product yield (FPY) is an important data base for the research of nuclear energy and nuclear devices. The development of reliable and efficient yield evaluation methods is of great significance for the establishment of highquality yield database. Previously, to obtain a FPY for a specific product, the related experimental data would be collected and analyzed to obtain the weighted average value as the recommended value. This method did not consider the physical relation among the yields of the precursor and daughter nuclides on the same mass chain. So a physical model is needed to describe the isobaric (with same mass) Z distribution, and could be used to evaluate the independent and cumulative yields simultaneously. Zp model is a better method to describe the isobaric charge distribution since its physic mechanic is more certain. And to connect with the cumulative yields, the conversion matrix is deduced from the decay data, with which the cumulative yields could be calculated from independent yields on the same chain. Based upon the above idea, a fitting program named ZpFit was established for the unified evaluation of isobaric yields, and was applied to evaluate the yields of n+235U fission. The related independent yield, cumulative yield and corresponding covariance data were obtained selfconsistently and formed a yield database in ENDF format. For examples, the evaluations for yields of A=125, 140 and 149 of nth+235U fission were expressed in this paper. For A=125, the cumulative yields in literature have discrepancy, roughly could be grouped with higher and lower parts with the help of ZpFit codes. An experiment was carried out to clear this discrepancy by the ream in CIAE, the yield for 125Sn is around 0025, about 20% and 10% lower than the data in ENDF /BⅧ0 and JEFF33, respectively, agrees with the lower group of the literature values. So finally, we got the fitted yields giving more weighting to the lower part. For A=140, the result shows that with the help of ZpFit, we could decide which yields are not reasonable and should be excluded. The uncertainties of nuclear density of kinf and some important nuclides were calculated for TMI1 cell in UAM burnup benchmark. Compared with those based upon ENDF/BⅧ0, the results are roughly in agreement. The burnup relative uncertainty transferred from independent fission yields of 149Nd vs other all product’s from nth+235U fission of this work were also calculated, the density uncertainties are mostly contributed from the uncertainty of 149Nd for nuclides 149Sm, 150Sm, 151Sm, 152Sm, and  151Eu, these nuclides have important impact for the reactor reactivity. In summary, ZpFit based upon Zp model was developed to evaluate the independent and cumulative yields on the same mass chain. The burnup uncertainties of kinf and some important FP were studied based upon the yield database of present work and other libraries.
    Study of Energy Dependence of Neutron-induced Fission Yield of 235U with Bayesian Machine Learning
    QIAO Chunyuan, PEI Junchen, WANG Ziao, CHEN Yongjing
    2022, 56(5):  937-943.  DOI: 10.7538/yzk.2022.youxian.0198
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    Nuclear fission fragment yields are the key infrastructure data in the field of nuclear engineering and nuclear applications. However, it is very difficult to obtain accurate and complete energydependent fission yields by experiments and theories. To supply the application needs, the twodimensional cumulative fission yields of neutroninduced fission of 235U are evaluated for energy dependencies and uncertainty qualifications by crossexperiment data fusion. The data fusion is aim to include more data correlations to produce more consistent and useful information. In this work, the Bayesian machine learning with a doublelayer neural network was adopted, which was particularly suitable for dealing with imperfect data. The conventional evaluation methods were not ideal for uncertainty quantifications. Furthermore, the experimental uncertainties of fission yields were taken into account in this work, which was essential for data fusion. This is reasonable that the yields with larger uncertainties would have smaller weights in the data fusion. Previously, the Bayesian evaluation of one dimensional mass yields in terms of Y(A) or Y(Z) was studied. As a further step, this work evaluated the two dimensional yields in terms of Y(N, Z) or Y(A, Z), which are of practical usefulness for developing novel nuclear reactors. The doublelayer networks with 18×18, 20×20 and 22×22 neutrons were tested and the network structure of 20×20 was chosen. The yieldenergy relations of some key fragments such as 99Mo, 97Zr, 127Sb, 131I, 140Ba, 143Ce and 147Nd were obtained. The full twodimensional cumulative fission yields at neutron incident energies of 2, 6, 8, 10, and 14 MeV were obtained. The resulted twodimensional fission yields can reasonably describe the energy dependencies of evolution of fission modes. The resulted uncertainties are dependent on specific fragments and incident energies. The evaluated uncertainties includes a background noise about 135, which is still very large. In the future, it is essential to develop physics-informed machine learning to obtain more reliable evaluations. It is promising that Bayesian machine learning can facilitate the maximum utilization of imperfect raw experience data.
    Development and Validation of Burnup Library for PWR
    WU Xiaofei, WU Haicheng, LIU Ping, GE Zhigang, WEN Lili, XIAO Yue
    2022, 56(5):  944-951.  DOI: 10.7538/yzk.2022.youxian.0212
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    Burnup calculation is an important field in the nuclear physics calculations by which the changes in the isotopic composition of materials is computed. Burnup calculation is of significant importance for a wide range of applications in various stages of design, licensing, operation, waste management and decommissioning. The role of the nuclear data library is very important to the accuracy of burnup calculations. CINDER90 is a widely used point nuclear depletion code and its library was released in the year 2000. The raw data which are used for processing CINDER90 library are primarily from ENDF/BⅤ、ENDF/BⅥ and EAF3. It is necessary to update this library with newly released evaluated nuclear data. In this paper, a CINDER90 burnup library for PWR was developed based on nuclear data library ENDF/BⅧ0 and EAF2010, which consists of three parts: neutron cross section, decay data and fission product yields. The processing of neutron cross section was divided into two parts, firstly the branching ratios of EAF-2010 were integrated into ENDF/B-Ⅷ.0 library using the inverted-stack algorithm and CRECTJ6 program, and then the pointwise cross section was processed into 63group infinitely diluted cross sections by the opensource nuclear data processing code NJOY2016. The crosssection data were Doppler broadened at 2936 K and the midlife PWR flux spectrum with a fusion peak added (IWT=5) was applied as weighting function. Decay data and fissionproduct yields data were processed based on the MF8/MT457 and MF8/MT454 separately. Continuum delayed photon emission data was converted into a series of discrete lines by defining a set of energy bins with a width of 10 keV that spans the range of the continuum. Fissionproduct yields data contain 60 datasets from 36 fissionable actinides. When yield data are not available for an isotope identified in the library as fissionable, 239Pu fast yield set is given for neutron induced fission and 252Cf spontaneous fission yield set is given for spontaneous fission. The Takahama-3 benchmark from SFCOMPO-2.0 database was used to validate the new library. For most of the nuclides, the inventory calculation results obtained by using the new library and the old library are in well agreement except for 242Amm, of which the calculated value using the new library is closer to the experimental value. After investigating the 63group neutron cross sections, we believe that the inelastic cross sections in the old library are unreasonable. It can be concluded that the processing method proposed in this paper is correct and the calculation accuracy is improved for some important nuclides.
    Research on Benchmarking Method of Burnup Database
    XIAO Yue, WU Haicheng, WU Xiaofei, ZHANG Huanyu
    2022, 56(5):  952-960.  DOI: 10.7538/yzk.2021.youxian.0872
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    The development of high precision burnup analysis requires high accuracy burnup database. The benchmarking method of burnup database is very important for developing high accuracy burnup database. Based on the burnup database composed of the multitemperature continuous point section database CENACE13 made by China Nuclear Data Center and Cinder90 burnup database, the modeling of SF95 samples in the post irradiation experiment of TAKAHAMA3 power water reactor was taken as an example. The spent fuel composition database SFCOMPO20 gives the detailed information of postirradiation experiment of TAKAHAMA3 power water reactor, including component information, power history, boron concentration history and nuclides inventory value. The influence of modeling elements on burnup calculation was studied, and the modeling method of burnup benchmark experiment was determined. The influence of modeling elements such as material temperature and density, boron concentration and burnup step on modeling calculation was determined. As an important part of burnup credit, burnup analysis aims to obtain the nuclide composition and inventory of spent fuel. Using the burnup benchmark experimental modeling method, the burnup credit was studied and the comparison of calculated values and experimental values of interested actinide nuclides and fission products nuclides was carried out. The preliminary results showed that the deviation between the calculated values and the experimental values of main actinide nuclides is less than 2%, the relative deviation between the calculated values and the experimental values of most minor actinides nuclides is less than 10%, and the deviation between the calculated values and the experimental values of most important fission products nuclides is less than 5%. In this paper, based on the adjacent burnup chain of 125Sb, the variation law of 125Sb inventory with burnup depth was theoretically analyzed, and it is confirmed that there are defects in the measurement results of destructive radiochemical experiment. In this study, the destructive experimental value of 125Sb was modified by nondestructive experimental measurement value and experimental value modification formula. And the corrected experimental values of 125Sb inventory were obtained, which reduced the calculation deviation between calculated value and experimental value from nearly 170% to less than 20%. This study shows that the benchmarking of burnup database using the spent fuel composition database requires not only the appropriate burnup benchmark experimental modeling method, but also the appropriate evaluation and verification of inventory measurement data, so as to be able to be used for benchmarking of burnup database and burnup credit research.
    Macroscopic Validation of Nuclear Data in Leadbased Reactor Based on VENUS-Ⅱ Facility
    JIANG Wei, ZHANG Lu, YU Rui, GU Long
    2022, 56(5):  961-968.  DOI: 10.7538/yzk.2022.youxian.0173
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    Lead-based fast reactor (LFR) is one of the advanced reactor concepts elected in the generation Ⅳ forum of international nuclear energy. Lead-based materials, including lead and leadbismuth, are used as coolant in the reactor core. The LFR core has the strong abilities of nuclear fuel proliferation and spent fuel transmutation, which is due to the fact that the neutron spectrum is very hard. The reactor core with the high inherent safety runs at atmospheric pressure, which can reduce the probability of coolant accident loss accident. Leadbased fast reactor is of vital importance to promote the sustainable development of nuclear energy in China. At present, the large uncertainties in the nuclear cross section data still exist in the neutronics simulations in the leadbased reactor. VENUSⅡ leadbased zeropower reactor, in which the neutron flux spectrum is close to that in the lead-based reactor, can be used to carry out the macroscopic validations of nuclear data of the materials in the lead-based reactor. In this paper, the reactor reactivity of VENUS-Ⅱ lead-based zero-power reactor in the supercritical state was measured by the period method, and the effective multiproliferation factor keff was obtained as 1001 14±0000 07. Meanwhile, the leadbased reactor model was accurately built by the MCNP code, and keff of the supcritical reactor in the experiment was calculated by MCNP with four worldwide cross section libraries (ENDF/BⅦ.0, ENDF/BⅦ.1, CENDL-3.1 and JENDL-4.0). The results show that the keff values calculated by the four cross section libraries are in good agreement with the experimentally measured one, and the maximum relative deviation is less than 1%. And the calculated result with ENDF/BⅦ.1 shows a better agreement with the experiment one and the relative and absolute deviations are respectively 025% and 251 pcm. For the interlibrary comparison, a single ENDF/B-Ⅶ.1 element was substituted with other libraries and the small changes in the keff value were calculated by MCNP code. It is found that the lead element causes the largest change in the keff, and the lead nuclear data in CENDL-3.1 and JENDL-4.0 respectively causes the keff change value of 219 pcm and 166 pcm. By comparing the fission rates in the fuel rods, it is concluded that keff values positively correlated with fission rates in the fuel rods. The fission rate results of ENDF/BⅦ.1 and other libraries show the largest difference in the fifth ring fuel rods of the reactor. The neutron spectrum curves calculated by the four libraries are almost consistent, and the shape of the neutron spectra is mainly determined by the nuclear fuel material and the matrix material. In this work, the nuclear data of some materials in the leadbased reactor have been primarily validated, which can provide the important reference for the subsequent neutronics experiments on VENUSⅡ leadbased reactor.
    Research on Generation Method for Photon-related Data in Nuclear Data Processing Code NECP-Atlas
    ZU Tiejun, XU Ning, YIN Wen, CAO Liangzhi, WU Hongchun
    2022, 56(5):  969-977.  DOI: 10.7538/yzk.2022.youxian.0126
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    The nuclear data processing code NECP-Atlas, developed at Xi’an Jiaotong University, has various functions to process the evaluated nuclear data to generate the applied nuclear data library for the nuclear reactor designs. In this paper, the photonrelated data generation capacities were developed in NECPAtlas, including the photon production cross sections and photon irradiation cross sections. In the conventional nuclear data processing code such as NJOY, only the production cross section for the prompt photon released with the neutroninduced reactions, and the photonatom reaction cross section can be produced, while the delayed photon released by the fission products is ignored regardless of its large contribution for the heating of the regions without fissionable materials such as control and reflector assemblies in fast neuron reactors. In this paper, the capacity for the generation of the production cross section for the delayed photon was developed in NECPAtlas. Besides, in the conventional nuclear data processing code, only the neutron irradiation damage cross section can be generated, while in the practical reactors, electron and positron also cause significant irradiation damage. The photonatom reactions, including Compton scattering, pair production and photoelectric effect, emit electron and positron. Therefore, the photon can also cause irradiation damage. To satisfy the demands from various application aspects, the capacity for the calculation of the photon irradiation damage cross sections was also developed in NECPAtlas. Meanwhile, the electron and positron irradiation damage cross section can also be produced, because it is the basis for the calculation of photon irradiation damage cross section. The delayed photon production cross sections were tested on the EBRⅡ fast reactor, and the results were compared with the library of the ERANOS which includes the production cross section of delayed photon. The difference of photon heating calculated based on the two libraries respectively from NECPAtlas and ERANOS is less than 0.93%, which means the correctness of the delayed photon data produced by NECPAtlas. Besides, the numerical results show that if the delayed photon is ignored, the difference of photon heating for the control assembly and reflector assembly can reach 3258% and 2041%, compared with the results of considering the delayed photon. The electron and positron displacement damage cross sections of Fe and Au calculated by NECPAtlas were verified with the results published by Oka Ridge National Laboratory, and the photon displacement cross sections calculated by NECPAtlas were verified with results reported by other researches. The numerical results show good agreement with the reference results.
    Development and Validation of Broad-group Shielding Library Based on CENDL-3.2
    SHU Wenyu, CAO Liangzhi
    2022, 56(5):  978-987.  DOI: 10.7538/yzk.2021.youxian.1060
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    The neutron and photon shielding is an important part of the reactor design for safety of people and facilities. CENDL-3.2 is the latest Chinese evaluated nuclear data library released by China Nuclear Data Center (CNDC) in 2020. Many nuclides were improved with new evaluation techniques and experimental data. Thanks to the high calculation efficiency, broadgroup shielding libraries are widely used in shielding calculation. To carry on the research on reactor shielding calculation, the codes NECPAtlas and NECPShield were applied to develop the broadgroup shielding library NECLCP29 based on CENDL32. Generating a broadgroup shielding library needs three stages. Firstly, NECLCP199, a finegroup library with 199 neutron groups and 42 photon groups, was produced using NECPAtlas and NECPShield. When producing the finegroup library, the neutron weight function consists of a 1/E spectrum, a fission spectrum and a thermal Maxwellian spectrum. The photon weight function consists of a 1/E spectrum at the middle energies. At the high and low energy, the spectrum goes down. Secondly, a typical 1D PWR model used in producing BUGLE libraries was calculated using NECLCP199 to obtain the finegroup flux distribution in the model. Before calculating the model, the nuclides were selfshielded using several different models, including a PWR pin cell model for the core, an ironwater model for the down comer, a concrete model for the concrete, a carbon steel model for the pressure vessel and a stainlesssteel model for the baffle and barrel. Lastly, the finegroup fluxes in core, down comer, pressured vessel and concrete were chosen to collapse finegroup cross sections of selfshielded nuclides to generate a broadgroup library. Moreover, the finegroup flux in concrete was applied to collapse the finegroup infinite diluted cross sections of all nuclides from NECLCP199. The energy group structure has an important influence on the accuracy and efficiency for shielding calculation. To obtain an improved broadgroup structure for better accuracy and efficiency, particle swarm optimization (PSO) was used. The research on application of PSO in optimizing the energy group structure for shielding calculation was performed, including the combination of PSO process and the generation of the broadgroup library, the mapping of PSO variables and the construction of fitness function. A 29group structure was optimized using PSO. NECLCP29 with the optimized 29group structure was generated. To validate NECLCP29 broadgroup shielding library, some benchmarks in SINBAD, the international shielding benchmark library, were calculated, including Iron88, ASPISNG, and HBR2. The calculated results were compared with the measured results. Moreover, the results of BUGLEB7 and BUGLE96 which are famous international broadgroup shielding libraries were also used for comparison. The numerical results show that the calculated values using NECLCP29 agree well with the measured values. Compared with BUGLEB7 and BUGLE96 whose numbers of neutron groups are all 47, the calculation accuracy of NECL-CP29 is better. Especially in calculating the reaction rate of 197Au(n,γ) in Iron88 benchmark, the results given by NECLCP29 are much closer to the measured values than BUGLEB7. Thanks to the optimized energy group structure with less neutron energy group number, the computational efficiency is obviously improved.