We report the development and validation of a new neutron detector, the spherical moderator counter (SMC), which represents the next generation of long counter technology. Unlike conventional cylindrical designs, the SMC employs a spherical moderator geometry, optimized by Monte Carlo transport simulations to achieve a near isotropic angular response and a broad, flat energy sensitivity. The prototype detector has been extensively tested with a wide range of neutron energies, i.e., neutrons coming from d-D and d-T accelerators, Cf-252 spontaneous fission, Am-Be (alpha and n) reactions, fission reactors, and spallation sources, covering the entire energy range from thermal to several tens of MeV. The experimental results demonstrate that SMC achieves a uniform response over 4 pi steradians, with an angular deviation of less than 6% for fast neutrons, and maintains adequate sensitivity from 0.01 eV to 20 MeV. These findings confirm that SMC not only addresses the inherent limitations of traditional cylindrical long counters but also establishes a versatile and reliable platform for neutron metrology, with significant potential for applications in next-generation fusion and advanced nuclear technologies.

This paper presents the development and validation of China's first benchmark measurement system for neutron leakage time-of-flight (TOF) spectra using a Cf-252 spontaneous fission source and spherical polyethylene sample. EJ-309 and CLYC scintillation detectors were used for neutron detection, and a shadow cone was employed for background suppression. Notably, the SiC detector was, for the first time on this platform, applied as the start-time signal generator in TOF spectrum measurement. The TOF spectrum covering the energy range of 0.15-8.00 MeV was measured, and the results were systematically compared with evaluated data from four major nuclear libraries: ENDF/B-VIII.1, JEFF-3.3, JENDL-5, and CENDL-3.2. The comparison revealed strong agreement across the full spectrum, with calculated to experimental (C/E) deviations remaining within 5% in the high-energy region and within 13% at low energies. These results verify the system's stability and suitability for integral experiments. The established benchmark platform provides a strong technical foundation for future neutron nuclear data validation, particularly in shielding applications and the improvement of fission-spectrum nuclear databases.

An interesting isotope, sulfur-33 (a stable isotope), has been proposed as an additional neutron-capture agent and a cooperative target for improving boron neutron capture therapy. The 33S (n, α)30Si reaction emits the α particles useful for neutron capture therapy. In this study, the dosimetric characteristics of sulfur-33 neutron capture therapy (SNCT) were investigated at the cellular level using the Monte Carlo method. The Monte Carlo N-Particle (MCNP) code was employed to simulate the transport of α particles emitted from different source regions of a single-cell model, including the cell surface (CS), cytoplasm (Cy), and nucleus (N). The model cells were defined as two concentric spheres consisting of the nucleus and cytoplasm. The radius of the cell model was Rc = 6 μm Rn = 3 μm. The S-values (SC←C, SC←CS, SN←N, SN←Cy, SN←CS) were compared to the results with the Medical Internal Radiation Dosimetry (MIRD) method. Finally, S-values of the cellular model of several geometries were calculated.The relative differences of MCNP versus MIRD method S-value for SNCT ranged from 0.28% to 4.29% for C to C (SC←C), -2.82% to 0.05% for Cs to C (SC←Cs), 0.54% to 4.42% for N to N (SN←N), -0.72% to 3.14% for Cy to N (SN ←Cy), and -1.76% to 0.36% for CS to N (SN←CS). The ratios of SN←N/SC←C, SN←CS/SC←C, SN←Cy/SC←C decreased with increasing the cell and nucleus size. The S-value simulated by MCNP is in agreement with the MIRD method, which is proved to be reliable for the cellular dosimetry of SNCT by the Monte Carlo method. The Monte Carlo method can be used instead of the MIRD method to accurately calculate cellular S-value, which solves the problem of cell shape limitation in MIRD method (which is suitable for regular spherical cellular model).

Critical benchmark experiments constitute the cornerstone for validating and refining nuclear data libraries. Recent advancements, however, have revealed fundamental incompatibilities within multiple beryllium-reflected critical configurations-a paradox that challenges conventional nuclear data validation paradigms. This study pioneers a diagnostic framework integrating sensitivity analysis with uncertainty quantification to establish a similarity metric for Be-related benchmark systems. A paradigmatic analysis of the PMF021 benchmark reveals a critical contradiction: while nuclear data modifications could theoretically induce maximum ke f f variations of 278.7 pcm, the observed 1135 pcm discrepancy between configurations exceeds theoretical adjustment limits by a factor of 4.1. This breakthrough necessitates two fundamental needs in nuclear data practices: (1) Introduction of a quantitative reliability index for benchmarking conflicting configurations, (2) Mandatory re-evaluation protocols for existing Be-reflected benchmarks. Our findings establish a new parameter, providing an essential discriminator for identifying inconsistent benchmarks that risk misleading the nuclear data community.

With the advancement of burning plasma operation goals, diagnostic systems in nuclear fusion reactors face a critical challenge: their key components will be continuously exposed to high-flux neutron and gamma radiation over long-term operation. The diagnostic Port Plug (PP), serving as the platform for installing diagnostic systems and other auxiliary devices, constitutes the first line of defense for the diagnostic systems against the nuclear environment. Based on the current Equatorial J Port Plug (JPP) design, this study employed Monte Carlo methods to perform a comprehensive neutronics analysis of the PP and its key components, covering neutron and gamma fluxes, nuclear heating, and displacement per atom (DPA).The results indicate that the flux distribution is mainly influenced by component layout and diagnostic openings. Nuclear heating analysis shows that gammainduced heating dominates in high-Z optical mirror assemblies, while the maximum DPA reaches 6.43 & times; 10-3, suggesting that plasma-facing key components may be subject to potential radiation-induced damage. This study provides a quantitative assessment of the neutron and gamma environment, nuclear heating, and radiation damage for the equatorial J-port diagnostic systems, offering fundamental data to guide the design optimization and reliability assurance of future diagnostic systems.

The present study provides new measurements of the 78Kr(n,2n)77Kr reaction cross sections at neutron energies of 13.89 and 14.57 MeV, along with a comprehensive evaluation of the excitation function up to 15 MeV based on the present and literature data. Additionally, theoretical calculations using the TALYS 2.2 and EMPIRE 3.2.3 codes were performed up to 20 MeV to provide broader model constraints. Enriched 78Kr, prepared as a high-density solid, served as the target material. Quasi-monoenergetic deuterium-tritium (D-T) neutrons were employed for activation analysis, followed by off-line high-resolution gamma-ray spectrometry. The measured cross sections are 159.7 ± 5.9 mb at 13.89 ± 0.15 MeV and 268.2 ± 10.1 mb at 14.57 ± 0.15 MeV, with total uncertainties below 3.76%. The two present data sets and four experimental data sets from the literature were systematically examined and recalibrated using the most recent nuclear parameters. The recommended excitation function from the reaction threshold to 15 MeV was determined by combining a third-order polynomial and an SPCC spline fit. Furthermore, a correlation covariance matrix for this reaction in the 14-MeV region was constructed for the first time. Comparisons with the TALYS- and EMPIRE-based calculations provide valuable constraints for model optimization and further development of nuclear data libraries.

Abstract The data on fission by neutrons is widely applied in reactor burnup, neutron fuel verification, and the identification of nuclear fuels. Due to the abundance of 238U, which is the most common isotope of uranium, cumulative data of 238U fission yield in the D-D neutron energy range is significant in fission research in Generation-IV (Gen-IV) reactors. In this article, measurements of the fission yield of the 238U(n,f) reaction at a neutron energy of 2.9 ± 0.3 MeV have been carried out by the activation technique based on off-line γ-ray spectrometry. The neutron irradiation experiment was conducted in the CPNG-600 neutron generator installed in the China Institute of Atomic Energy (CIAE). A quasi-monoenergetic beam of neutrons was made using the D(d,n)3He reaction. The γ-rays of the activation products were characteristically measured with the help of a low-background HPGe spectrometry system, and fluctuations of neutron flux were measured with the aid of an Au-Si surface barrier detector. Once the correction factor needed was applied, highly accurate cumulative yields of seven fission products were derived. The obtained measurements were related to the existing experimental values and data on estimating yield provided by the ENDF/B-VIII.0 library. Fission yields of a 238U(n,f) reaction were also determined with the help of the TALYS-1.96 code. The current outcomes have dependable information to confirm the behavior of energy-dependent fission yield and a supplementary nuclear reaction database as applied in reactor design and operation.

The cross-sections of K-39(n, p)Ar-39 at an energy of 2-3 MeV play an important role in nuclear structure research and Ar-40/Ar-39 geochronology application. Due to the limitations of n-He-3 coincidence technology and counting instruments, the data in literature are from before 1967, and existing data are scarce and significantly diverge. Meanwhile, there are large discrepancies between the measured and evaluation results. By taking advantage of the high sensitivity and resolution of the noble gas mass spectrometer at the Institute of Geology and Geophysics, Chinese Academy of Sciences (IGGCAS), the cross-sections of K-39(n, p)Ar-39 were measured by combining neutron activity analysis and noble gas mass spectrometry, and the uncertainties are discussed in detail. The cross-sections of K-39(n, p)Ar-39 were measured as 103.84 +/- 16.33, 109.76 +/- 15.88, and 150.27 +/- 24.19 mb at 2.56 +/- 0.08, 2.69 +/- 0.08, and 2.96 +/- 0.12 MeV energies, respectively. The measured data filled the data gaps and provided more accurate data support for Ar-40/Ar-39 dating. Furthermore, the theoretical excitation function of K-39(n, p)Ar-39 was calculated using TALYS-1.97 computer codes. Then, the experimentally determined cross-sections were analyzed by comparing them with the data from the EXFOR database and evaluated nuclear data in ENDF/B-VIII.0, JEFF-3.2, TENDL-2021, BROND-3.1, and JENDL-5 databases. According to the comparative results, the measured cross-sections of K-39(n, p)Ar-39 exhibit a rapid energy-dependent increase between 2-3 MeV, aligning with higher literature values and resolving previous discrepancies. Compared with the previously reported data, the precision of the determined cross-sections in this study showed considerable improvement. The comparison of measured data indicates that the combined detection method of neutron activity analysis and noble gas mass spectrometry techniques is suitable for measuring the cross-sections of nuclear reactions with long-lived product nuclei and the application of Ar-40/Ar-39 geochronology with a D-D neutron source.

To achieve high-accuracy in-situ calibration of the neutron flux monitor on the Experimental Advanced Superconducting Tokamak (EAST), a systematic comparison of continuous and multi-point calibration approaches is conducted using a full three-dimensional (3D) engineering model. 3D neutron transport simulations for 2.45 MeV neutrons in the complex structural environment of EAST are carried out using MCNP and PHITS. Calibration calculations are conducted for three different source configurations: a toroidal ring line source, 32 coplanar point sources, and a D-shaped volume source representing realistic plasma conditions. Neutron flux responses and energy spectrum at the spherical long counter (SLC) location are quantitatively analyzed. The results show that the differences in neutron flux between the ring source and point source calibration approaches remain within 2.2%. An operational calibration strategy for an upcoming in-situ experiment on EAST using a 1 × 109 n/s D-D accelerator neutron source is proposed. This work establishes quantitative criteria for relating volumetric and point-source models and provides a transferable framework for neutron calibration in EAST.

As one of the six critical nuclides in nuclear engineering applications, 239Pu is widely utilized in fields such as fast breeder reactors and accelerator-driven subcritical systems. Accurate neutron-induced fission cross-section data are critical for nuclear facility design, operation, and fuel cycle management. However, discrepancies become apparent in the fast neutron energy range regarding the 239Pu(n,f) cross-section, where the five major evaluated nuclear data libraries not only differ in their evaluated values but also in their maximum reported energies. Notably, the CENDL-3.2 library, released in 2020, contains neither an updated evaluation for the 239Pu(n,f) crosssection nor the associated covariance data. This work has systematically compiled the experimental measurements of the 239Pu(n,f) cross-section from the EXFOR database. An evaluation and covariance calculation for the 239Pu(n,f) cross-section experimental data have been performed. The excitation function curve has been fitted through using the Origin and SPCC codes. The correlation covariance matrix has been calculated using the ASEU2.0 code. Additionally, the 239Pu(n,f) reaction cross-section has been comparatively analyzed and validated using the theoretical model codes TALYS-1.96 and FUNF (version 2012). Ultimately, an evaluated curve and covariance for the 239Pu(n,f) cross-section in the 0.2-200 MeV energy region have been provided, with a relative average uncertainty of 1.12%. The study extends the existing methodology for evaluating experimental data on cross-sections, resulting in a high-precision evaluated excitation function for 239Pu(n,f) across the 0.2-200 MeV energy range, complete with corresponding covariance data. This work also provides critical reference information for supplementing the 239Pu(n,f) cross-section covariance data in the CENDL.

Investigation of the cross-section for the 78Kr(n,2n)77Kr reaction is essential for studies of nuclear reaction mechanisms and fusion plasma diagnostics. Currently, experimental data are limited and exhibit significant discrepancies. This study provides new measurements within this energy range and a comprehensive evaluation up to 20 MeV. Enriched 78Kr, prepared as a high-density solid, served as the target material. Quasi-monoenergetic deuterium–tritium (D–T) neutrons were employed for neutron activation analysis, followed by offline high-resolution gamma-ray spectrometry. The measured cross-sections are 159.7 ± 5.9 mb at 13.89 ± 0.15 MeV and 268.2 ± 10.1 mb at 14.57 ± 0.15 MeV, with total uncertainties below 3.76%. All six available experimental data sets were systematically examined and recalibrated using the most recent nuclear parameters. The recommended excitation function, spanning from the reaction threshold to 15 MeV, was determined by combining a third-order polynomial and an SPCC spline fit. Additionally, a correlation covariance matrix for this reaction in the 14 MeV region was constructed for the first time. Comparisons with TALYS-2.2 and EMPIRE-3.2.3 calculations provide valuable constraints for model optimization and further development of nuclear data libraries.

To provide complementary benchmark data for validating evaluated neutron nuclear data with time-of-flight techniques, a neutron leakage experiment for spherical natural-iron samples was conducted on a 252Cf spontaneous-fission-source-based integral experiment platform established at Lanzhou University. Neutron leakage spectra from spherical natural-iron samples with diameters of 12, 24, and 36 cm were measured over the energy range of 0.15–8.00 MeV. The 0.15–0.80 MeV region was measured using a CLYC detector, while the 0.80–8.00 MeV region was measured using an EJ-309 liquid scintillation detector.Monte Carlo simulations were performed using the MCNP code with four evaluated nuclear data libraries: ENDF/B-VIII.1, CENDL-3.2, JEFF-3.3, and JENDL-5. The calculated leakage spectra were compared with the experimental results. The calculated-to-experimental (C/E) ratios indicate that JENDL-5 and ENDF/B-VIII.1 show comparatively good agreement with the measurements, whereas CENDL-3.2 and JEFF-3.3 exhibit larger deviations, particularly in the 0.80–8.00 MeV region.The NDPlot-based diagnostic analysis of 252Cf-spectrum-weighted secondary-neutron emission spectra suggests that these deviations may be related to differences in the treatment and representation of elastic and discrete-level inelastic scattering processes in the evaluated nuclear data libraries and the numerical reconstruction procedure. However, since the measured leakage spectra represent integral responses involving multiple scattering and coupled reaction channels, further sensitivity and uncertainty analyses are required to identify the dominant contributors quantitatively.This work provides complementary modern TOF-based benchmark data for natural-iron shielding calculations using a 252Cf-driven integral experiment platform. The measured leakage spectra provide additional experimental information for validating recently updated evaluated nuclear data libraries and may support future sensitivity, uncertainty quantification, and nuclear-data improvement for shielding applications in advanced nuclear energy systems.

A high-performance neutron and gamma-ray collimator was developed to address the challenges of high-precision multi-particle diagnostics, such as deuterium-deuterium (D-D) neutron, deuterium-tritium (D-T) neutron, and gamma-ray, in the extreme radiation environment of fusion devices. The design integrates geometric shielding, material attenuation, and energy-selection optimization. Full-scale three-dimensional Monte Carlo radiation transport simulations were conducted to evaluate the collimator performance in a complex structural environment on the EAST device. After collimation, the peak contributions of 2.5 MeV and 14 MeV neutrons reach 57.47 % and 21.31 % of the total energy spectrum, respectively, demonstrating effective energy discrimination. Predicted neutron spectra under d-T operation in EAST show that the collimator significantly suppresses environmental scattering and enhances directional sensitivity. The results indicate strong potential for dual-mode neutron-gamma-ray diagnostics in future d-T fusion reactors and high applicability to ongoing d-D experiments such as EAST. The collimator has been fabricated and installed on EAST and will be deployed in the upcoming d-T experimental campaigns.

与传统剂量率放射治疗（CONV-RT）相比，超高剂量率放射治疗（FLASH RT）能够减少对正常组织的损害，同时有效杀灭癌细胞。为了满足临床需求并适应大多数医院的安装条件，有必要开发一种室温紧凑型FLASH直线加速器。由中国工程物理研究院应用电子学研究所和中久大光医疗科技有限公司共同打造的FLASH-RT平台能够产生超高剂量率的X射线。本研究旨在测量实验平台的剂量并验证我们的设计。将EBT4胶片安装在固体水中，以测量不同射野大小下的百分深度剂量（PDD）曲线和剂量分布轮廓。基于实验平台和初始束流参数，使用MCNP6（蒙特卡罗N粒子输运代码，第6版）建立了蒙特卡罗模型，以优化转换靶厚度并测量PDD和束流轮廓。同时，为了满足用户对不同剂量率的要求，并实现同一台机器的实验，通过蒙特卡罗建模分析了几种剂量率调整方法的可能性。结果表明，在 X 射线 FLASH-RT 实验平台上，源皮距（SSD）为 50 厘米时，最大平均剂量率超过 330 Gy/s，脉冲内剂量率约为 7500 Gy/s，远高于触发 FLASH 效应所需的条件（≥40 Gy/s）。所有 PDD 曲线和轮廓的蒙特卡罗模拟结果与实验数据之间的差异均小于 3%。此外，伽马分析（2 毫米/2%标准）表明通过率超过 95%，证实结果完全在临床可接受的公差范围内。对于转换靶材料厚度的优化，结合最大转换效率原则所获得的平均剂量比连续减速近似（CSDA）范围增加了约 12.7%。建议仅调整平均剂量率的方法来改变脉冲频率，而同时改变平均剂量率和瞬时剂量率的方法是添加相应厚度的屏蔽。在次级准直器前使用该材料。此剂量率范围能够满足在同一台机器上进行 CONV-RT 和 FLASH-RT 的要求，最大程度地减少实验生物变异性。

The China Spallation Neutron Source (CSNS) is a newly constructed large-scale facility located in Dongguan, which was completed and operational in 2018. It is generating neutrons by bombarding 1.6 GeV protons into a tungsten target for multidisciplinary research. A back-streaming neutron beamline (Back-n) at CSNS is built at the reverse direction opposite to the proton beam mainly for the nuclear data measurement. Back-n is characterized by its wide energy range (from 0.3 eV to 300 MeV), high flux (~107 n/cm2/s at 55 m) and good energy resolution (less than 1% below 1 MeV), standing as one of the state-of-the-art white neutron beams worldwide. Fission cross-sections of a series of nuclei, such as 235, 236, 238U, 232Th, 239Pu have been measured in wide energy ranges since 2018, and more isotopes (such as minor actinides) are planned to be measured in the near future. The CSNS Back-n facility and the progress in the fission cross-section measurements are reviewed in this paper. Then the prospects of the fission cross-section measurement at Back-n are discussed.

Benchmark experiments are indispensable for the development of neutron nuclear data evaluation libraries. Given the lack of domestic benchmarking of nuclear data in the fission energy region, this study developed a neutron leakage spectrum measurement system using a spherical sample based on the 252Cf spontaneous fission source. The EJ309 detector (for high-energy measurements) and CLYC detector (for low-energy measurements) were combined to measure the time-of-flight spectrum using the γ tagging method. To assess the performance of the system, the time-of-flight spectrum without a sample was measured first. The experimental spectra were consistent with those simulated using the Monte Carlo method and the standard 252Cf spectrum from ISO:8529-1. This demonstrates that the system can effectively measure the neutron events in the 0.15--8.0 MeV range. Then, a spherical polyethylene sample was used as the standard to verify the accuracy of the system for the benchmark experiment. The simulation results were obtained using the Monte Carlo method with evaluated data from the ENDF/B-VIII.0, CENDL-3.2, JEFF-3.3, and JENDL-5 libraries. The measured neutron leakage spectra were compared with the corresponding simulated results for the neutron spectrum shape and calculated C/E values. The results showed that the simulated spectra with different data libraries reproduced the experimental results well in the 0.15--8.0 MeV range. This study confirms that the leakage neutron spectrum measurement system based on the 252Cf source can perform benchmarking and provides a foundation for evaluating neutron nuclear data through benchmark experiments.

Geochemical logging is essential for identifying formation lithology, and its quantitative analysis of the composition of formation elements relies heavily on accurate response gamma spectra. Measuring the response spectra by experiment is complicated, but using the Monte Carlo method for simulations is a feasible approach. We have developed and verified a high-precision technique to establish a comprehensive and highly accurate database of response spectra for geochemical logging tools with isotope neutron sources.We established a large cylindrical formation geometry using the MCNP6.1 code, including a geochemical logging tool with a Phi 7.62 x 15 cm bismuth germanate detector and an Am-Be neutron source. A suitable neutron-induced gamma-ray nuclear data library was selected, and the method for obtaining the formation gamma-ray and capture response spectra was enhanced. The constructed database was then evaluated by analyzing the spectra of actual formations after ensuring that the simulated spectra were consistent with the measured response spectra and characteristic gamma rays.We applied the weighted least-squares method to predict the element yields, achieving a maximum mean absolute error compared to the standard yields of less than 0.05. Meanwhile, the reconstructed spectra showed good agreement with the measured spectra, indicating the high accuracy of the spectra database generated by this technology. This can provide assistance for geochemical logging to determine the composition of formation elements, providing an essential foundation for geochemical logging's application in oil exploration and uranium development.

Innovation in neutron detection has supported basic science, the development of large-scale scientific facilities, and the development of clean nuclear energy. However, due to the uncharged nature of neutrons, accurately measuring neutron flux and spectrum simultaneously over a wide range of energies has always been challenging. A new type of neutron detector is designed based on PHITS-334, which utilizes a hydrogen-3He gas combination, combined with the nuclear recoil method and nuclear reaction method, to simultaneously measure neutron flux and spectrum in a broad energy region. Based on the isotope neutron sources of Am-B, Am-Be, Cf-252, and C-252 f -D2O given by ISO8529-1, the DD fusion neutron source of Experimental Advanced Superconducting Tokamak (EAST), and the theoretical fusion DT neutron source of International Thermonuclear Experimental Reactor (ITER), the energy spectrum measurement capability of the detector is evaluated. It has been demonstrated that the new detector can measure neutron flux over the entire energy range. The low limit of the conventional energy spectrum measurement is extended from 0.1 to 0.001 MeV through neutron spectrum analysis. The proposed method provides a new reference for developing neutron detectors. It should be noted that the new neutron detector can simultaneously measure neutron flux and spectrum information in a broad energy area, which will be the basis for integrating and miniaturizing neutron detection systems, such as fusion reactors.

In order to improve the quality of neutron data for 238U in CENDL, considering the impact of new measurements, a brand-new evaluation of the complete set of neutron induced 238U reaction data up to 20 MeV has been performed. Important reactions, such as (n,tot), (n,gamma), (n,f), (n,2n) and (n,3n) reaction cross sections and average number of fission neutrons have been evaluated based on experimental data analysis. Also, using existing optical model potential parameters, new theoretical calculations based on Hauser-Feshbach and pre-equilibrium model have been carried out. Resonance parameters and prompt fission neutron spectrum from ENDF/B-VIII.0 have been adopted. Guided by integral benchmark, (n,inl), (n,gamma) and (n,f) reaction cross sections have undergone multiple adjustments, leading to a noticeable improvement in data quality, as indicated by final benchmark results.

Lead (Pb) plays a significant role in the nuclear industry and is extensively used in radiation shielding, radiation protection, neutron moderation, radiation measurements, and various other critical functions. Consequently, the measurement and evaluation of Pb nuclear data are highly regarded in nuclear scientific research, emphasizing its crucial role in the field. Using the time-of-flight (ToF) method, the neutron leakage spectra from three ^nat Pb samples were measured at 60^∘ and 120^∘ based on the neutronics integral experimental facility at the China Institute of Atomic Energy (CIAE). The ^nat Pb sample sizes were 30 cm × 30 cm × 5 cm, 30 cm × 30 cm × 10 cm, and 30 cm × 30 cm × 15 cm. Neutron sources were generated by the Cockcroft-Walton accelerator, producing approximately 14.5 MeV and 3.5 MeV neutrons through the T(d,n) ^4 He and D(d,n) ^3 He reactions, respectively. Leakage neutron spectra were also calculated by employing the Monte Carlo code of MCNP-4C, and the nuclear data of Pb isotopes from four libraries: CENDL - 3.2, JEFF - 3.3, JENDL - 5, and ENDF/B-VIII.0 were used individually. By comparing the simulation and experimental results, improvements and deficiencies in the evaluated nuclear data of the Pb isotopes were analyzed. Most of the calculated results were consistent with the experimental results; however, a few areas did not fit well. In the (n,el) energy range, the simulated results from CENDL - 3.2 were significantly overestimated; in the (n,inl)D and the (n,inl)C energy regions, the results from CENDL - 3.2 and ENDF/B-VIII.0 were significantly overestimated at 120^∘ , and the results from JENDL - 5 and JEFF - 3.3 are underestimated at 60^∘ in the (n,inl)D energy region. The calculated spectra were analyzed by comparing them with the experimental spectra in terms of the neutron spectrum shape and C/E values. The results indicate that the theoretical simulations, using different data libraries, overestimated or underestimated the measured values in certain energy ranges. Secondary neutron energies and angular distributions in the data files have been presented to explain these discrepancies.