Analysis of integral experiment on erbia-loaded thermal spectrum cores using Kyoto University Critical Assembly by MCNP code with various cross section libraries

Under the project on high burnup nuclear fuel development using erbium as a burnable poison, a series of experiments were performed at the Kyoto University Critical Assembly. The experimental results have formed the basis for this study which aims to analyze the suitability of various evaluated nuclear data libraries for using them in neutronic calculations under the project. The MCNP code was used for the analysis. Calculation model geometry was fully detailed, and ENDF, JENDL, JEFF, and TENDL libraries were used during calculation. For the cross sections of erbium nuclides, the analysis revealed that calculated results upon all the libraries corresponded with experimental data within the errors. However, in some libraries, significant differences were found in case of carbon and uranium nuclides under certain conditions.     

MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination

The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10^?4. Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition.     

MCNP在临界计算中的影响因素分析

简单说明计算所用的实验资料,在此基础上,介绍了采用MCNP计算MOX燃料实验堆有效增殖系数的过程。说明以不同的随机数序列重复计算问题,得到的计算结果总存在一定的涨落,并随着总迭代次数的增加逐渐减小;紧接着在统计涨落足够小的基础上,讨论了数据库的差异对计算结果的影响;对于含中子热化问题,指出采用S(α,β)热散射处理是得到合适的有效增殖系数计算值的必须条件。     

Trend in criticality calculation results of borated-water-moderation experimental cores relating to 10B(n, α) cross section

ABSTRACT

In connection with the accuracy of the ¹⁰B(n, α) cross section in the thermal- and epithermal-neutron energy regions, criticality calculation results were examined for six benchmark sets of light-water-moderation critical experiments of UO₂ and MOX fuel lattice cores with un-borated and borated water. Two of the benchmark sets were those implemented in the Tank-Type Critical Assembly (TCA). The others were taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP), and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). The enrichments of the UO₂ fuel range from 1.9 wt% to 2.6 wt%, and the Pu contents of the MOX fuel do from 2.0 to 6.6 wt%. The boron concentrations in water are up to 1511 ppm. The effective neutron multiplication factors (k_eff ) were taken from the published documents. They were calculated with continuous-energy Monte Carlo calculation codes in combination with JENDL-4.0, and other evaluated nuclear data libraries. It was confirmed that the k_eff values of the critical cores increased with the boron concentrations, which indicates that the ¹⁰B(n, α) cross section in the thermal- and epithermal-neutron energy regions should be larger than those in JENDL-4.0 and other libraries.     

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

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

Performance of the CENDL-3.2 and other major neutron data libraries for criticality calculations

Nuclear data are the cornerstones of reactor physics and shielding calculations.Recently,China released CENDL-3.2 in 2020,and the US released ENDF/B-Ⅷ.0 in 2018.Therefore,it is necessary to comprehensively evaluate the criticality computing performance of these newly released evaluated nuclear libraries.In this study,we used the NJOY2016 code to generate ACE format libraries based on the latest neutron data libraries（including CENDL-3.2,JEFF3.3,ENDF/B-Ⅷ.0,and JENDL4.0）.The MCNP code was used to ...     

Impact of cross section libraries on CASMO-4E and CASMO-4 results

A detailed comparison between the code CASMO-4 and its extended version CASMO-4E has been made. In addition to the standard library, CASMO-4E calculations have been performed also with its extended libraries. The differences are significant enough to be considered when choosing the library to be used for a particular problem. The differences in the multiplication factor k_∞ range up to several hundred pcm depending on the void history, burnup and other parameters. The differences in fuel temperature or void coefficients are smaller especially at small void fraction and low burnup. At large void and low burnup CASMO-4E with the standard library gives significantly different results than the other combinations. The microscopic cross sections show small differences when calculated with the same library but clear differences due to the extended libraries.     

用于剂量计算的核反应总截面经验公式

在核物理及核物理应用研究中，利用核反应总截面经验公式进行理论计算具有现实意义。这方面的计算公式很多，本文以中能区应用的Kox经验公式为基础，结合重离子治疗肿瘤计算剂量深度模型的需要，对Kox经验公式的参数进行了修正，使其具有更广泛的应用价值，同时对修正参数的物理意义进行了探讨。     

X光管滤光片效应的蒙特卡罗仿真

在X射线荧光分析中,使用滤光片可有效降低原级光谱本底对待测元素的干扰,提高分析灵敏度。本文采用MCNP5方法仿真X光管滤光片对原级谱线的影响,比较了滤光片对不同能量电子束产生的X射线谱的差异,分析了不同滤光片材料及厚度与X射线原级谱衰减的关系,得到不同滤光片对X射线谱的衰减规律。     

Status of the problem of nuclear cross section data for IBA

The progress achieved in an IAEA Coordinated Research Project (CRP) to improve the cross section data for IBA is reported. The objective of the CRP, started in 2005, is to create a nuclear cross section database for IBA that contains reliable and usable data freely available to the entire IBA community. The major results achieved so far by the CRP participants are discussed. The results include compilation and assessment of the existing cross sections, new experimental data and evaluation of the most wanted cross sections. The experimental results are incorporated into the IBANDL database (www-nds.iaea.org/ibandl/) and evaluated data are presented at the SigmaCalc cross section calculator (www-nds.iaea.org/sigmacalc/).     

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C,JENDL-3.3 and ENDF/B-Vl data libraries

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for ^natZr, ^natMo, ^natCr, ^natFe, ^natNi, ^natSi, and ^natMg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ²⁸, δ²⁵, ρ²⁵, and C¹ were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries.     

燃耗截面基本库对输运燃耗耦合计算的影响分析

基于蒙特卡罗中子输运程序和ORIGEN2点燃耗程序的蒙特卡罗输运燃耗耦合计算方法应用广泛。但现有评价库中子连续截面的核素个数远小于燃耗计算涉及到的核素数量,即通过输运计算得到的燃耗截面不足以完全替代燃耗计算的基本库。采用经过栅元验证的蒙特卡罗燃耗程序MCBMPI,对最新的VERA燃耗计算基准题进行验证计算,对比分析不同的燃耗截面基本库对输运燃耗计算的影响。分析结果表明:1)在实际应用中尽量不要采用典型热中子截面库,会带来较大偏差;2)在燃耗计算核素替换较多的情况下,对该基准题而言,选取典型压水堆基本库还是典型快堆基本库,对结果影响不大,二者keff偏差在8‰以内,燃耗末期235U偏差在4‰以内,135Xe偏差在5‰左右;3)建议选取与研究对象能谱相近的基本库。     

Development and testing of the code for automatic generating of multi-temperature continuous-energy neutron cross section libraries

In the nuclear reactor design, a code for automatically generated multi-temperature continuous-energy neu- tron cross section data library, which is called AMTND for short, was designed and developed to meet the need of the reactor core design coupled with thermal-hydraulic design. The code can provide a point-wise cross- section at any temperature for a Monte Carlo neutron transport program, such as MCNE In ensuring that the nuclear data produced by AMTND meets the testing of critical benchmark experiments, the time-consumed by the nuclear data generating of AMTND compared with NJOY＇s was carried out and the result shows the code＇s excellence. In order to test the accuracy of the code, out and the results verified the code preliminarily. the Doppler coefficient test benchmark was also carried     

Displacement damage cross section and mechanical properties calculation of an Es-Salam research reactor aluminum vessel

Nuclear facility aging is one of the biggest problems encountered in nuclear engineering. Radiation damage is among one of the aging causes. This kind of damage is an important factor of mechanical properties deterioration. The interest of this study is on the Es-Salam research reactor aluminum vessel aging due to neutron radiation. Monte Carlo(MC) simulations were performed by MCNP6 and SRIM codes to estimate the defects created by neutrons in the vessel. MC simulations by MCNP6 have been performed to determine the distribution of neutron fluence and primary knock-on atom(PKA) creation. Considering our boundary conditions of the calculations, the helium and hydrogen gas production in the model at a normalized total neutron flux of 6.62×10~(12) n/cm~2 s were determined to be 2.86 × 10~8 and 1.33 × 10~9 atoms/cm~3 s,respectively. The SRIM code was used for the simulation of defects creation(vacancies, voids) in the aluminum alloy of the Es-Salam vessel(EsAl) by helium and hydrogen with an approximate energy of 11 MeV each.The coupling between the two codes is based upon postprocessing of the particle track(PTRAC) output file generated by the MCNP6. A small program based on the Mat Lab language is performed to condition the output file MCNP6 in the format of a SRIM input file. The concentration of silicon was determined for the vessel by the calculation of the total rate of ~(27)Al(n,γ)~(28)Si reaction. The DPA(displacement per atom) was calculated in SRIM according to R.E. Stoller recommendations; the calculated value is 0.02 at a fast neutron fluence 1.89 × 10~(19) n/cm~2.RCC-MRx standard for 6061-T6 aluminum was used for the simulation of the evolution of mechanical properties for high fluence. The calculated values of nuclear parameters and DPA obtained were in agreement with the experimental results from the Oak Ridge High Flux Isotope Reactor(HFIR) reported by Farrell and coworkers.     

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

ABSTRACT

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

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

Determination of the neutron skin thickness from interaction cross section and charge-changing cross section for B,C, N,O, F isotopes

The effective neutron and proton root-meansquare radius of stable and unstable nuclei(~(12-15,17)B,~(12-20)C,~(14-21)N,~(16-24)O and ~(18-21,23-26)F) were deduced from the charge-changing cross section,σ_(cc),and the interaction cross sections,σ_I,by using a statistical abrasion-ablation model calculation.The extracted proton radii are in good agreement with the data from the Atomic Data and Nuclear Data Tables within the errors.Furthermore,we can observe that the neutron skin thickness increases monotonously with the increasing neutron number in these isotopes,which is consistent with the systematical trend of theoretical calculations.     

Inactivation cross section of yeast cells irradiated by heavy ions

1IntroductionThecalculationsofinactivationcrosssectionsforheavyionsinthetrackwidthregaldisplayingthindownforE.ColtB/randBs-1,andforBaciUusSubtilusarestraightforwardforthese1-hitdetectorsbasedontheKatztracktheory.[1]CalculationsforV-79hamstercellsaremorecomplex.Theyfollowtheoriginaldevelopmentofitusmodelforeucaryoticcens,andmalleuseofthecrosssectionscalculatedforhypotheticalinternaltargetswhicharethenassertedtobeproportionaltothemeasuredcellularinactivationcrosssections.Inmaxnmallancensinac…