Validation of the Monte Carlo model supporting core conversion of the Portuguese Research Reactor (RPI) for neutron fluence rate determinations

The Monte Carlo code MCNP was used to calculate absolute values of thermal, epithermal and fast neutron fluence rates in the new RPI core using fresh LEU fuel. Discrepancies smaller than 20% were obtained between calculated results and activation foil measurements. A previous knowledge of general characteristics of the neutron energy spectra, provided by the MCNP reactor model itself, has been fundamental to determine the conditions yielding a proper comparison of simulated and measured results. An excellent agreement (6%) was also obtained for the relative neutron fluence rate profiles along the fuel height. The MCNP model of the reactor core was therefore validated for a tri-dimensional determination of neutron fluence rates in the fuel assemblies and neighbouring irradiation positions.     

堆内超临界水回路辐照装置物理参数模拟研究

堆内超临界水回路对我国超临界水堆燃料和结构材料的辐照腐蚀实验具有重要意义,辐照装置位于反应堆堆芯栅格,是超临界水回路的核心部件。采用MCNP程序模拟研究辐照装置的关键物理参数,并考虑超临界水热物特性对物理参数的反馈效应。计算得到辐照装置热中子注量率为4.72×10¹³ cm^-2•s^-1,快中子注量率为1.55×10¹⁴ cm^-2•s^-1,辐照产热率为14.7 kW,反应性引入为0.045%。     

The definition of damage functions from irradiation test data

This report describes a method of determining irradiation effectiveness of different neutron spectra in causing radiation effects to fuel cladding and reactor structural materials. It involves the definition of a semi-empirical damage function or cross section using measured data from specimens irradiated in thermal and fast test reactor spectra. The damage function is applied to design problems involving irradiation effects to reactor structural and fuel cladding materials to predict the fluence which would produce a specific change in material properties.     

超高通量快中子研究堆需求分析

实现超高快中子通量是世界先进研究堆的重要发展方向,对于加快第四代先进核能系统燃料及材料创新发展具有重要意义。本文从先进核能堆内结构材料与核燃料的辐照考验、长反应链超钚元素生产等角度,初步分析了我国建设超高通量快中子研究堆的必要性。在此基础上,确定了超高通量快中子研究堆的堆芯最大中子注量率及其冷却剂,给出了反应堆主要参数及冷却剂流动方案。反应堆热功率为200 MW,冷却剂为铅铋合金,最大中子注量率大于1016 cm?2·s?1。      

中子吸收材料累积中子注量研究

中子注量可作为加速辐照实验的辐照指标。为了通过加速辐照的方式检验中子吸收材料的中子吸收性能,计算了中子吸收材料贮存不同时间下的中子注量。通过对乏燃料组件初始富集度、燃耗深度以及乏池温度、可溶硼浓度的研究,得到中子吸收材料在乏池贮存时中子注量的包络值,同时计算得到不同贮存时间材料¹⁰B的消耗量。结果表明,材料的中子吸收性能在贮存10～60 a的情况下并无明显变化。本文结果可为检验材料的中子吸收性能提供支持。     

Calculation of neutron fluence in the region of the pressure vessel for the history of different reactors by using the Monte-Carlo-method

Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture.     

Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor

This study aims to estimate burnup of the fuel elements for the Istanbul Technical University TRIGA Mark II Research and Training Reactor using a Monte Carlo-based burnup-depletion code. Effect of burnup on the core neutronic parameters, effective core multiplication factor, fast/epithermal/thermal neutron fluxes, and core-average neutron spectrum, and incoming neutron spectrum of the piercing beam port (PBP), is investigated at the Beginning of Life (BOL) and End of Life (EOL). Operational data peculiar to a selected operation sequence, which contains positions of CRs, power level of the reactor, material temperatures and latest core map, are used to determine the current fuel burnup of fuel elements at the time under consideration. A specific operation sequence is selected for the analysis. Furthermore, all control rods are considered fully withdrawn to assess the excess reactivity. Results are obtained using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library for a full power of 250 kW. Neutron cross-section libraries at the full-power operating temperatures are generated using NJOY. From the results, the calculated burnup values of the core at the sequence considered and EOL are found to be 420 MWh and 560 MWh, respectively. Remaining excess reactivity is calculated to be less than 0.3 $. It is observed that core average thermal neutron flux reduces by 1 % while the fast and epithermal neutron fluxes remain almost unchanged.     

超高通量快中子研究堆核燃料概念设计研究

提高中子注量率是高通量研究堆的发展趋势,能够大幅加速反应堆材料研发进程。但若提高中子注量率至1016 cm?2·s?1将导致功率密度峰值相较于现有研究堆高数倍,对反应堆和核燃料设计带来许多挑战。为此,本文从中子学、传热、燃料材料堆内行为等方面半定量分析了提高中子注量率对核燃料性能的影响,并提出应对超高通量和功率密度挑战的设计措施,为发展超高通量快中子研究堆燃料设计提供指导。      

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

In the design of fast reactor core with higher burnup and higher linear power, prediction accuracy of burnup history of fuel pin should be upgraded so as to assure fuel integrity without extra design margin under increased neutron fluence and burnup. A method is studied to predict fuel pin-wise power and its burnup history in fast reactors accurately based on an analytic solution of diffusion theory equation on hexagonal geometry with boundary condition from core calculation by finite-differenced diffusion calculation code. The present method is applied to a fast reactor core model, and its accuracy in predicting fuel pin power is tested. The result is compared with the reference solution by the finite difference calculation with very fine mesh. It is found that the present method predicts the power peaking factors in fuel assemblies accurately. The fuel pin-wise nuclide depletion calculation is also done using neutron fluxes for each fuel pin. The result shows that the fuel pin-wise depletion calculation is very important in predicting the burnup history of the fuel assembly in detail.     

Gas-cooled fast breeder reactor fuel bundle irradiations in the BR2 helium loop

In the BR2 helium loop at Mol, Belgium, a 12-pin test fuel element of gas-cooled fast breeder reactor (GCFR) design and materials will be irradiated at a 500 W/cm maximum pin rating and a 700°C maximum cladding temperature to a target burnup of 60 MWd/kg (extension to 100 MWd/kg is intended). The design of the test element and the loop is described in detail. To fabricate the test element, parts of the GCFR fuel development had to be anticipated. Preliminary out-of-pile testing was successfully performed, and irradiation is scheduled to start in early 1977 and will be completed between mid-1978 and mid-1979, depending on the final burnup objective. GCFR operating conditions will be completely simulated except for the full size of the fuel element and the fast neutron flux. In combination with out-of-pile performance testing of full-size dummy elements and fast flux experience from the liquid metal fast breeder reactor program, the helium loop irradiation is regarded as an adequate basis for the design of a fuel element for a GCFR demonstration plant serving as the final test bed.     

Monte Carlo simulation of the Greek Research Reactor neutron irradiation facilities

A Monte Carlo simulation of the Greek Research Reactor was carried out using MCNP-4C2 code and continuous energy cross-section data from ENDF/B-VI library. A detailed model of the reactor core was employed including standard and control fuel assemblies, reflectors and irradiation devices. The model predicted neutron flux distributions within the core in good agreement with calculations performed using the deterministic code CITATION and measurements using activation foils. The model is used for the prediction of the neutron field characteristics at the reactor irradiation devices and enables the design and evaluation of experiments involving material irradiations.     

中国实验快堆结构材料辐照装置设计

为在中国实验快堆(CEFR)上开展国产快堆包壳材料的辐照试验,进行了CEFR首个结构材料辐照装置的设计。材料辐照装置的创新设计基于CEFR的辐照条件和堆芯组件的基本结构,通过在辐照装置内部设置不同气隙尺寸的辐照罐,实现了在快堆不同功率稳态运行条件下(40%和100%额定功率)对材料样品不同辐照温度(450~600℃)的要求。辐照装置具有样品辐照温度与中子注量率的非在线监测功能,其结构具有通用性,能满足材料辐照标准试样最大装载的需要。通过对辐照装置进行热工分析和堆外的传热验证试验、流阻特性和结构稳定性验证试验,保证了辐照装置的设计能满足材料辐照任务的要求。     

Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1.The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are also presented.     

反应堆压力容器中子辐照损伤的分析和评价

对于材料已经确定的反应堆压力容器,其辐照脆化效应的主要因素是快中子积分通量。本文应用中子输运格林函数法验算了秦山核电站压力容器1/4厚度处最大快中子通量。分析和评价结果表明,该压力容器的设计对中子辐照是安全的。     

Four energy group neutron flux distributions in the Syrian Miniature Neutron Source Reactor using the WIMSD4 and CITATION codes

A 3-D (R, θ, Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the pointwise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation sites with relative differences less than 7% and 5%, respectively.     

Development of Thermal/Irradiation Stress Analytical Code “VIENUS” for HTTR Graphite Block

A new thermal/irradiation stress analysis code “VIENUS” has been developed for the graphite block in the High-Temperature Engineering Test Reactor (HTTR). The VIENUS is a two- dimensional finite element visco-elastic analysis code to take account of graphite behavior under irradiation in detail. In the analysis, the effects of both fast neutron fluence and temperature on material properties are considered.

The code has been evaluated by the irradiation test results of the Peach Bottom fuel elements to confirm the thermal/irradiation stresses in the graphite block. It is clarified that the calculated results are able to estimate a tendency of the test results, and that both the irradiation- induced creep and dimensional change are the most important parameters in the thermal/irradiation stress analysis. From the present study, it is suggested that the VIENUS code is a useful tool to evaluate the thermal/irradiation stresses in the HTTR graphite blocks.     

Performance of TRISO particles fueled with plutonium and minor actinides in a PBMR-400 core design

The TRISO particle design of high temperature reactors fueled with plutonium (Pu) and/or minor actinides (MAs) is investigated by calculating the failure fraction of TRISO particles during irradiation. For this purpose, a fuel depletion, neutronics and thermal-hydraulics code system, which delivers the fuel temperature, fast neutron flux and power density profiles, is coupled to an analytical stress analysis code. The latter is being further developed for the calculation of a reliable and realistic failure fraction. The code system has been applied to a PBMR-400 design containing TRISO particles fueled with 1st and 2nd generation plutonium and with a target burn-up of 700 and 600 MWd/kgHM, respectively. It is shown that the pebble-bed type high temperature reactor under consideration is a promising option for burning Pu and MAs if very high burn-ups can be achieved. The TRISO particle failure fraction is also calculated for both Pu and MA fuels, and compared to U-based fuel. It is shown by the present stress analysis code that the Pu-based fuel particles need a better design and this has been achieved for the MA-based fuel, in which helium gas atoms have a significant contribution to the buffer pressure.     

基于蒙特卡罗-离散纵标双向耦合方法的堆坑粒子注量率计算

为有效解决大型复杂核设施屏蔽计算问题,研究了三维蒙特卡罗(MC)-离散纵标（S_N）双向耦合方法,通过自主开发接口程序实现MC粒子概率分布与S_N角通量密度之间的相互转换,实现MC-S_N双向耦合计算。将基于MC-S_N双向耦合方法的程序用于某反应堆堆坑底部粒子注量率计算。利用MC程序建立堆芯及堆坑处的精细模型进行计算,三维S_N程序用于堆芯下表面与压力容器底面之间区域的计算。通过MC-S_N-MC两步耦合计算,给出堆坑通道及小室内的中子和光子注量率。三维MC-S_N双向耦合方法计算结果与单一MCNP程序结果吻合较好,初步验证了该方法是解决大型复杂核装置屏蔽问题的有效工具。     

Assessment of Radiation Damage to the Structural Material of EAST Tokamak

Radiation damage to structural material of fusion facilities is of high concern for safety. The superconducting tokamak EAST will conduct D-D plasma experiments with the neutron production of 1015 neutrons per second. To evaluate the material radiation damage a programme system has been devised with the Monte Carlo transport code MCNP-4C, the inventory code FISPACT99, a specific interface, and the fusion evaluated nuclear data library FENDL-2. The key nuclear responses, i.e. fast neutron flux, displacement per atom, and the helium and hydrogen production, are calculated for the structural material SS-316L of the first wall, and the vacuum vessel, using this programme. The results demonstrate that the radiation damage to the structural material is so little that it will not lead to any significant change of material properties according to the reference design. This indicates that there is a large potential space for EAST to test advanced operation regime from the viewpoint of structural material safety.     

Optimization of flux distribution in a molten-salt reactor with a 2-region core for plutonium burning

Molten-salt reactors (MSRs) are selected as one of the candidates of Generation IV reactor concepts. In GLOBAL2005 held in Tsukuba, Japan, one paper discussed the flattening of fast neutron flux in the core for a longer life of graphite moderator. In the paper a 3-region reactor concept was presented. The authors tried many cores changing configurations such as volume of each region and fractions of fuel salt in the regions or fuel compositions.

We investigated the other possibility of a 2-region core for the simplicity. Using one energy group neutron diffusion theory and considering extrapolation distance, the optimum selection of region wise neutron multiplication factors can be theoretically and easily obtained. In MSRs, there is no burnup distribution of the fuel. The region wise neutron multiplications can be obtained by adjusting the volume fraction of fuel in a cell with a given composition of the fuel salt. Using the theoretical results, the optimization of the actual core configuration was determined by a nuclear analysis code SRAC2002 with the nuclear data library of JENDL3.3.

In this paper, we considered MSRs using plutonium as a fissile material. Ordinary MSRs use uranium-233, which doesn't exist naturally, and utilizing plutonium is easier to startup.