Comparison of calculated and measured reaction rates obtained through foil activation in the subcritical dual spectrum facility YALINA-Booster

Reaction rates were measured by the foil activation technique to obtain neutron spectrum information in a subcritical core driven by an external neutron source. The experimental results are compared with Monte Carlo calculations in order to examine the capability of the Monte Carlo code MCNP together with ENDFB-6.8, JEFF-3.1.1 and CENDL-3.1 neutron cross section libraries to predict the neutron spectrum dependent reaction rates correctly in a subcritical core. The focus lies on fast neutrons. A discrepancy is found in the calculated-to-experimental values of the reaction rates and an inaccurate cross section is identified in CENDL-3.1.     

强脉冲源法测量次临界系统中子学时间常数的分析

讨论了在现有源强及测试条件下，用强脉冲源法测量次临界系统中子学时间常数本征值的可行性，并在此基础上对用γ强度谱诊断时间常数进行了研究。数值计算表明，这种新的测试方法能够得到很好的结果。     

Accelerator driven subcritical system as a future neutron source in Kyoto University Research Reactor Institute (KURRI) —Basic study on neutron multiplication in the accelerator driven subcritical reactor—

A basic study on the nuclear characteristics in the accelerator driven subcritical reactor (ADSR) was performed through a series of neutronics calculations in view of a future neutron source in Kyoto University Research Reactor Institute (KURRI) for the joint use program among researchers of Japanese universities. In this series of calculations, it was assumed that three kinds of monoenergetic neutrons were isotropically generated at the center of spherical and homogeneous cores with different moderator-to-fuel volume ratios in order to examine the spectrum mismatching effect between injected neutrons and fission neutrons born in the subcritical core. The results of calculations clearly showed the spectrum mismatching effect on the neutron multiplication in the ADSR.     

ADS实验大厅中子防护工程技术研究

为了保障加速器驱动次临界系统(ADS)散裂靶与反应堆耦合特性及影响验证实验的顺利进行,以原子能院现有的临界实验装置为基础,对堆厅部分墙体进行屏蔽改造。建造由聚乙烯、镉、铅、钢以及混凝土等材料构成的屏蔽装置,以防止临界装置产生的射线外泄,使工作人员受到的照射保持在合理水平。通过MCNP模拟计算,完成了屏蔽结构的优化设计。基于槽钢支撑结构、铅屏蔽层、镉屏蔽层和聚乙烯屏蔽层等材料组成的组合屏蔽结构建立简化模型,采用ANSYS有限元分析程序计算分析得出各部分应力小于许用应力,稳定性符合要求。最后通过工程实践,完成对屏蔽性能理论计算结果的验证。     

Power spectral analysis for a thermal subcritical reactor system driven by a pulsed 14 MeV neutron source

A series of power spectral analyses for a thermal subcritical reactor system driven by a pulsed 14 MeV neutron source was carried out at Kyoto University Critical Assembly (KUCA), to determine the prompt-neutron decay constant of the accelerator-driven system (ADS). The cross-power spectral density between time-sequence signal data of two neutron detectors was composed of a familiar continuous reactor noise component and many delta-function-like peaks at the integral multiple of pulse repetition frequency. The prompt-neutron decay constant inferred from the reactor noise component of the cross-power spectral density was consistent with that obtained by a pulsed neutron experiment. However, the reactor noise component of the auto-power spectral density of each detector was hidden by a white chamber noise in the higher-frequency range and this feature resulted in a considerable underestimation of the decay constant. For several runs with a low pulse-repetition frequency, furthermore, we attempted to infer the decay constant from point data of the delta-function-like peaks. The analysis for a run under a slightly subcritical state resulted in the consistent decay constant; however, those for other runs under significantly subcritical states underestimated the decay constant. Considering the contribution of a spatially higher mode to the point data, the above underestimation was solved to obtain the consistent decay constant. While the Feynman-α formula for a pulsed neutron source is too complicated to be fitted directly to variance-to-mean ratio data, the present analysis on frequency domain is much simpler and the conventional formula based on the first-order reactor transfer function is available for fitting to power spectral density data.     

On the use of different analytical solutions for recalculation of the YALINA-Booster experiment SC3A

The SC3A experiment in the YALINA-Booster facility in Belarus is described and investigated. For this investigation the very special configuration of YALINA-Booster core, consisting of a fast and a thermal zone, decoupled with a neutron ‘valve’ is analyzed in detail based on a full HELIOS model for the calculations. The two region design causes unexpected results in the experiments. The special problems for the analysis of the experiments are shown. The results for different analytical solution (one group diffusion, one group P₁ transport and two group diffusion) are analyzed and compared. To model the streaming of neutrons from the thermal area into the fast area, a special two group analytical solution for the space–time dependent neutron flux with two sources is developed from the available Green’s functions for two groups. The new analytical solutions show very good agreement in the comparison with the experimental results. Especially, with the two group and two source solution the unexpected behavior at the outermost detector can be reproduced. Thus analytical solutions without separation of space and time are a very promising tool to develop a new method for the analysis of ADS experiments.     

裂变-聚变中子源

为适应聚变堆的发展和处理高放废物的需要,提出裂变-聚变中子源的概念,它是采用LiD组件放在高通量反应堆中或中国先进研究堆(CARR)重水区中,通过慢中子与6Li(n,a)反应产生2.739 MeV氚离子,它与LiD中的D发生聚变反应,产生聚变中子;随着LiD中氚的快速积累,14 MeV 中子产生的D反冲粒子流与氚发生聚变反应,增长聚变中子产额,使 14 MeV 中子注量率逐渐升高.当氚浓度接近0.5×1022时,D反冲粒子流与氚的聚变反应率的产额接近于1,聚变中子将成倍的增长,类似于连锁反应,使聚变中子产额达到饱和,即t时刻产生氚,都被用于产生聚变反应,形成裂变-聚变中子源.这时的通量非常高,必须在接近饱和前对设定的通量(如3.5×1014n/cm2·s)下逐步降低反应堆功率,如降低CARR 中子注量率,使其在设定的通量下达到饱和,适应聚变堆中子注量率的需求.论述了裂变-聚变中子源的原理,聚变中子产生率,氚的积累速率和浓度,D反冲粒子流和与氚的聚变反应速率,以及其影响因素.在均匀中子场下(即不考虑中子降抑的情况下)计算了外径180 mm、内径100 mm的LiD管道中聚变中子注量率.     

On the calculation of flux in slightly subcritical reactors with external neutron sources

The behaviour of the flux inside a subcritical reactor in the presence of external neutron sources is examined. It is shown, in particular, that the flux can be approximated by the flux resulting from eigenvalue calculation as the reactor approaches its critical state. A method based on the perturbation technique is described, allowing an estimation of spatial effects on the flux by the local sources.     

Analytical solution for the Feynman-Alpha formula for ADS with pulsed neutron sources

The theory of Feynman-alpha measurements is elaborated for the case of a “stochastically pulsed” subcritical system. The corresponding physical situation is when a pulsed neutron source is used, and no synchronisation between the start of the measurement time gate and the pulsing is made. This is the case in the European Community supported research project MUSE.

The solution to the Feynman-alpha formula was obtained for such a case through complex function techniques in an analytical form by Laplace transform and residue calculus. The final expression is a smoothly regular function with a simple periodic modulation. It consists of a Feynman-curve corresponding to a stationary source, plus an infinite sum of periodic sine functions squared. The series converges as 1/n⁶ with the summation index n, thus in practice two or three terms are sufficient for a high accuracy quantitative result. This few-term representation amounts to a compact closed form analysis solution. Such a solution is well suitable for use in the determination of the subcritical reactivity from measurements, in contrast to the case of deterministic pulsing (measurement start synchronized with pulsing), where no simple solution is available, and where no explicit relationship between the continuous and pulsed forms of the Feynman-alpha exists.     

A source efficient ADS for minor actinides burning

Taking advantage of the good neutron economy of nitride fuel, a compact accelerator-driven system (ADS) for burning of minor actinide fuels has been designed, based on the fuel assembly geometry developed for the European Facility for Industrial Transmutation (EFIT) within the EUROTRANS project. The small core size of the new design permits reduction of the size of the spallation target region, which enhances proton source efficiency by about 80% compared to the reference oxide version of EFIT. Additionally, adoption of the austenitic steel 15/15Ti as clad material allows to safely reduce the fuel pin pitch, which leads to an increase of fuel volume fraction and therefore makes the neutron energy spectrum faster, consequently increasing minor actinides fission probabilities. Our calculations show that one can dramatically increase neutron source efficiency up to 0.95 without a significant loss of neutron source intensity, i.e. having high proton source efficiency. Consequently, the accelerator current required for operation of the ADS with a fission power of 201 MWth and a burn-up of 27 GW d/t per year (365 EFPD) is reduced by 67%.     

Neutron kinetics in subcritical cores with application to the source modulation method

The subject of this paper is an investigation of the performance of the so-called source modulation technique for the measurement of reactivity in subcritical, source-driven cores. Methods of measuring reactivity by a single detector, including the source modulation method, are based on the assumption of point kinetic behaviour of the core. Deviations from point kinetic behaviour will lead to an inaccurate estimation of the reactivity. Hence, first, the conditions of point kinetic behaviour in subcritical source-driven cores are revisited. In addition to the known conditions for such behaviour, which have an analogy to those in critical cores, some additional cases are found which only exist in subcritical cores. Then the performance of the source modulation technique is investigated. It is found that the error of the method, originally thought to be due exclusively to the deviation of the local detector signal from the amplitude factor of point kinetics, remains finite and non-zero even in the limit of exact point kinetic behaviour (e.g., with low frequencies or deep subcriticalities). This is demonstrated and explained by analytical formulae. Some remedies for this shortcoming of the method are also suggested and discussed.     

Application of FLUKA and OpenMC in coupled physics calculation of target and subcritical reactor for ADS

The study of accelerator-driven subcritical reactor systems(ADSs) has been an important research topic in the field of nuclear energy for years. The main code applied in ADS research is MCNPX, which was developed by Los Alamos National Laboratory. We studied the application of the open-source Monte Carlo codes FLUKA and OpenMC to a coupled ADS calculation. The FLUKA code was used to simulate the reaction of highenergy protons with the nucleus of the target material in the ADS, which produces spallation neutrons. Information on the spallation neutrons, such as their energy, position,direction, and weight, can be recorded by a user-defined routine called FLUSCW provided by FLUKA. Then, the information was stored in an external neutron source file in HDF5 format by using a conversion code, as required by the OpenMC calculation. Finally, the fixed-source calculation function of OpenMC was applied to simulate the transport of spallation neutrons and obtain the distribution of the neutron flux in the core region. In the coupled calculation, the high-energy cross-section library JENDL4.0/HE in ACE format produced by NJOY2016 was applied in the OpenMC transport simulation. The OECD–ADS benchmark problem was calculated, and the results were compared with those obtained using MCNPX. It was found that the flux calculations performed by FLUKA–OpenMC and MCNPX were in agreement, so the coupling calculation method for ADS is reasonable and feasible.     

Preparation and verification of mixed high-energy neutron crosssection library for ADS

An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the energy of the spallation neutrons can reach several hundred megaelectron volts. However, the upper neutron energy limit of nuclear cross-section databases, which are widely used in critical reactor physics calculations, is generally 20 MeV.This is not suitable for simulating the transport of highenergy spallation neutrons in the ADS. We combine the Japanese JENDL-4.0/HE high-energy evaluation database and the ADS-HE and ADS 2.0 libraries from the International Atomic Energy Agency and process all the data files for nuclides with energies greater than 20 MeV. We use the continuous pointwise cross-section program NJOY2016 to generate the ACE-formatted cross-section data library IMPC-ADS at multiple temperature points. Using the IMPC-ADS library, we calculate 10 critical benchmarks of the International Criticality Safety Benchmark Evaluation Project manual, the 14-MeV fixed-source problem of the Godiva sphere, and the neutron flux of the ADS subcritical core by MCNPX. To verify the correctness of the IMPCADS, the results were compared with those calculated using the ENDF/B-VII.0 library. The results showed thatthe IMPC-ADS is reliable in effective multiplication factor and neutron flux calculations, and it can be applied to physical analysis of the ADS subcritical reactor core.     

A cold neutron beam for the experiment in grenoble

The neutron beam for the experiment is produced at the HFR reactor of the ILL Institute in Grenoble. Cold neutrons are transported by means of a slightly bent neutron guide with a 6 × 12 cm² cross section 60 m long to the experimental area, where they are injected into a divergent guide specially designed in order to reduce the beam divergence. The measurements of the beam intensity and profile are compared to the Monte Carlo calculations.