Performance of the improved version of the Monte Carlo A3MCNP code in the design of cask shielding

Abstract

A³MCNP (automatic adjoint accelerated MCNP) is a revised version of the MCNP Monte Carlo code that automatically prepares variance reduction parameters for the CADIS (consistent adjoint driven importance sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and radiation particle transport biasing within the weight-window technique. The current version of A³MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A³MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A³MCNP provided only a point source configuration option for large-scale shielding problems, such as spent fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A³MCNP (referred to as A³MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A³MCNPV in solving the cask neutron and gamma-ray shielding problem.     

A comparison of dose rate calculations for a spent fuel storage cask by using MCNP and SAS4

This paper presents a detailed comparison of the surface dose rate calculations for the NAC-UMS spent fuel storage cask by using MCNP and SAS4 computer codes. Their accuracy and computation efficiencies are compared. For such a real world deep penetration and streaming problem, effective variance reduction techniques are indispensable for a Monte Carlo simulation to obtain results of small statistic errors within reasonable computing time. The TORT-coupled MCNP calculation based on the CADIS methodology has been used in this study. The main differences between MCNP and SAS4 calculations are the underlying cross-section libraries and the adjoint functions used for variance reduction in Monte Carlo simulations. The cross-section libraries and their formats should be the root cause for some significant discrepancies between the MCNP and SAS4 results. In addition, limited by the 1D adjoint biasing scheme, SAS4 is inefficient in calculating the dose rates near inlet/outlet apertures. Considering all the computer time spent and the statistical errors of results obtained, the overall computation efficiency by using the TORT-coupled MCNP is better than SAS4 in the shielding calculations of spent fuel storage casks. More specifically, although the SAS4 efficiency is better when the cask side calculation is the only concern, the TORT-coupled MCNP technique is more efficient for the gamma-ray transport in cask top configurations and almost all the vent-streaming problems.     

全局减方差方法在乏燃料干式贮存容器屏蔽计算中的应用

开发了基于离散纵标(S_N)方法的蒙特卡罗(MC)全局减方差方法,针对乏燃料干式贮存容器,分别建立了中子源及光子源MC直接计算模型、S_N计算模型及全局减方差方法计算模型,并进行了计算精度和效率的比较。数值结果表明,全局减方差方法计算结果与无偏的MC及S_N计算结果相比吻合良好。其中S_N计算的次级光子剂量率与全局减方差方法计算的偏差较大,这主要是由于MC计算和S_N计算时的数据库差异导致的。和无偏的MC结果相比,全局减方差方法计算的中子及次级光子输运计算收敛效率提高了近2个数量级,初级光子输运计算收敛效率提高了1~2个数量级。     

蒙特卡罗模拟CFETR中子输运计算中的全局减方差方法应用及对比

在具有全局特性的蒙特卡罗输运精细计算的问题中,传统的MCNP(Monte Carlo N Particle Transport Code)局部减方差方法很难得到理想的计算结果,全局减方差方法(Global Variance Reduction,GVR)则是一种有效的解决方法。针对中国聚变工程试验反应堆(Chinese Fusion Engineering Testing Reactor,CFETR)的中子输运过程中减小全局方差的问题,将多种形式的GVR方法应用到柱状CFETR中子学模型的计算中。依据不同的中子分布信息,在算例中应用和对比了6种不同形式的GVR权窗,并对不同GVR方法的品质因子(FOMG)、标准差(σ)和有效计数率(Scoring)进行了分析。与AN(MCNPanalog method)相比,GVR方法的FOMG有很大的增长,误差在空间的分布也更加平缓,且具有更高的Scoring。在前人提出的全局减方差的基础上,在计算中应用一些新的GVR形式(能量、径迹数等),计算结果表明,基于中子通量的GVR方法的全局计算效率较AN提高了6.43倍。此外,基于中子能量的全局减方差方法也是一种可行的GVR应用形式,其与AN比较,计算效率提高了5.11倍。综上,基于中子通量的GVR方法具有最佳的全局减方差效果。     

反应堆临界-燃耗耦合蒙特卡罗计算

基于连续点截面MCNP程序 ,研制了三维多群P3 中子输运蒙特卡罗程序MCMG ,并与栅元均匀化程序WIMS耦合 ,实现了临界 燃耗耦合计算。采用WIMS产生的 69群共振、自屏宏观中子截面和BUGLE 80u47群微观中子截面 ,分别计算了简单反应堆和临界实验堆问题 ,计算结果与其它输运方法的计算结果和试验结果一致。在相同计算精度下 ,MCMG的计算时间较MCNP的计算时间少     

TRIPOLI-4® Monte Carlo code ITER A-lite neutronic model validation

3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4^® is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4^®, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4^® A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4^® is shown; discrepancies are mainly included in the statistical error.     

Assessment of fusion facility dose rate map using mesh adaptivity enhancements of hybrid Monte Carlo/deterministic techniques

Three mesh adaptivity algorithms were developed to facilitate and expedite the use of the CADIS and FW-CADIS hybrid Monte Carlo/deterministic techniques in accurate full-scale neutronics simulations of fusion energy systems with immense sizes and complicated geometries. First, a macromaterial approach enhances the fidelity of the deterministic models without changing the mesh. Second, a deterministic mesh refinement algorithm generates meshes that capture as much geometric detail as possible without exceeding a specified maximum number of mesh elements. Finally, a weight window coarsening algorithm decouples the weight window mesh and energy bins from the mesh and energy group structure of the deterministic calculations in order to remove the memory constraint of the weight window map from the deterministic mesh resolution. The three algorithms were used to enhance an FW-CADIS calculation of the prompt dose rate throughout the ITER experimental facility and resulted in a 23.3% increase in the number of mesh tally elements in which the dose rates were calculated in a 10-day Monte Carlo calculation. Additionally, because of the significant increase in the efficiency of FW-CADIS simulations, the three algorithms enabled this difficult calculation to be accurately solved on a regular computer cluster, eliminating the need for a world-class super computer.     

Evaluation of Shutdown Gamma-ray Dose Rates around the Duct Penetration by Three-Dimensional Monte Carlo Decay Gamma-ray Transport Calculation with Variance Reduction Method

For the evaluation of gamma-ray dose rates around the duct penetrations after shutdown of nuclear fusion reactor, the calculation method is proposed with an application of the Monte Carlo neutron and decay gamma-ray transport calculation. For the radioisotope production rates during operation, the Monte Carlo calculation is conducted by the modification of the nuclear data library replacing a prompt gamma-ray spectrum with a decay gamma-ray spectrum. By multiplying each correction factor, which is ratio of the actual activation level after shutdown to the production rate during operation, with each decay gamma-ray flux due to each radioisotope, the decay gamma-ray dose rate is evaluated. In order to improve the statistical error, a variance reduction method is proposed by the application of the weight window importance technique and the specification of the decay gamma-ray generation location. We identify the cell producing the decay gamma-ray which can contribute the decay gamma-ray flux in evaluation locations, and forcibly terminate the gamma-ray transport calculation in the cells except for the identified cells. In order to validate the effectiveness of the method, shielding calculation for actual ITER (International Thermonuclear Experimental Reactor) configuration is performed, and small statistical errors below criteria are obtained. The effectiveness of the proposed method for ITER design analysis is demonstrated.     

Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP–ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB.     

Shut-Down Dose Rate analysis for ITER Diagnostic Equatorial and Upper Ports

The Shut-Down Dose Rate (SDDR) is an important criterion of radiation safety for the personnel access for maintenance operations in ITER ports after the cessation of the D-T 14 MeV neutron fusion source. Therefore, the problem of the SDDR calculations attracts the attention of fusion neutronics community because SDDR in such a large and geometrically complicated fusion device as the ITER tokamak is challenging to compute. This challenge has been faced and overcome by applying dedicated methodological approaches explained in this paper. The results of the SDDR analysis allowed us to propose several design solutions for improvement of the radiation shielding of the ITER Generic Diagnostic Equatorial and Upper Port Plugs (EPP and UPP). The SDDR analysis was focused on the interspace area located between the ITER bioshield and port plugs where the personnel access is envisaged at ∼12 days after the ITER shut-down. By this analysis the radiation streaming pathways and dominant sources of decay radiation were revealed and the methods to mitigate the streaming and subsequent activation were found. The optimization of the port shielding was targeted on minimization of the SDDR in the interspace area following the ALARA principle and taking into account the feasibility to implement proposed shielding options with the actual hardware. Among them, wrapping the EPP walls with the B₄C tiles improves the EPP shielding performance. While void around the ELM/in-vessel coils and blanket manifolds leads to the performance reduction. The SDDR inside the Generic UPP interspace depends mainly on the environment (blanket, manifolds, and gaps).     

Biased Selection of Particle Flight Directions as a Variance Reduction Technique in Monte Carlo Calculations

As a practical variance reduction technique applicable to Monte Carlo shielding calculations, the present article shows a new simple biased sampling technique on particle flight directions. Scattered particles not directed towards the detector positions are killed if they are not so important, that is, if the particle weights are sufficiently small compared to the source weight. In this way, we can reduce the sample size required for obtaining an accurate estimate for the detector response.

The present technique was incorporated into the multigroup neutron and γ-ray transport code MORSE, and sample calculations were performed on spherical fast neutron systems. The results have shown that this biased technique is effective for dealing with neutron multiplication as well as neutron transmission problems. The neutron flux or the effective multiplication factor of a nuclear reactor is estimated more accurately than from the method of path-length stretching with about the same computation time. In addition, it is shown that the flight-direction biasing can further effectively be used by combining it with other variance reduction techniques.     

New Sampling Method in Continuous Energy Monte Carlo Calculation for Pebble Bed Reactors

A pebble bed reactor generally has double heterogeneity consisting of two kinds of spherical fuel element. In the core, there exist many fuel balls piled up randomly in a high packing fraction. And each fuel ball contains a lot of small fuel particles which are also distributed randomly. In this study, to realize precise neutron transport calculation of such reactors with the continuous energy Monte Carlo method, a new sampling method has been developed. The new method has been implemented in the general purpose Monte Carlo code MCNP to develop a modified version MCNP-BALL. This method was validated by calculating inventory of spherical fuel elements arranged successively by sampling during transport calculation and also by performing criticality calculations in ordered packing models. From the results, it was confirmed that the inventory of spherical fuel elements could be reproduced using MCNP-BALL within a sufficient accuracy of 0.2%. And the comparison of criticality calculations in ordered packing models between MCNP-BALL and the reference method shows excellent agreement in neutron spectrum as well as multiplication factor.

MCNP-BALL enables us to analyze pebble bed type cores such as PROTEUS precisely with the continuous energy Monte Carlo method.     

核聚变实验装置HT-7U停机辐射剂量率三维计算与分析

在基于三维蒙特卡罗方法的聚变装置停机剂量率计算方法“严格两步法”(R2S)的基础上,首先建立了核聚变托卡马克实验装置HT-7U三维精确模型,然后对HT-7U各种D-D放电模式下的停机剂量率进行了详细的三维计算与分析,从而为装置实验方案及实验维修人员的安全操作规程的制定提供了理论基础,也对装置的辐射防护问题具有参考价值。     

Fast assessment of the error in the neutron distribution in a large reactor in Monte Carlo calculations

It is suggested that there is a close analogy between the statistical error of local characteristics in a Monte Carlo calculation of a large reactor and random deviations of the multiplication properties of these cells from a nominal value within technological tolerance limits. It is well-known that the latter result in global and strongly correlated deformations of the neutron field which are especially noticeable in large reactors. The scale of the deformations, or the statistical error, of the neutron field is determined by a formula obtained from an analysis of the influence of technological tolerances. Model Monte Carlo calculations confirm that this analogy is correct. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 115–119, August, 2007.     

Shielding design and analyses of KIPT neutron source facility

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of a neutron source facility. An electron accelerator drives a sub-critical facility (ADS) is used for generating the neutron source. The facility will be utilized for performing basic and applied nuclear researches, producing medical isotopes, and training young nuclear specialists. Monte Carlo code MCNPX has been utilized as the major design tool for the design, due to its capability to transport electrons, photons, and neutrons at high energies. However the ADS shielding calculations with MCNPX need enormous computational resources and the small neutron yield per electron makes sampling difficulty for the Monte Carlo calculations. The high energy electrons (E > 100 MeV) generate very high energy neutrons and these neutrons dominant the total radiation dose outside the shield. The radiation dose caused by high energy neutrons is ∼3-4 orders of magnitude higher than that of the photons. However, the high energy neutron fraction within the total generated neutrons is very small, which increases the sampling difficulty and the required computational time. To solve these difficulties, the user subroutines of MCNPX are utilized to generate a neutron source file, which record the generated neutrons from the photonuclear reactions caused by electrons. This neutron source file is utilized many times in the following MCNPX calculations for weight windows (importance function) generation and radiation dose calculations. In addition, the neutron source file can be sampled multiple times to improve the statistics of the calculated results. In this way the expensive electron transport calculations can be performed once with good statistics for the different ADS shielding problems. This paper presents the method of generating and utilizing the neutron source file by MCNPX for the ADS shielding calculation and similar accelerator facilities, and the accurate radiation dose analyses outside the shield using modest computational resources.     

Importance Determination Method for Geometry Splitting with Russian Roulette in Monte Carlo Calculations of Thick and Complicated Core Shielding Structure

In case of a shielding analysis of the geometry having thick and complicated structures with a Monte Carlo code, it is a serious problem that it takes too much computer time to obtain results with good statistics. Therefore, it is very important to reduce variances in the calculation. In this study, a method to determine the importance function in 3-dimensional Monte Carlo calculation with geometry splitting with Russian roulette was developed for the shielding analysis of thick and complicated core shielding structures. Only two essential importance ratio curves for one material enable us to determine the importance function easily in the shielding calculation.

The validity of this method was confirmed through a simple benchmark calculation. From the comparison with the result obtained by using weight window (W-W), it was shown that the present method can give an accurate result on the same level with W-W method with less trial and errors. And this method was applied to an actual reactor core shielding analysis to confirm its applicability to a 3-dimensional thick and complicated structure.

Using this method, the variance reduced calculation can be easily realized with the developed importance determination procedure, especially in case that parameter survey calculations are required in order to determine the shield thickness in a design work of a thick and complicated structure. Accordingly, it became easier to use Monte Carlo method as a powerful tool for a reactor core shielding design.     

网格权窗减方差技术及其在聚变堆屏蔽分析中应用研究

在聚变堆辐射屏蔽计算中,如何有效解决深穿透问题是近年来国际聚变辐射安全领域关注的焦点之一。针对该问题,本文研究了直角坐标系与圆柱坐标系下基于网格的权窗减方差技术。本文基于超级蒙卡核模拟软件系统SuperMC实现了该方法,并选取减方差技巧的基准例题进行测试与分析,初步得出"粗划真空或密度很小的区域、细分密度大的区域"的网格划分规律,能有效提高网格权窗计算效率。基于该规律对聚变屏蔽基准问题进行对比分析,新的网格划分与原始网格划分的计算效率相比,FOM因子提高了1.92倍。减方差技巧的基准例题和聚变屏蔽基准问题计算中,SuperMC通量计算结果与MCNP相比偏差均在0.5%以下,证明了本文中方法的正确性。     

Application of local Monte Carlo method in neutronics calculation of EAST radial neutron camera

The Local Monte Carlo (LMC) method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera (RNC) diagnostic system on the experimental advanced superconducting tokamak (EAST), and the radiation distribution of the RNC and the neutron flux at the detector positions of each channel are obtained. Compared with the results calculated by the global variance reduction method, it is shown that the LMC calculation is reliable within the reasonable error range. The calculation process of LMC is analyzed in detail, and the transport process of radiation particles is simulated in two steps. In the first step, an integrated neutronics model considering the complex window environment and a neutron source model based on EAST plasma shape are used to support the calculation. The particle information on the equivalent surface is analyzed to evaluate the rationality of settings of equivalent surface source and boundary. Based on the characteristic that only a local geometric model is needed in the second step, it is shown that the LMC is an advantageous calculation method for the nuclear shielding design of tokamak diagnostic systems.