Criticality qualification of a new Monte Carlo code for reactor core analysis

In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal–hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the “Accelerator part” of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the “Reactor part” of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.     

Experiments for a core catcher concept based on water addition from below

A core catcher concept is proposed to be integrated into a new PWR design based on the standard German PWR. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core-melt based on the process of water inlet from the bottom through the melt.To ge more detailed information on the very important process of water penetrating into the melt, simulant experiments have been conducted using a transparent plastic and a solder melt representing the oxidic and metallic part of the core-melt. It appears from the results that fragmentation of the melts can be achieved by proper selection of water supply pressure and water feed cross-section.The important part of the transient medium scale experiments with thermite melts, conducted since mid 1993, is to get information on the process of evaporation of water by water ingression in hot melts from below and to investigate whether there is a possibility of strong melt-water interactions, or even steam explosions. The experimental set-up represents a section of the core catcher. A thermite melt is located on the catcher plate with water supply from the bottom. After ignition of the melt, the upper sacrificial layer is eroded until water penetrates into the melt from the bottom through the holes in the supporting plate and fragmentation and simultaneous solidification of the melt occurs. The experiments, up to now, show that flooding and early coolability of the melt by water addition from the bottom are achieved.These experiments serve also as pretests for the COMET-H experiments with sustained heating planned to be conducted in the BETA-facility at the beginning of next year.     

CEFR-MOX新燃料组件运输货包临界安全计算

易裂变材料运输过程中重要的安全问题之一是临界安全。在对运输货包进行临界安全分析中必须要同时考虑多货包阵列形式、事故后货包损伤对临界安全影响、最佳水慢化条件等因素。本文采用MCNP 程序针对CEFR-MOX新燃料组件运输货包进行了临界安全计算。计算结果表明:MCNP程序(采用核截面库为ENDF/B-V库)对本问题的次临界限值为0.924 6;正常运输条件下无限个运输货包的最大k_eff值为0.574 4,运输事故条件下无限个运输货包的最大k_eff值为0.659 7。根据临界安全指数的定义,确定CEFR-MOX新燃料组件运输货包的临界安全指数为0。     

CNSC乏燃料组件运输容器临界安全分析

临界安全作为乏燃料组件运输容器的一项重要安全指标,需经过计算和分析以判断其是否满足法规标准。为分析中国核工业集团有限公司(China National Nuclear Corporation,CNSC)乏燃料组件运输容器临界安全设计是否满足《放射性物品安全运输规程》的要求,使用蒙特卡罗程序MCNP(Monte Carlo N Particle Transport Code)构建了保守临界计算模型,对正常和事故工况下CNSC乏燃料组件运输容器进行了临界计算分析。分析表明:正常运输条件下单个货包和货包阵列的k_(eff)最大值为0.804 25,小于次临界限值,临界安全指数为0;事故工况下单个货包和货包阵列的k_(eff)最大值为0.813 17,小于次临界限值,临界安全指数为0。可见,正常和事故工况下,CNSC乏燃料组件运输容器的keff最大值均小于0.94的次临界限值,临界安全指数为0,满足法规标准要求。     

三门核电一号机组新燃料组件在冲洗井中临时存放的临界安全计算

新燃料贮存主要涉及到的核安全问题是临界安全。对新燃料组件临时存放的方案进行临界安全计算时必须考虑两项独立事件的情况下组件的临界安全。采用SCALE对三门核电将首炉新燃料组件存放于冲洗井中的方案进行了临界计算。计算结果显示,即使在极限的两项独立事件同时发生的前提下,三门核电首炉机组的燃料组件在冲洗井中存放的临界安全也是有保障的。     

Criticality analysis of pulse core and laser module coupled small reactor with low enriched uranium

Nuclear-pumped laser can directly convert nuclear energy to optical energy. A coupled reactor which consists of two pulse cores with highly enriched metallic uranium and a subcritical thermal laser module with highly enriched metallic uranium is one of the reactors for nuclear-pumped laser. In this paper, criticality analysis of a coupled reactor which consists of pulse cores with 20% enriched metallic uranium and a subcritical thermal laser module with 20% enriched metallic uranium was performed by Monte Carlo calculation. The result of criticality analysis showed the following three points. First, a coupled reactor with 20% enriched metallic uranium can achieve criticality. Second, using eight pulse cores in axial direction is effective to achieve flattened axial power distribution in the laser module. Third, less than 20% of the energy released from fissions in the whole coupled reactor has the possibility to be converted to optical energy for a coupled reactor with 100% enriched uranium, and less than 7% for a coupled reactor with 20% enriched uranium.     

Corium retention for high power reactors by an in-vessel core catcher in combination with External Reactor Vessel Cooling

If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.     

Investigation of ex-vessel core catcher for SBO accident in VVER-1000/V528 containment using MELCOR code

Nuclear Science and Techniques - An S-band high-gradient accelerating structure is designed for a proton therapy linear accelerator (linac) to accommodate the new development of compact,...     

Criticality and burnup analyses of a PBMR-400 full core using Monte Carlo calculation method

In this study, the criticality and burnup analyses have been performed for full core model of Pebble Bed Modular Reactors, such as PBMR-400, using the computer codes MCNP5.1.4 and MONTEBURNS 2.0. Three different pebble distributions, namely; Body Centered Cubic (BCC) (packing fraction = 68%), Random Packing (RP) (packing fraction = 61%) and Simple Cubic (SC) (packing fraction = 52%) were selected for the analyses. The calculated core effective multiplication factor, k_eff, for BCC, RP and SC came to be 1.2395, 1.2357 and 1.2223, respectively. The core life for these distributions were calculated as ～1200, 1000, and 800 Effective Full Power Days (EFPDs), whereas, the corresponding burnups came out to be ～99,000, ～92,000 and ～86,000 MWD/T, respectively, for end of life k_eff set equal to 1.02.     

Structural homogenized analysis for a nuclear reactor core

A nuclear reactor core is composed of a great number of tubular beams with periodic structure, which are immersed in an acoustic fluid. In the present paper, a 3-D homogenization model is developed to predict its overall dynamic behavior. An approximate solution to the local problem is given. The application to an 1-D example shows that approximate expressions of the natural frequency, the added fluid mass and the equivalent sound speed can be used in engineering estimation.     

Preliminary safety analysis on loss of flow accidents and external source transients for LBE cooled ADSR core

In this paper, preliminary safety studies on the 800 MWth accelerator-driven system (ADS) proposed by Xi'an Jiaotong university are presented. The system is a pool type facility coupling a proton accelerator with current in the range of 17–23 mA and a sub-critical core by means of a spallation target. The RELAP5/MOD3.3 code is selected as a base tool. In order to simulate the system, the point kinetics model is modified and the property of lead-bismuth is implemented to meet the requirement of ADS analysis. This paper focuses on the assessment of its response to the loss of flow events. The first part is originated from the failure of the pump and the second part derives from the significant flow blockage at a fuel assembly inlet. The reactivity insertion accidents are caused by the change of the proton beam current. The results show that the safety and criteria are satisfied and the system is tolerant to the loss of flow accidents and proton beam doubled accident and is sensitive to the external neutron changing.     

Burnup in Criticality Safety for Spent Fuel Casks

Abstract

Criticality safety margins must be based upon the combination of the best available prediction of the margin and all uncertainties in the prediction. Inclusion of the effects of burnup in the evaluation of spent fuel shipping or storage casks must be based upon a thorough understanding of the prediction of the effects of burnup and the uncertainties in the measurements (or predictions) of burnup and predictions of the effects. A preliminary estimate of the effects of burnup and its uncertainties is presented. This will serve as a first step in the effort to develop acceptance criteria that assure public safety. An assembly average burnup of 20,000 MWD/MTU represents an increase in the criticality safety margin of about 20% (δk/k), and the current estimate of the uncertainty in this value is close to 4% (δk/k). The uncertainties in the components of the effects of burnup were based upon relevant literature citations and — where no other information was available — upon estimates. Consequently, the margins and uncertainties in the margin presented here should be considered as initial estimates upon which more refined analyses should build to develop a defensible basis for predicting and reviewing the criticality safety margins which include the effects of burnup.     

Performance improvement of artificial neural networks designed for safety key parameters prediction in nuclear research reactors

The present work explores, through a comprehensive sensitivity study, a new methodology to find a suitable artificial neural network architecture which improves its performances capabilities in predicting two significant parameters in safety assessment i.e. the multiplication factor k_eff and the fuel powers peaks P_max of the benchmark 10 MW IAEA LEU core research reactor. The performances under consideration were the improvement of network predictions during the validation process and the speed up of computational time during the training phase.To reach this objective, we took benefit from Neural Network MATLAB Toolbox to carry out a widespread sensitivity study. Consequently, the speed up of several popular algorithms has been assessed during the training process. The comprehensive neural system was subsequently trained on different transfer functions, number of hidden neurons, levels of error and size of generalization corpus.Thus, using a personal computer with data created from preceding work, the final results obtained for the treated benchmark were improved in both network generalization phase and much more in computational time during the training process in comparison to the results obtained previously.     

A two-time step algorithm for seismic analysis of a PWR core

The present analysis considers the colliding vibration of a row of fuel assemblies under seismic excitation from the core plates. The differential equation model of each assembly is integrated separately, the impact forces due to the shocks between the spacer grids, being considered as external forces and reevaluated at each time step. The linear behaviour of the fuel assemblies for low amplitude vibrations allows a modal superposition.Two different time steps are considered in the analysis; the first one, related to the seismic excitation, applies to the fuel assemblies not involved in shocks; the second one, a submultiple of the first and related to the impact characteristics, is used for those assemblies involved in shocks.The resulting computer program, CLASH, is relatively economical to use; this allows such analysis becoming part of the standard design procedures.     

Development of , a seismic analysis program for FBR core components

A computer program, (seismic analysis program for fuel assemblies) has been developed to analyze core component vibration in fast breeder reactors (FBRs) during seismic excitation. Since an FBR core is composed of as many as 1000 core subassemblies (fuel assemblies, blanket assemblies, neutron shield assemblies, etc.), which are immersed in a coolant fluid, seismic analysis of FBR cores must consider the vibrations generated in a system with a large number of degrees of freedom with impacts under fluid-structure interactions. models subassemblies as finite beam elements. Fluid interaction forces are considered as added mass and time integration is done using mode superposition and the Nigam method. The load pad impact is modeled using a gap, a linear spring and a linear damper. The program also uses a new method to determine the nonlinear impact force, making it unnecessary to use convergence iteration. Comparison with experimental results confirms that the program can closely predict the seismic response of FBR cores.     

A seismic analysis method for a block column gas-cooled reactor core

An analytical method for predicting the behavior of a prismatic high-temperature gas-cooled reactor (HTGR) core under seismic excitation has been developed. In this analytical method, blocks are treated as rigid bodies, are constrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions. Coulomb friction between blocks and between dowel holes and pins is also considered. A spring dashpot model is used for the collision process between adjacent blocks and between blocks and boundary walls.Analytical results are compared with experimental results and are found to be in good agreement. The analytical method can be used to predict the behavior of the HTGR core under seismic excitation.     

Criticality theory for a reactor in the case of fluctuations in the fuel density

Conclusions In Fig. 2 we show graphs of the dependence of the additional reactivity that arises as a result of fluctuations of the fuel density. As follows from Fig. 2, the increase in the reactivity for sufficiently large reactors and for (/₀) 0.1–0.05 is comparable with the contribution of the delayed neutrons. Thus, it is in principle possible to regulate the criticality of the reactor by exciting fluctuations of the density of a gaseous fuel. Regulation in this manner has decided advantages. Thus, the time in which the reactivity changes which determines the transient processes, is short — of the order of one period of the fluctuation. Moreover, there is practically no danger of accidents, since the reactivity falls as soon as the fluctuations cease. The amplitude of the fluctuation of the neutron flux (see, for example, the expression (20)) always exceeds the amplitude of the fluctuations in the fuel density by (k–1)^–1/2. This circumstance may be exploited to obtain a neutron flux pulsating with a large amplitude.This effect of a growth in thereactivity as a result of fluctuations of the fuel density may prove important in the study of the possibility of self-oscillatory conditions of operation in reactors with a high neutron flux.Translated from Atomnaya Énergiya, Vol. 27, No. 2, pp. 107–111, August, 1969.     

MCNP程序在反应堆临界计算中的应用

用三维的蒙特卡罗程序（MCNP）进行临界计算，着重介绍堆芯和反射层的建模，利用MCNP程序的重复结构功能简化对堆芯的描述，以JRR3为例计算了几个不同棒位于Keff值，计算结果与参值吻合较好，表明MCNP程序能够用于反应堆的临界计算。     

Development of a safety analysis code for molten salt reactors

The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.