WEBEXPIR: Windowless target electron beam experimental irradiation

The windowless target electron beam experimental irradiation (WEBEXPIR) program was set-up as part of the MYRRHA/XT-ADS R&D effort on the spallation target design to investigate the interaction of a proton beam with a liquid lead-bismuth eutectic (LBE) free surface. In particular, possible free surface distortion or shockwave effects in nominal conditions and during sudden beam on/off transient situations, as well as possible enhanced evaporation were assessed. An experiment was conceived at the IBA TT-1000 Rhodotron, where a 7 MeV electron beam was used to simulate the high power deposition at the MYRRHA/XT-ADS LBE free surface. The geometry and the LBE flow characteristics in the WEBEXPIR set-up were made as representative as possible of the actual situation in the MYRRHA/XT-ADS spallation target. Irradiation experiments were carried out at beam currents of up to 10 mA, corresponding to 40 times the nominal beam current necessary to reproduce the MYRRHA/XT-ADS conditions. Preliminary analyses show that the WEBEXPIR free surface flow was not disturbed by the interaction with the electron beam and that vacuum conditions stayed well within the design specifications.     

Accelerator driven systems for transmutation: Fuel development, design and safety

European R&D for ADS design and fuel development is driven in the 6th FP of the EU by the EUROTRANS Programme. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT. The XT-ADS is designed to provide the experimental demonstration of transmutation. The EFIT, the European Facility for Industrial Transmutation, aims at a conceptual design of a full transmuter. A key R&D issue is the choice of an adequate fuel. Various fuel forms have been assessed and CERCER and CERMET fuels, specifically the matrices MgO and Mo, have finally been selected. Within EUROTRANS, the domain ‘AFTRA’ is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel database for the EFIT. The EFIT is optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. In the current paper the fuels under investigation are described, including their production route and their safety limits. First core designs of CERCER and CERMET fuelled 400 MWth EFITs have been developed within AFTRA. The trends found in the design studies and first safety analyses are presented.     

European research on HLM thermal-hydraulics for ADS applications

The objective of the European 6th framework project EUROTRANS is to demonstrate the technical feasibility of transmutation of high level nuclear waste using accelerator driven systems. Within this objective the design of a European experimental ADS should demonstrate the technical feasibilities to transmute a sizeable amount of waste and to operate an ADS safely. This ADS will be a subcritical reactor system having liquid lead-bismuth eutectic as coolant. The liquid lead-bismuth eutectic is also intended to serve as target material for the spallation reaction which forms a crucial part to the subcritical reactor core. Since lead-bismuth eutectic is used as core coolant and spallation material, knowledge of its thermal hydraulic behaviour is essential. Within the DEMETRA domain of the EUROTRANS project, basic thermal hydraulic studies in order to support the design and safety analysis of XT-ADS components and the development of measurement techniques have been started.     

Addressing the heavy liquid metal – Water interaction issue in LBE system

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.     

Compatibility of ferritic-martensitic steel T91 welds with liquid lead-bismuth eutectic: Comparison between TIG and EB welds

The 9 wt.% chromium ferritic-martensitic steel T91 is being considered as candidate structural material for a future experimental accelerator driven system (XT-ADS). This material and its welded connections would need to be used in contact with liquid lead-bismuth eutectic (LBE), under high irradiation doses. Both unirradiated tungsten inert gas (TIG) and electron beam (EB) welds of T91 have been examined by means of metallography, scanning electron microscopy (SEM-EDX), transmission electron microscopy (TEM), Vickers hardness measurements and tensile testing in both gas and liquid lead-bismuth environment. The TIG weld was commercially produced and post weld heat treated by a certified welding company while the post weld heat treatment of the experimental EB weld was optimized in terms of the Vickers hardness profile across the welded joint. The mechanical properties of the T91 TIG and EB welds in contact with LBE have been examined using slow strain rate tensile testing (SSRT) in LBE at 350 °C. All welds showed good mechanical behaviour in gas environment but total elongation was strongly reduced due to liquid metal embrittlement (LME) when tested in liquid lead-bismuth eutectic environment. The reduction in total elongation due to LME was larger for the commercially TIG welded joint than for the EB welded joint.     

Operations strategy for workload balancing of crews in an advanced main control room

The advanced main control room (advanced-MCR) is the one that allows for reactor operations based on digital instrumentation and control (I&C) technology. Thus, the operators of an advanced-MCR operate the plant through digital I&C interfaces, and for this purpose, an additional digital manipulation task for the operating equipment should be performed that cannot be observed in a conventional-MCR. As a prior study proposing the cognitive, communicative, and operational activity measurement approach (COCOA), COCOA enables an evaluation of the operator's workload in an advanced-MCR, which includes newly generated tasks for Man-Machine Interface System based secondary operation under a digital environment, which does not exist in a conventional-MCR. As a result of observations on the workload level by utilizing COCOA for a reference plant with an advanced-MCR when conducting an emergency operating procedure, it was observed that the workload of the shift supervisor is about two times greater than that of other operators. This is because operators therein stuck to the old guidelines customized to a conventional-MCR and failed to accomplish load balancing in consideration of the operation environment that an advanced-MCR provides. In this context, it would be imperative to develop and apply an operations strategy for an advanced-MCR operation. This study proposes an operations strategy in an attempt to make a balanced workload of operators in an advanced-MCR.     

Smearing in an axisymmetric finite element structural analysis application to the GCFR steam generator cavity closure plug

A methodical technique for smearing material properties in an elastic finite element analysis to derive an approximate axisymmetric model of a structure is presented. In this process, material properties are smeared circumferentially about an axis of symmetry in the structure's outline. Attention is given to the subsequent problem of unsmearing the results to obtain an estimate of circumferential variation in the solution. The scheme is illustrated with an application to an analysis of a GCFR steam generator cavity closure plug.     

核电设备公路运输的风险分析与控制

核电设备公路运输过程是设备制造完成后到安装投入使用前的一个重要环节,是设备制造价值向安装使用价值转化的一个必要阶段;在运输各环节,存在不同的风险因素。本文以秦山二期扩建工程3号机组稳压器运输过程为例,对核电设备公路运输各阶段的风险因素进行分析,提出相应的风险控制措施,以期完善核电设备公路运输过程的质量控制。     

Spent fuel transportation risk assessment: cask impact analyses

Abstract

The Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. This assessment considered the response of three certified casks to a range of impact accidents in order to determine whether or not they would lose their ability to contain the spent fuel or maintain effective shielding. The casks consisted of a lead shielded rail cask that can be transported either with or without an inner welded canister, an all-steel rail cask that is transported with an inner welded canister, and a DU shielded truck cask that is transported with directly loaded fuel. Finite element analyses were performed for impacts at speeds of 48, 97, 145 and 193 kilometres per hour into a rigid target. Impacts in end-on, side-on, and CG-over-corner orientations were analysed for each cask and impact speed. Calculations were performed to equate these impacts onto rigid targets with higher speed impacts onto the yielding targets that exist in the real world. These analyses indicated that a cask with an inner welded canister or a truck cask would not release radioactive material in any impact accident and that only very high-speed impacts onto hard rock targets could result in either release of material or significant degradation of shielding for rail casks without an inner canister. Impacts other than those onto flat unyielding targets were also considered. Analyses show that an impact that bypasses the impact limiters on the ends of the casks does not result in seal failure and neither does an impact by a locomotive also between the impact limiters.     

A MAUT approach for selecting a dismantling scenario for the thermal column in KRR-1

In this paper, an evaluation methodology for selecting an optimal dismantling scenario at the stage of a decommissioning planning is presented. The selection of a dismantling scenario is critical for an entire decommissioning project because by the selected dismantling scenario, the dismantling schedule, the dismantling cost, a worker’s exposure, and the work safety level are determined. So it is necessary that an optimal scenario for dismantling an object is selected through an evaluation of the relevant scenarios in consideration of several evaluation items. In this research, as an evaluation model, the multi-attribute utility theory (MAUT) is quantitatively and qualitatively applied to assess these scenarios. Quantification of an evaluation item in the MAUT method has been performed by using a single utility function as well as a multi-utility function for each item. As an implementation, the MAUT method is applied to select the best scenario for the thermal column in Korea Research Reactor-1 (KRR-1) and its advantages and disadvantages compared with the (analytic hierarchy process) AHP method about the application to a decommissioning project.     

大规模核与辐射事故受照人员生物剂量估算的应对策略

在发生大规模核与辐射事故后对受照人员的早期临床分类诊断中,生物剂量估算起着不可或缺的重要作用,制定有效的生物剂量估算应对策略是值得学者们关注和讨论的问题。本文在简要介绍事故照射事件的情景、特征、生物剂量计的作用基础上,对发生大规模核与辐射事故时对估算受照人员生物剂量的应对策略进行简要综述。     

Formal qualification of the various sources of an uncertainty employed in the level 2 PSA and the underlying issues

In the Level 2 PSA, a probabilistic treatment of complex phenomenological accident pathways inevitably leads to two sources of an uncertainty: (a) an incomplete modeling of these accident pathways and their subsequent impacts on the Level 2 risk, and (b) an expert-to-expert variation in the their probabilistic estimates. The impacts of the foregoing sources of an uncertainty on the Level 2 risk measures are different from each other, thus leading to a different conclusion in the decision-making process. An important aspect of the foregoing distinction of an uncertainty is that it plays an essential role in identifying what sources of an uncertainty impact more on the Level 2 risk and what sources among them should be focused on for a greater reduction of the overall Level 2 uncertainty. Another aspect is closely related to its importance in combining the Levels 1 and 2 uncertainties. A primary objective of this paper is to qualify formally the typical sources of an uncertainty that would often be employed in the Level 2 PSA and the underlying issues for a further clarification.     

U型管阵流体弹性不稳定性分析

从U型管振动方程出发,推导出 U 型管阵流体诱发振动方程,给出了管阵出现流体弹性不稳定性临界流速的计算方法及算例。结果表明:管阵的临界流速与其频率取 U 型管阵中各管子一阶频率平均值的直管管阵的流速基本一致。     

Calculation for Three-Dimensional Structures of Two-Phase Flow in Enlarged Flow Area

Superconducting tunnel junctions in use for X-ray detection ale expected to produce an excellent energy resolution. An inclusive model is employed to give an overview of the detector characteristics. This method is devised to implicitly take properties of superconductor materials into account. With introducing an empirical relationship between junction parameters, the detector characteristics such as signal magnitude and energy resolution are reproduced to an acceptable degree. This method gives a helpful suggestion for understanding the detector performance such as a Fano-like factor, and may also serve as a guide line for designing tunnel junctions.     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

半球形CdZnTe伽马谱仪在放射性"热点"测量中的应用

核设施退役过程中,对退役场所进行"热点"调查(包括"热点"定位,活度测量等)是非常重要的.为了配合伽马相机在热点调查后对"热点"进行更为精确的测量,建立了一套配有铅屏蔽准直器的CdZnTe伽马能谱系统,用Visual c#语言编写了便于现场操作的谱分析软件MiniAnalysis,对现场用CdZnTe谱仪测得的能谱进行...     

A Fusion Development Facility Based on the Spherical Torus

The Fusion community could soon have available an opportunity to significantly advance fusion development, an opportunity afforded by the potential of the spherical torus as a plasma confinement system. That opportunity is, in a single device at an affordable price and at a suitable site, to move sequentially through the major fusion objectives of advanced confinement physics, burning plasma and DT physics, blanket and other fusion nuclear technology development, tritium self-sufficiency, and, perhaps at the end with an upside performance outcome, a chance at net electric breakeven. Meeting this set of objectives would dramatically advance the development of fusion.     

Effect of multi-shot X-ray exposures in IFE armor materials

As part of the High Average Power Laser (HAPL) program the performance of tungsten as an armor material is being studied. While the armor would be exposed to neutrons, X-rays and ions within an inertial fusion energy (IFE) power plant, the thermomechanical effects are believed to dominate. Using a pulsed X-ray source, long-term exposures of tungsten have been completed at fluences that are of interest for the IFE application. Modeling is used in conjunction with experiments on the XAPPER X-ray damage facility in an effort to recreate the effects that would be expected in an operating IFE power plant. X-ray exposures have been completed for a variety of X-ray fluences and number of shots. Analysis of the samples suggests that surface roughening has a threshold that is very close to the fluences that reproduce the peak temperatures expected in an IFE armor material.     

Detailed Evaluation of the Natural Circulation Mass Flow Rate of Water Propelled by Using an Air Injection

One-dimensional (1D) air-water two-phase natural circulation flow in the “thermohydraulic evaluation of reactor cooling mechanism by external self-induced flow—one-dimensional” (THERMES-1D) experiment has been verified and evaluated by using the RELAP5/MOD3 computer code. Experimental results on the 1D natural circulation mass flow rate of water propelled by using an air injection have been evaluated in detail. The RELAP5 results have shown that an increase in the air injection rate to 50% of the total heat flux leads to an increase in the water circulation mass flow rate. However, an increase in the air injection rate from 50 to 100% does not affect the water circulation mass flow rate, because of the inlet area condition. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it has no influence on the local pressure. An increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not have an influence on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases.     

Numerical comparison between the ICF and the ICF-spherical pinch

The spherical pinch concept is an outgrowth of the inertial confinement model. The salient feature of the spherical pinch concept is the creation of a hot plasma in the center of a sphere.^(1,2) This plasma is then compressed by a strong shock wave launched from the periphery of the vessel by an imploded plasma acting as a driver. This scheme, reveals that convergence of the shock, which is the main feature of all inertial confinement schemes, is a component of the spherical pinch model. The reasons for classifying the spherical pinch as a particular ICF model and designating it as a ICF-SP are given here. The fluid mechanics and high-temperature hydrodynamics of the spherical pinch can be briefly described as an explosion within an implosion. The structure of the shock wave for such explosion within an implosion and for, an implosion alone is determined by solving numerically the governing equations of the phenomena. We present here a detailed computational comparison of the inertial confinement model and the spherical pinch in terms of density, pressure, temperature, confinement time, total accumulated number of neutrons, and time-resolved neutron flux from reactions in deuterium-tritium mixture. It is shown that temperature, confinement time, and total number of neutrons for the ICF-Spherical Pinch improve upon the classical ICF.