Mock-up Experiment Research Articles

A Neutronics Shielding Mock-Up Experiment for Reduction of Uncertainty on the Prediction of the ITER-TFC Nuclear Heating

A mock-up of the inboard shield and vacuum vessel of ITER was set up at ENEA Frascati and irradiated with 14-MeV neutrons produced by the Frascati Neutron Generator. The mock-up includes the coil region, and its dimensions and materials composition are consistent with the latest ITER design. The objective of the experiment is to validate the calculations of nuclear heating in the ITER toroidal field coil performed with the Monte Carlo code MCNP-5 and the FENDL-2.1 library, through measurement performed by thermoluminescent dosimeters. The goal was to reach a calculated-to-experiment accuracy of less than or equal to ±10%, possibly as low as ±8%.

조명장치의 이동과 자동 교체가 가능한 LED 무대조명 시스템

A performance delivers the contents and feeling through human bodies and stage settings in the limited space of a stage. Among the stage settings, the lighting creates overall atmospheres effectively according to the situations of the performance such as a time flow, a change of seasons, an expression of the mental state of the performer. As the LED lighting is used more widely, the number and the size of the lighting is being decreased, which makes it possible for various devices to be used. However, just as conventional lighting devices, most of them are fixed. So it is difficult to use them at a blind spot and to replace them. To solve this problem, this article suggests a system which replaces lighting automatically and moves. As it can moved to a spot we want on a rail, and the lighting system can be replaced by itself, it can have various lighting effects, increasing the effectiveness of the performance. Additionally, it can also decrease the danger of a worker operating the lighting device located at a high place. At the mock-up experiment, its mobility and stability on a rail were tested using the wheel of the Modular Robot.

XT-ADS Windowless spallation target thermohydraulic design & experimental setup

The objective of the European 6th framework Integrated Project (IP) EUROTRANS (EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in an Accelerator Driven System) is to demonstrate the feasibility of transmutation of high level nuclear waste using subcritical Accelerator Driven Systems (ADS). The spallation target represents the most challenging new component in an ADS since it is the component coupling the accelerator and the nuclear core and is subjected to very high thermal load in a high radiation field. In this document the thermal hydraulic activities which led to reliable design rules for a windowless target are presented and the status of the heavy liquid metal target mock-up experiment at the KArlsruhe Liquid metal LAboratory (KALLA) are reported.

Novel methods of tritium production rate measurements in HCLL TBM mock-up experiment with liquid scintillation technique

Two novel methods of tritium production rate (TPR) measurements applied in HCLL TBM mock-up neutronic experiment are described and discussed. In the first method LiF TLD detectors used for gamma radiation dose measurements are applied as detectors of −β decay of tritium (T) produced by neutrons. TL signal expressed as mGy/h is proportional to energy accumulated in self-irradiation process (SI). However, the TL signal is proportional to tritium activity generated in detector material, recalculation to Bq/mg requires calibration by independent measurement of T activity in the TLD. In our experiments determination of this activity was performed by LSC technique. Good correlation between the 3H activity and TLD signal was observed. The second presented method is based on direct measurement of T in LiPb material with the use of LSC spectrometer. Dissolution of LiPb in nitric acid, and than composing aliquot containing well known mass of irradiated LiPb mixed than with scintillation cocktail, is measured in LS spectrometer providing signal proportional to the T activity. It was proved by series of experiments that procedure performed at controlled conditions ensures no loss of tritium while preparation. Optimization measurement conditions with respect of type of the scintillator used and composition of aliquot/scintillator mixture provide counting efficiency above 22%. Sample material for testing the both techniques was irradiated by neutron flux either at nuclear research reactor “Maria”, Świerk, Poland or at the Frascati Neutron Generator Laboratory in HCLL TBM neutronic experiment. Li 2CO 3 pellets used as reference were irradiated in the same experiment with TLDs, and samples of LiPb-eutectic and then analysed by modified Dierckx method, a well established technique. Measurement results obtained applying the both techniques described above as well as the control Li 2CO 3 pellets were in very good consistency. Direct measurements of T produced in LiPb at different depth in HCLL TBM mockup agrees well with MCNP simulation results. Parameters of the simulation provide additionally information on 6Li abundance and Li content in the LiPb mixture. Spread of the T activity observed at the same depth for different samples correspond with inhomogeneity of the eutectic composition investigated and reported. Direct measurement of T activity seems to be the very promising and relatively simple method of validation of neutronic codes and nuclear data capabilities for calculation TPR in fusion relevant materials provided parameters of the irradiated material are well known.

Sensitivity and uncertainty analysis of nuclear responses in the EU HCLL TBM of ITER

Within the EU Fusion Technology Programme dedicated theoretical and experimental efforts are conducted to provide reliable nuclear data and computational tools for design analyses for fusion devices like ITER including qualified uncertainty estimates. In this respect, the present paper reports on sensitivity and uncertainty analyses for the EU HCLL Test Blanket Module (TBM) of ITER. Neutron flux spectra and tritium production rates have been calculated using MCNP with a modified version of the ITER Alite torus sector model with integrated TBMs. Sensitivities of such parameters to nuclear cross sections of isotopes contained in the TBM as well as in the ITER device have been calculated using the Monte Carlo code MCSEN. Uncertainties could be obtained by using existing covariance data of the important nuclear cross section files, mainly from ENDF/B-VI, SCALE6.0, but also from recent JEFF/EFF evaluations. Like in the HCLL mock-up experiment two positions at front and back of the TBM have been selected. In both cases the calculated uncertainties of the responses (tritium production rate, neutron flux) are in the range of 2–4%.

A212 熱流動解析による高温ガス炉圧力容器温度の評価(OS8 高温ガス炉・高速炉・金属燃料)

In the passive decay heat removal system in high-temperature gas-cooled reactor (HTGR), the decay heat is transferred from the reactor pressure vessel(RPV) to cooling panel by radiation and natural convection. The thermal-hydraulic analysis for 1/8 scale mockup experiment simulating passive decay heat removal system of HTGR was performed to demonstrate the numerical calculation accuracy on temperature of RPV. The 3-dimensional 30-deg. sector model which included heater, RPV and cooling panel, was created on analytical code FLUENT v6.3. The temperature on outer face of RPV was numerically calculated within the difference of 10℃ from the experiment data with the condition of helium pressure of 0.47MPa and 0.11MPa in RPV.

Sensitivity and uncertainty analyses of the HCLL mock-up experiment

Within the European Fusion Technology Programme dedicated computational methods, tools and data have been developed and validated for sensitivity and uncertainty analyses of fusion neutronics experiments. The present paper is devoted to this kind of analyses on the recent neutronics experiment on a mock-up of the Helium-Cooled Lithium Lead Test Blanket Module for ITER at the Frascati neutron generator. They comprise both probabilistic and deterministic methodologies for the assessment of uncertainties of nuclear responses due to nuclear data uncertainties and their sensitivities to the involved reaction cross-section data. We have used MCNP and MCSEN codes in the Monte Carlo approach and DORT and SUSD3D in the deterministic approach for transport and sensitivity calculations, respectively. In both cases JEFF-3.1 and FENDL-2.1 libraries for the transport data and mainly ENDF/B-VI.8 and SCALE6.0 libraries for the relevant covariance data have been used. With a few exceptions, the two different methodological approaches were shown to provide consistent results. A total nuclear data related uncertainty in the range of 1–2% (1σ confidence level) was assessed for the tritium production in the HCLL mock-up experiment.

적합한 조명시뮬레이션 사용을 위한 주광 시뮬레이션과 Mock-up 실험 비교에 관한 연구

Available daylight in inside offers comfortable view environment, and psychological and physical advantages to people in a room. Architectural design has a need for daylight prediction. But it is difficult to predict the illuminance values in daylight system. Using various Daylight simulation software is easy and simple, but we can find different results under the same conditions. This study compares with illuminance values from simulation (Lightscape 3.2, Relux 2007, Radiance 3.9) and Mock-up experiment with a same space and parameter of daylight (location of space and window, date, time and sky condition).

Numerical modelling for the interpretation of a laboratory mock-up experiment of bentonite/granite interface

Performance assessment of a deep geological repository requires understanding diffusion and determining diffusion parameters under real conditions because diffusion is a key transport mechanism in hosting geological formation. FEBEX (Full-scale Engineered Barrier Experiment) is a demonstration and research project dealing with the bentonite engineered barrier designed for sealing and containment of the high-level radioactive waste repository. To support field investigations of FEBEX in situ test, a large–scale laboratory mock-up experiment (MUE) is being performed at CIEMAT facilities to study tracer migration at the bentonite/granite interface. Numerical models of MUE are presented here for HTO, 36Cl- and 137Cs+. Experiments are modeled with 2-D axi-symmetric finite element grids and are solved with CORE2D V4. Model results indicate that numerical solutions with reference parameters reproduce measured data for HTO and 36Cl- but show large discrepancies for 137Cs+. Relevant diffusion and retention parameters are identified by sensitivity analysis for tracer concentrations in borehole, bentonite and granite, respectively. Interpretation of 137Cs+ data measured in the tracer chamber is perfomed by taking into account the uncertainties in initial activity C0 and initial time t0. Optimum values of C0 and t0 are obtained. The best fit is obtained with De-filter equal to 2.03.10-10 m2/s and Kd-bentonite equal to 5m3.Kg-1.

Uncertainty evaluation of nuclear reaction model parameters using integral and microscopic measurements. Covariances evaluation with CONRAD code

In the [eV;MeV] energy range, modelling of the neutron induced reactions are based on nuclear reaction models having parameters. Estimation of co-variances on cross sections or on nuclear reaction model parameters is a recurrent puzzle in nuclear data evaluation. Major breakthroughs were asked by nuclear reactor physicists to assess proper uncertainties to be used in applications. In this paper, mathematical methods developped in the CONRAD code[2] will be presented to explain the treatment of all type of uncertainties, including experimental ones (statistical and systematic) and propagate them to nuclear reaction model parameters or cross sections. Marginalization procedure will thus be exposed using analytical or Monte-Carlo solutions. Furthermore, one major drawback found by reactor physicist is the fact that integral or analytical experiments (reactor mock-up or simple integral experiment, e.g. ICSBEP, …) were not taken into account sufficiently soon in the evaluation process to remove discrepancies. In this paper, we will describe a mathematical framework to take into account properly this kind of information.

Nuclear Data Benchmarking for Lead by LiAl/Pb Fusion Blanket Mock-Up Experiment

For the International Thermonuclear Experimental Reactor (ITER), the European Union intends to introduce a test blanket module using liquid lithium lead. In the present study, a preliminary experiment was conducted using a LiAl/Pb assembly, which has a neutronics performance similar to lithium lead. The neutron spectrum at the inside of the assembly was measured with an NE-213 detector. We found that the spectra calculated with MCNP5 and JENDL-3.3 underestimate the 14-MeV peak by no less than 30%, while ENDF/B-VII.0 and JEFF-3.1 gave good agreement. The neutron nuclear data of lead stored in JENDL-3.3 may have some problems.

Chemical stability of montmorillonite buffer clay under repository-like conditions—A synthesis of relevant experimental data

A current semi-empirical model of conversion of smectite to illite implies that the process is associated with formation of quartz and chlorite and will not cause significant illitization in the hydrothermal period of a repository if the activation energy is of the commonly assumed magnitude and the temperature is lower than about 100 °C. The model is supported by several natural analogues and hydrothermal experiments in laboratories and field tests. In all the experiments, which were conducted under open chemical conditions, the confined clay was exposed to thermal gradients for at least 1 year. Most of the tests were small-scale and of “1D-type”, one being performed with strong gamma radiation, and the others with different porewater compositions. One was a Mock-up experiment simulating the KBS-3V concept on 50% scale, and one a full-scale KBS-3V field experiment at 400 m depth in granitic rock. They all showed that while smectite remained the major clay mineral a number of processes changed the physical properties in bulk, the most important changes being reduction of the expandability and increase of the hydraulic conductivity of the hottest part of the buffer. The latter is concluded to have been caused by contraction of clay aggregates that became permanent by precipitation of silica and iron compounds, the cementing agents emanating from dissolved accessory minerals as well as from montmorillonite. Fe-rich montmorillonite underwent particularly important dissolution.

Blanket mock-up experiment with a LiAlPb assembly irradiated with 14 MeV neutrons in preparation of the HCLL-TBM neutronics experiment

Irradiation experiments with a LiAlPb assembly have been performed using the DT neutron generator of Technical University of Dresden. The main goal was to qualify LiF thermoluminescence detectors (TLDs) for tritium breeding experiments and to test a recently improved NE-213 spectrometer, both for application in a neutronics mock-up experiment of the European HCLL-TBM at ENEA Frascati and Technical University of Dresden (TUD) planned for 2008. The LiF-TLD were set up inside the LiAlPb assembly and irradiated with the DT neutron generator of TUD in continuous mode. A good agreement with TPR calculated with MCNP was found. Fast neutron spectra inside the assembly have been obtained with the NE-213 detector. The flux of the thermal and epithermal neutrons inside the assembly, which is important for the tritium production rate from Li 6 , was measured with a He 3 counter and the DT neutron generator operated in pulsed mode.

Impact of reflected neutrons on accuracy of tritium production rate prediction in blanket mock-ups for fusion reactors

A series of breeding blanket neutronics experiments were carried out with DT neutrons in the Japanese ITER test blanket module at JAEA Fusion Neutronics Source (FNS), and tritium production rates (TPRs) inside different blanket mock-ups were measured in detail. Most of calculations made using FENDL-2.1 and MCNP agreed well with the measured TPRs. On the other hand, there were overestimations of the calculation for the TPR in the experiment with a reflector and the TPR around the boundary between the rear part of the breeder layer and the beryllium in the pebble bed mock-up experiment. In this study, we investigated the possibility of improvement of the calculation by modifying the part of angular distributions in backward directions in FENDL-2.1. By uniformly decreasing the backward elastic scattering cross-sections of 56Fe by 20% below the original data for incoming neutrons in the energy range from 0.5234 MeV up to 15.0 MeV, the overestimation of the calculations for the TPR in the enriched breeding blanket experiment with a reflector was reduced. Also, by uniformly decreasing the backward elastic scattering cross-sections of 9Be by 20% and 30% below the original data for the incoming neutrons in the energy range from 0.93 MeV up to 14.94 MeV, and by 20% for those from 0.62 MeV up to 14.94 MeV, the overestimation of the calculations for the TPR around the boundary between the rear part of the breeder layer and the beryllium was lessened.

Advanced neutronics tools for BWR design calculations

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific “2005–2007 BWR Plan” settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy.

Waste Package Corrosion Studies Using Small Mockup Experiments

AbstractUnderstanding the corrosion of spent nuclear fuel (SNF) and the subsequent mobilization of released radionuclides, particularly under oxidizing conditions, is one of the key issues in evaluating the long-term performance of a nuclear waste repository. However, the large amounts of iron in the metal waste package may create locally reducing conditions that would lower corrosion rates for the SNF, as well as reduce the solubility of some key radionuclides, e.g., Tc and Np. In order to investigate the interactions among SNF-waste package-fluids, four smallscale (∼1:40 by length) waste package mockups were constructed using metals similar to those proposed for use in waste packages at the proposed repository at Yucca Mountain. Each mockup experiment differed with respect to water input, exposure to the atmosphere, and temperature. Simulated Yucca Mountain process water (YMPW) was injected into three of the mockups at a rate of 200 μL per day for five days a week using a calibrated needle syringe. The YMPW was prepared by equilibrating 50 mg/L silica as sodium metasilicate with air, and adding enough HCl to lower the pH to 7.6 in contact with an excess of powdered calcite.X-ray powder diffraction and scanning electron microscopy confirm that, where corrosion occurred, the dominant corrosion product in all cases was magnetite. In the high temperature (60°C) experiment, hematite and a fibrous, Fe-O-Cl phase were also identified. The Fe(II)/Fe(III) ratios measured in the corrosion products using a wet chemistry technique indicate extremely low oxygen fugacities (10-36 bar). Experiments are in progress in which 0.1g powdered UO2 was included in the mock-up in order to investigate the relative kinetics of Fe and U oxidation and to identify the U corrosion products formed under these conditions.

Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL) concept will be performed in 2008 in the Frascati Neutron Generator (FNG) in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n) and (n,3n) reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

Sensitivity and uncertainty analyses of the tritium production in the HCPB breeder blanket mock-up experiment

A neutronics experiment performed at the Frascati Neutron Generator (FNG) on a mock-up of the Helium-Cooled Pebble Bed (HCPB) breeder test blanket module (TBM) has been analysed on the basis of neutron transport, sensitivity and uncertainty calculations using both deterministic and probabilistic computational methods. The calculations revealed a slight but systematic underestimation of the measured tritium activities by 5 to 10% on average, independent of the nuclear data used that were taken from the EFF-3 and FENDL-2.0, -2.1 data files. The uncertainty assessments showed data related uncertainties of the calculated tritium production rates at a level of 4% (2 σ confidence level). The beryllium cross-section data were shown to be the main source of these uncertainties. The total uncertainties of the tritium production prediction including the data uncertainties, the statistical uncertainties of the Monte Carlo calculation and the experimental uncertainties are in the order of 8 to 10% (2 σ). The observed underestimation of the measured tritium production is at the lower bound of the assessed uncertainty margin. The obtained results indicate that design calculations for the tritium breeding ratio (TBR) of fusion power reactors employing a HCPB type breeder blanket are conservative. Thus, an additional TBR margin is provided which allows compensation for potential other uncertainties.

Feasibility Study for Evaluation of Control Rod Worth in Pressurized Water Reactors using Neutron Count Rate during a Control Rod Drop Testing

Applicability of the modified Neutron Source Multiplication (NSM) method with extraction of the fundamental mode to subcriticality measurement has been proposed. Following the feasibility verification in the previous study based on numerical analyses, its applicability has been proven in a more realistic situation; in a withdrawal sequence of control rod banks during the PWR startup. Subcriticalities with various control rod insertion configurations were estimated based on the modified NSM method. The subcriticality could be evaluated with a good accuracy even with the mockup experiment where any special treatments for accurate measurement were not taken into account and furthermore the insensitivity of measured signals by reactivity changes and their large fluctuations were seen. Based on this fact, we further investigated a feasibility to use neutron count rate data obtained during the control rod drop testing, which is carried out before the reactor physics tests at hot zero power condition. When it is proven that these data could be used for the estimation of each control rod worth, the following reactor physics tests could be performed with the advanced knowledge of each control rod worth and procedures for detailed control rod worth measurement could be simplified or eliminated from the reactor physics tests.

Plasma diagnostic system using optical fibers with high numerical aperture

A diagnostic system using optical fibers with high numerical apertures is constructed for investigating a structure of magnetically confined plasma. Since the fiber has no convex lens at its front facet, the radiation from all plasmas included in the viewing angle of the fiber is utilized. First, theoretical forms are derived to determine the structure of the plasma from optical signals of a few fibers arranged axisymmetrically around the plasma. Second, the numerical aperture is measured by a simple method using a circular fluorescent lamp. Third, the theoretical forms are ascertained by a mock-up experiment where a light source representing a model plasma is constructed by a matrix array of straight fluorescent lamps. In this experiment, the effect of a reflected light from a wall of a vacuum vessel is investigated. Finally, the optical system is applied to a field-reversed-configuration plasma. Obtained structures of the plasma for n=1 and 2 toroidal modes are compared with the results of the conventional method using the fibers with convex lenses. The former results agree well with the latter ones in spite of the use of one-sixth the number of fibers in the latter method.