Modeling, analysis and prediction of neutron emission spectra from acoustic cavitation bubble fusion experiments

Self-nucleated and external neutron nucleated acoustic (bubble fusion) cavitation experiments have been modeled and analyzed for neutron spectral characteristics at the detector locations for all separate successful published bubble fusion studies. Our predictive approach was first calibrated and validated against the measured neutron spectrum emitted from a spontaneous fission source (²⁵²Cf), from a Pu–Be source and from an accelerator-based monoenergetic 14.1 MeV neutrons, respectively. Three-dimensional Monte-Carlo neutron transport calculations of 2.45 MeV neutrons from imploding bubbles were conducted, using the well-known MCNP5 transport code, for the published original experimental studies of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109; Taleyarkhan, et al., 2006a. PRL 96, 034301; Taleyarkhan, et al., 2006b. PRL 97, 149404] as also the successful confirmation studies of Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324], Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and Bugg [Bugg, W., 2006. Report on Activities on June 2006 Visit, Report to Purdue University, June 9, 2006]. NE-213 liquid scintillation (LS) detector response was calculated using the SCINFUL code. These were cross-checked using a separate independent approach involving weighting and convoluting MCNP5 predictions with published experimentally measured NE-213 detector neutron response curves for monoenergetic neutrons at various energies. The impact of neutron pulse-pileup during bubble fusion was verified and estimated with pulsed neutron generator based experiments and first-principle calculations. Results of modeling-cum-experimentation were found to be consistent with published experimentally observed neutron spectra for 2.45 MeV neutron emissions during acoustic cavitation (bubble) fusion experimental conditions with and without ice-pack (thermal) shielding. Calculated neutron spectra with the inclusion of ice-pack shielding are consistent with the published spectra from experiments of Taleyarkhan et al. [Taleyarkhan, et al., 2006a. PRL 96, 034301] and Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324] where ice-pack shielding was present, whereas without ice-pack shielding the calculated neutron spectrum is consistent with the experimentally observed neutron spectra of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109] and Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and also that from GEANT computer code [Agostinelli, S., et al., 2003. Nuclear Instrum. Methods Phys. Res. A 506, 250–303] predictions [Naranjo, B., 2006. PRL 97 (October), 149403] in which ice shielding was also absent.The results of this archive confirm for the record that the confusion and controversies caused from past reports [Reich, E., 2006. Nature (March) 060306. news@nature.com; Naranjo, B., 2006. PRL, 97 (October) 149403] have resulted from their neglect of important details of bubble fusion experiments. Results from this paper demonstrate that ice-pack shielding between the detector and the fusion neutron source, gamma photon leakage and neutron pulse-pileup due to picosecond duration neutron pulse emission effects play important roles in affecting the spectra of neutrons from acoustic inertial confinement thermonuclear fusion experiments.     

An assessment of acoustic emission for nuclear pressure vessel monitoring

Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques.

Published data suggest that AE can make an important contribution to weld fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise.

It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment.     

Deuterated target comparison for pyroelectric crystal D-D nuclear fusion experiments

Different target materials were investigated to determine which ones are favorable to increasing the theoretical neutron yield using pyroelectric crystal D-D nuclear fusion. Calculations show that deuterated polyethylene (CD₂) will potentially yield the highest number of neutrons compared to the other targets investigated. However, deuterated plastic targets have been found to erode over the course of experiments.     

Direct containment heating experiments for Vandellos and ASCo nuclear power plants

The ongoing IPE studies for the Vandellos and ASCo nuclear power plants require evaluation of accident phenomena that have been perceived to potentially challenge containment integrity including direct containment heating (DCH). Analyses and scaled experiments performed to date indicated that the lower containment structures play a substantial role in mitigating the extent of DCH given a high pressure melt ejection. Since the geometry is judged to be of major importance, linearly scaled experiments were conceived and conducted to evaluate the role of such structures in the Vandellos and ASCo specific configurations. The Vandellos test configuration with an initally dry cavity and significant exhaust area for the instrument tunnel resulted in the dispersal of a majority of the debris from the instrument tunnel into the lower compartment. The test of the ASCo configuration with an initially wet reactor cavity and limited exhaust area from the instrument tunnel exhibited the retention of the majority of the debris within the instrument tunnel and reactor cavity. The observed pressure responses in these scaled experiments for the seal table room, lower containment vessel, and upper containment vessel were all less than the containment design basis pressure. These test results contribute to the existing technical basis for concluding that direct containment heating would not represent a challenge to the integrity of these containments.     

核物理实验中的慢变化信号自动记录仪

介绍了一种由单片机控制,用于核物理实验中对慢变化信号进行自动监测的数字记录仪,并对其工作原理,硬件和软件设计性能指标作了较详细的阐述。     

A fast active linear gate for experiments in nuclear and particle physics

This paper presents an active linear gate that is able to transmit fast pulses with minimum distortion. The gate exhibits switching times better than 3 ns, pedestal, and a feedthrough of 0.3%. In the field of experimental high-energy physics, the gate was successfully used in measuring the speed response of a fast scintillating calorimeter.     

基于微机的多参数核物理实验数据获取分析系统

介绍一台由微机控制的用CAMAC插件组成的多参数数据获取和离线数据分析系统。该系统采用软件控制的方法进行信号种类的甄别以及划定信号收集的范围。与用NIM系统构成的数据获取系统相比,在达到同样测量精度的基础上,降低了对硬件设备的要求。该系统再配合微型电离室、芪晶体探测器,被用来做裂变中子飞行时间谱的测量,给出了测量结果。     

Irradiation experiments in the testing nuclear power plant VAK

Within the German Research Programme “Integrity of Components” the first two capsules were irradiated in the Testing Nuclear Power Reactor VAK. The materials are of the 22 NiMoCr 3 7 and 20 MnMoNi 5 5 types and represent the lower bound of the base material regarding upper shelf energy and chemical composition (Cu, S, P), as well as a state of material which does not meet both chemical and toughness requirements (low upper shelf test melt). Tensile, Charpy, drop-weight, and fracture mechanics specimens were irradiated up to a range of 1.5 to 2 × 10¹⁹ cm⁻² (E > 1 MeV). Despite the materials being at or beyond the specification limits, the results show irradiation sensitivity which can be predicted from the US Reg. Guide Trend Curves (1.99) and KWU Trend Curves in a conservative manner. The procedure to determine the adjusted reference temperature RT_NDT (adj.) on the basis of ΔT_41J (following ASTM E 185) could also be confirmed as conservative by comparing the different criteria derived from Charpy and drop weight tests in the unirradiated and irradiated condition.The results of fracture mechanics testing in the linear elastic range show a remarkable temperature margin to the K_Ic-curve of ASME XI.Prestrained compact tension specimens CT 40 mm made of 22 NiMoCr 3 7 material with an upper shelf energy of approx. 100 J were wedge loaded in a range up to 30 MPa m and exposed to the water environment during radiation. Macroscopic examination gave no indications of stress corrosion cracking.From tests of these specimens in the linear elastic range, a fracture toughness K_Ic^*, which was not affected by the prestrain and environment history, was found depending only on the overload applied during the prestraining procedure.     

核子皮带秤物理模拟实验

用皮带运输物料煤进行的核子皮带秤物理模拟实验测量了核子秤的线性、重复性和长期稳定性。初步研究了煤的堆积形状、偏载、成分含量变化对核子秤计量的影响,为核子皮带秤的安装、调试和运行提供了物理依据。     

Application of remote gamma spectrometry in automated systems for monitoring emissions from nuclear power plants

Methodological questions concerning the application of remote gamma spectrometry for radiation monitoring of air emissions from nuclear power plants and detecting trans-boundary transport of products of an accident are examined for the example of the development and experimental operation of the components of the Spaider data-acquisition system produced by the KonverSia company. A system of collimated γ spectrometers arranged along the monitoring contour at the site boundary of a nuclear power plant makes it possible to increase the accuracy of the computational predictions and to obtain and accumulate, during normal operation of a nuclear power plant, emissions data for radioactive inert gases and iodine in a continual monitoring regime. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 35–42, January, 2009.     

Utilizing PUNITA experiments to evaluate fundamental delayed gamma-ray spectroscopy interrogation requirements for nuclear safeguards

ABSTRACT

Present safeguards verification methods for small samples of high-radioactivity nuclear material (e.g. spent nuclear fuel solution) in reprocessing facilities use destructive analysis techniques since passive non-destructive techniques are incapable of directly determining the nuclear material content. To supplement these methods, the Japan Atomic Energy Agency and European Commission Joint Research Centre (JRC) are collaborating to develop delayed gamma-ray spectroscopy non-destructive assay technology for composition analysis of the fission nuclides. Multiple experiments were performed in the JRC-Ispra site using the Pulsed Neutron Interrogation Test Assembly (PUNITA) to study the signature from short-lived fission products from low-radioactivity U and Pu standard samples. From these spectra, we identify many gamma rays useful to determine the composition of a mixed nuclear material sample. Here we present the results of these experiments along with correlations to the interrogation, mass, volume, and sample homogeneity.     

Preliminary spectrum shifter design for intermediate energy nuclear data benchmark experiments with DT neutrons

Integral benchmark experiments with DT neutrons are not always sufficient for nuclear data benchmarking in the MeV region, below 10 MeV. A neutron spectrum shifter, which will be placed between a sample and a DT neutron source, is effective to moderate DT neutrons incident to the sample. In order to estimate effects of the spectrum shifter, the ratio of the contribution of 14 MeV neutrons in the leakage neutron and gamma-ray spectra was calculated with MCNP-4C for an experimental configuration at FNS of JAEA, Japan. The calculations were carried out for a Li₂TiO₃ sample with a Be, D₂O, or ⁷LiD spectrum shifter. It was found out that the Be shifter was superior to others and the Be shifter was effective to decrease the contribution of 14 MeV neutrons especially for secondary gamma-ray spectrum measurements.     

Importance of the MUSE experiments for emerging ADS concepts from the nuclear data viewpoint

The current investigation, conducted in the general framework of the MUSE program (“MUltiplication avec une Source Externe”), considers the representativity of a specific configuration of its fourth phase (M4SC2), which is driven by an external D(d,n)He³ or T(d,n)He⁴ neutron source, with respect to current concepts of eXperimental Accelerator Driven Systems (XADSs) with gas (He), Na and Pb/Bi coolants. The study has been carried out from the nuclear data viewpoint, with the external source being accounted for in an appropriate manner. In this context, data sensitivity/uncertainty analyses based on first-order perturbation theory calculations have been performed using the deterministic code ERANOS (Version 2.0) in conjunction with its adjusted nuclear data library ERALIB-1.It is found that the M4SC2 configuration, independent of the external source, is quite representative of the different XADSs for actinide capture reactions at the centre of the fuel zone, relative to ²³⁹Pu fission at the same location. For the case of a threshold fission reaction, such as that in ²³⁸U, the sensitivity to the external source is significantly higher. With respect to the corresponding spectral index, M4SC2 with the D(d,n)He³ source remains quite representative of the He- and Na-cooled XADSs. For the system with Pb/Bi coolant, on the other hand, effects of uncertainties associated with the data for these two nuclides and their low content in the MUSE configuration result in significantly lower associated representativity factors. A better overall representativity of the Pb/Bi-cooled XADS is expected to be achieved by the new MUSE_Na/Pb configuration.     

ATCA/AXIe compatible board for fast control and data acquisition in nuclear fusion experiments

An in-house development of an Advanced Telecommunications Computing Architecture (ATCA) board for fast control and data acquisition, with Input/Output (IO) processing capability, is presented. The architecture, compatible with the ATCA (PICMG 3.4) and ATCA eXtensions for Instrumentation (AXIe) specifications, comprises a passive Rear Transition Module (RTM) for IO connectivity to ease hot-swap maintenance and simultaneously to increase cabling life cycle.The board complies with ITER Fast Plant System Controller (FPSC) guidelines for rear IO connectivity and redundancy, in order to provide high levels of reliability and availability to the control and data acquisition systems of nuclear fusion devices with long duration plasma discharges.Simultaneously digitized data from all Analog to Digital Converters (ADC) of the board can be filtered/decimated in a Field Programmable Gate Array (FPGA), decreasing data throughput, increasing resolution, and sent through Peripheral Component Interconnect (PCI) Express to multi-core processors in the ATCA shelf hub slots. Concurrently the multi-core processors can update the board Digital to Analog Converters (DAC) in real-time. Full-duplex point-to-point communication links between all FPGAs, of peer boards inside the shelf, allow the implementation of distributed algorithms and Multi-Input Multi-Output (MIMO) systems. Support for several timing and synchronization solutions is also provided.Some key features are onboard ADC or DAC modules with galvanic isolation, Xilinx Virtex 6 FPGA, standard Dual Data Rate (DDR) 3 SODIMM memory, standard CompactFLASH memory card, Intelligent Platform Management Controller (IPMC), two PCI Express x4 (generation 2) ATCA Fabric channels (dual-star topology), eleven Xilinx Aurora x1 (or other ATCA compatible communications protocol) ATCA fabric channels (full-mesh topology) and two Fast Ethernet (Precision Time Protocol – PTP IEEE1588-V2 and Lan eXtensions for Instrumentation – LXI compatible) ATCA base channels (dual-star topology).     

Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

We report the measurement of elastic constants of non-irradiated UO₂, SIMFUEL (simulated spent fuel: UO₂ with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young’s modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.     

Tomography of nuclear fuel experiments with an electronic line scan camera

A line scan camera has been developed, which may be applied for tomography in the non-destructive examination of nuclear fuel experiments. The camera has been tested with different radiation sources like 400 kV X-ray tubes, an 18 mV betatron and a thermal neutron beam. The camera uses silicon photo-diode-arrays with an integrated shielding suppressing most of the radiation coming from elsewhere than from the source. Tomograms of preirradiated fuel pins were generated showing the fuel debris distribution after an over-power test. Various phantoms have been tomographed, which demonstrate the feasibility of imaging cross sections of fuel bundles nondestructively. The system is able to measure density distributions in a 0.3 mm thick cross section with about 0.3 mm spatial resolution. The camera itself is small and the control computer of standard industry type. Thus the tomography system is transportable and may be integrated in existing nondestructive facilities.     

Photon production through multi-step processes important in nuclear resonance fluorescence experiments

We present calculations describing the production of photons through multi-step processes occurring when a beam of gamma-rays interacts with a macroscopic material. These processes involve the creation of energetic electrons through Compton scattering, photo-absorption and pair production, the subsequent scattering of these electrons, and the creation of energetic photons occurring as these electrons are slowed through Bremsstrahlung emission. Unlike single Compton collisions, during which an energetic photon that is scattered through a large angle loses most of its energy, these multi-step processes result in a sizable flux of energetic photons traveling at large angles relative to an incident photon beam. These multi-step processes are also a key background in experiments that measure nuclear resonance fluorescence by shining photons on a thin foil and observing the spectrum of back-scattered photons. Effective cross sections describing the production of back-scattered photons are presented in a tabular form that allows simple estimates of backgrounds expected in a variety of experiments. Incident photons with energies between 0.5 MeV and 8 MeV are considered. These calculations of effective cross sections may be useful for those designing NRF experiments or systems that detect specific isotopes in well-shielded environments through observation of resonance fluorescence.