Personal neutron dosimetry at a research reactor facility

Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve.     

Measurement of the near infrared xenon scintillation yield near atmospheric pressure

We have measured the xenon near infrared (NIR) scintillation yield in the wavelength range from about 700 to 850 nm for gas pressures from 0.3 to 1.5 atm. We have found the NIR yield to be nearly constant above 0.9 atm, increasing by about a factor of 4.5 as pressure is lowered to 0.3 atm. Measured in number of photons emitted we estimate that at 0.3 atm the NIR yield is 45% of the ultraviolet yield near 170 nm. The NIR yield may be sufficient to form the basis of detector systems utilizing low-pressure xenon gas.     

A causal model of productivity in a research facility

The problem addressed concerns the conditions that foster productivity among natural scientists in a large research laboratory. We take several variables identified as important in two major perspectives in the literature on productivity, and use these variables to construct a causal model. Using path analysis, we test the model by employing data from a sample of 295 scientists working at an atomic research facility in West Germany. In general, educational level of the scientists has an important, positive impact on productivity; years of service has a varying and more modest positive effect. Rank of the scientist has an intermediate positive impact on productivity; psychological factors have a negligible effect. Finally, the influence the scientist has on his research endeavors has a modest positive impact on productivity.I am happy to acknowledge several people who provided considerable help on this research: Professors Edward O.Laumann, of the University of Chicago, Franz U.Pappi, of the Universitat zu Kiel, and Michael S.Flynn and Mary FrankFox, of the University of Michigan, and a number of research assistants, including in particular HaraldKlingemann and E. DanielAyres. Financial support for the data collection was freely provided by the Landesamt fur Forschung in Ministerium für Wissenschaft and Forschung des Landes Nordrhein-Westfalen. Financial support for the data analysis came from the National Science Foundation (GS-32002). Data collection was conducted under the auspices of the Zentralarchiv der Universität zu Köln.     

Monte Carlo optimisation of a BNCT facility for treating brain gliomas at the TAPIRO reactor

An epithermal boron neutron capture therapy facility for treating brain gliomas is currently under construction at the 5 kW fast-flux reactor TAPIRO located at ENEA, Casaccia, near Rome. In this work, the sensitivity of the results to the boron concentrations in healthy tissue and tumour is investigated and the change in beam quality on modifying the moderator thickness (within design limits) is studied. The Monte Carlo codes MCNP and MCNPX were used together with the DSA in-house variance reduction patch. Both usual free beam parameters and the in-phantom treatment planning figures-of-merit have been calculated in a realistic anthropomorphic phantom ('ADAM').     

Problems of waste chemical handling at a large biomedical research facility

Present handling and disposal methods for the chemical wastes arising at a large U.S. biomedical laboratory complex are described, with emphasis on the operational and management problems posed by these wastes. A study of incinerators which may be suitable for dealing with a large proportion of the wastes is described, which has led to a more detailed investigation of one of the most promising units. The effects of recent legislation are indicated. It is concluded that improvements are desirable, and these are now being implemented.     

Influence of specific data on a research reactor probabilistic model

Deterministic safety calculations are usually required and included in the Safety Analysis Report of research reactors. To estimate the risk of a research reactor, Probabilistic Safety Assessment (PSA) is rarely used. In this paper, a PSA of a TRIGA Mark II research reactor with generic and specific data is described. The results are discussed to show the need for PSA and the usefulness of specific examination of the research reactor.Beside the deterministic calculations, PSA has proved to be a powerful tool for safety evaluation of the research reactor. It is recommended that as much specific data is used as is possible for initiating event definitions and frequencies estimation. We do not recommend the building of an extensive data base for components. When safety of non-standard technologies is estimated it is recommended that a preliminary PSA is carried out first. It enables definition of needed specific data to be collected and contributes to better employment of human resources.     

A comparison of different neutron spectroscopy systems at the reactor facility VENUS

The VENUS facility is a zero-power research reactor mainly devoted to studies on LWR fuels. Localised high-neutron rates were found around the reactor, with a neutron/gamma dose equivalent rate ratio as high as three. Therefore, a study of the neutron dosimetry around the reactor was started some years ago. During this study, several methods of neutron spectroscopy were employed and a study of individual and ambient dosemeters was performed.A first spectrometric measurement was done with the IPSN multisphere spectrometer in three positions around the reactor. Secondly, the ROSPEC spectrometer from the Fraunhofer Institut was used. The spectra were also measured with the bubble interactive neutron spectrometer. These measurements were compared with a numerical simulation of the neutron field made with the code TRIPOLI-3. Dosimetric measurements were made with three types of personal neutron dosemeters: an albedo type, a track etch detector and a bubble detector.     

Results of field trials using the NPL simulated reactor neutron field facility

The NPL simulated reactor neutron field facility provides neutron spectra similar to those found in the environs of UK gas-cooled reactors. Neutrons are generated by irradiating a thick lithium-alloy target with monoenergetic protons between 2.5 and 3.5 MeV (depending on the desired spectrum), and then moderated by a 40-cm diameter sphere of heavy water. This represents an extremely soft workplace field, with a mean neutron energy of 25 keV and, more significantly, a mean fluence to ambient dose equivalent conversion coefficient of the order of 20 pSv cm(2), approximately 20 times lower than those of the ISO standard calibration sources (252)Cf and (241)Am-Be. Results of field trials are presented, including readings from neutron spectrometers, personal dosimeters (active and passive) and neutron area survey meters, and issues with beam monitoring are discussed.     

Prototype development of a battery power supply at theelectromagnetic launcher research facility

Electromagnetic launcher (EML) research at Elgin Air Force Base has progressed beyond available power source levels. Therefore, an inexpensive, high-power, high-energy battery system is under construction to fulfil present and future requirements. Data on the subscale battery power supply (BPS) system are examined. Decisions made on the design and operation of pneumatic switches, batteries, and contactors as a result of this testing program are described. Discharges to date have proved that the gang concept of operation is successful at current levels significantly higher than the standard operational level of 50 kA to 60 kA expected in the full BPS system. The electrical integrity of the prototype system has also been proved at energy levels of approximately 45 MJ, which is 15 to 20 MJ greater than energy levels that will be exhibited at the gang level in the final system     

GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities.In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]).The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10¹¹ ncm²/s.This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted.The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line (http://www.info.cern.ch/asd/geant4/geant4.html[4]).To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al₂O₃) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al₂O₃) filters, respectively.To get a good cadmium ratio, GEANT 4 simulations were used to define the design of the moderator in the inlet of the radiation channel. A graphite block of 22 cm thickness seems to be the optimal neutron moderator.The results showed that the combination of 5 cm of bismuth with 5 cm of sapphire permits the filtration of gamma-rays, epithermal neutrons as well as fast neutrons in a considerable way without affecting the neutron thermal flux.     

Risk implications of digital reactor protection system operating experience

This paper summarizes an in-depth review of the US nuclear operating experience with the first generation of digital reactor protection systems. The accumulated operating experience from 1984 to 2006 on these first generation digital reactor protection system functions exceeds 1.27 million hours (145.5 yr). A review of failure event reports identified 141 specific events associated with these systems on seven US nuclear power plants. Twenty-six of these events involved some type of common cause failure mechanism (predominantly redundant sensors/channels being out of calibration), which temporarily rendered redundant portions of the overall trip function degraded. Most of these failures were found not to be unique to digital systems. Six of the common cause failure events were more severe and involved situations where incorrect addressable constant data sets were systematically loaded into all redundant computer channels due to personnel errors. One of these events involved a latent software design change error introduced during a software update, which would prevent proper operation, given an unlikely event involving failure of three out of four sensors of one type.Based upon this review of digital system operating experience, a series of risk assessment calculations were performed to evaluate the safety significance of the observed failure events. From the insights gained in this work, it is possible to develop a framework for establishing digital reactor protection system reliability requirements that can be related back to regulatory safety goal objectives and operating experience.     

Broadening of the particle resonance drift trajectory and tritium injection into a fusion reactor with an I=3 helical winding

Two fusion reactor problems, removal of helium ash and fuel (tritium) injection, can be solved using the concept of “drift island motion.” The motion of a drift island is an indicator of the broadening of the resonant trajectory of a charged particle guiding center. This trajectory broadening occurs if two conditions are satisfied. First, the drift pitch angle of the particle is equal to the resonance values i*=n/m, where n and m are the “wave numbers” of the perturbing magnetic field. Second, the drift pitch angle i*=n/m “moves” over the plasma cross section as the particle moves. This displacement is caused by a slow change in the helical magnetic field with time as the particle moves. It is shown that this effect may occur in a fusion reactor with an l=3 helical winding and may be used for tritium ion injection. Pis’ma Zh. Tekh. Fiz. 25, 5–13 (January 26, 1999)     

Effect of eddy formation on gas mixture homogeneity near a substrate in a vertical reactor for CdHgTe deposition

This paper presents simulation results obtained for an unsteady-state flow of a multicomponent gas mixture in a flow-type vertical reactor using a three-dimensional mathematical model that takes into account the convective mass and heat transport, compressibility, viscosity, and thermal conductivity of the components and gas flow turbulization. We identify the gas dynamic mechanism behind eddy formation in the reactor and examine the influence of the viscosity, density, and speed of the flow components on eddy formation and the homogeneity of the gas mixture over the substrate.     

Investigation of monitoring for internal exposure by urine bioassay in a biomedical research facility

A simple monitoring programme by urine bioassay was carried out to assess internal doses for workers in a biomedical research facility. Urine samples were measured without chemical treatment using a liquid scintillation counter or sodium iodide (NaI (Tl)) scintillation detector. The detection limits for the committed effective doses were below 1 mSv for 125I and 131I and below 1 x 10(-1) mSv for 3H, 14C, 32P, 35S and 51Cr, in the case that samples were collected within 3 d after handling. All of the urinary concentrations were below the detection limit, suggesting that no significant intake was detected during the investigation. The present results suggest that personal monitoring, such as the urine bioassay, is not necessary in many cases as a routine monitoring for workers engaged in tracer experiments using the above nuclides.     

Study of irradiation effects at the research reactor

We discuss irradiation tests performed at the Budapest research reactor operating with 10 MW using a new irradiation rig. Some examples of the use of the irradiation rigs are presented: results of reactor pressure vessel cladding, 9% Cr steels (Euroferr), application of Master Curve to irradiated reactor pressure vessel steels, as well as irradiation tests of Ti-alloys and tungsten.     

Conversion of biogas like mixtures to C2 hydrocarbon in a plug flow reactor supported by a DBD at atmospheric pressure

The conversion of biogas like mixtures of methane and carbon dioxide was studied in a plug flow reactor with a dielectric barrier discharge. A 13.56 MHz power supply generated the atmospheric plasma discharge. Studied concentration of methane ranged from 0 to 100%, with missing part filled up with carbon dioxide. This mixture was diluted with helium to 2.5% with small part of the product stream monitored online at a total pressure of 100 mbar by a Fourier transform infrared spectrometer supported by a White-cell and a quadrupole mass spectrometer. The reactor was driven at different flow rates. This DBD reactor produces all three hydrocarbons, with ethane being the major compound. The concentration of ethane increases when the power in the plasma region increases from 30 to 65 W. This product concentration also grows up, if the fraction of methane in the inlet flow is increased. The highest amount of ethane (5.4%) is produced, when the gas stream consist of 2.5% Methane and 97.5% helium at a flow rate of 200 sccm.     

Single-row equidistant facility layout as a special case of single-row facility layout

In this paper we discuss two particular layout problems, namely the Single-Row Equidistant Facility Layout Problem (SREFLP) and the Single-Row Facility Layout Problem (SRFLP). Our aim is to consolidate the two respective branches in the layout literature. We show that the SREFLP is not only a special case of the Quadratic Assignment Problem but also a special case of the SRFLP. This new connection is relevant as the strongest exact methods for the SRFLP outperform the best approaches specialised to the SREFLP. We describe and compare the exact approaches for the SRFLP, the SREFLP and Linear Arrangement that is again a special case of the SREFLP. In a computational study we showcase that the strongest exact approach for the SRFLP clearly outperforms the strongest exact approach tailored to the SREFLP on medium and large benchmark instances from the literature.     

Mass transfer from a finite strip near an oscillating stagnation point – implications for atherogenesis

The mass transfer from a finite-length strip near a two-dimensional, oscillating stagnation-point flow in an incompressible, Newtonian fluid is considered. The problem is investigated using a combination of asymptotic and numerical methods. The aim of the study is to determine the effect of the location of the strip, relative to the time-averaged position of the stagnation point, on the mass transfer into the fluid. The study is motivated by the problem of mass transfer from an injured region of the arterial wall into the blood, a process that may be of considerable importance in atherogenesis. For physiologically realistic parameter values, it is found that the fluid flow is quasi-steady, but the mass transfer exhibits genuine time-dependence and a high-frequency asymptotic solution provides an accurate prediction of the time-average mass transfer. In this regime, there is a significant reduction in mass transfer when the centre of the strip is located at the point of zero time-averaged wall shear rate, or equivalently wall shear stress, which may serve to explain, at least partially, the correlation between arterial disease and regions of low wall shear stress.     

Some problems of emergency cooling of the thermal research reactor

The cooling regime of a reactor in emergency shutdown of the circulating pumps is studied on a type MN-7 electronic model and on a model with electrical heating of a fuel element assembly.