bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of a Jet Fuel Thermal Stability Flow Device which Employs Direct Gravimetric Analysis of Both Surface and Fuel Insoluble Deposits
Authors: Hardy DR , Beal EJ , Burnett JC
Published: March 2009

This paper describes the development of a liquid flow reactor which is capable of emulating the JFTOT apparatus conditions but which is constructed of simple off-the-shelf components. This new gravimetric fuel thermal oxidation tester allows the determination of deposit weights directly, simply and non-destructively by gravimetry. This new device is quite compact and offers vastly improved flexibility of operating conditions including a wide range of test temperatures, pressures, flows and reactant types. It is ideally constructed for the attachment of ancillary analytical devices. The results of tests with this device on 15 specification jet fuels at standard ASTM D3241 (Jet Fuel Thermal Oxidation Tester Procedure) conditions are presented. The gravimetric tester yielded quantitative and repeatable measurements of both heater strip deposits and fuel insoluble concentrations.
Electrical Resistance of Wires Used as a Corrosion Rate Monitor
Authors: Sproles ES
Published: March 2009

As a wire corrodes, its electrical resistance increases due to a loss of metallic cross-sectional area. By continuously monitoring resistance, corrosion can be monitored in the ambient where the wire is exposed. Although simple in principle, this technique has a number of practical problems when applied to atmospheric corrosion tests used with electronic equipment. These problems include: (1) the need to use very small diameter wires which are quite fragile; (2) the need to clean the wire in a reproducible and effective initialization process; and (3) the need to hold the wire in a resistance-measuring fixture without using solder and solder flux. This paper describes methods developed to overcome these problems and to implement the technique in a practical way for use as corrosion rate monitors for mixed gas atmospheric corrosion tests. Data are shown to illustrate the use of the method and to show the degree of reproducibility obtained between separate monitoring wires. The application of the method to fundamental studies of the effect of shielding on atmospheric corrosion is also illustrated.
The Effect of Conversion Coated and Plated Components on the Corrosion of Cobalt Alloy Magnetic Disks
Authors: Goodson K , Cormia R
Published: March 2009

Common disk drive components with chromate conversion coatings or electroless nickel platings were analyzed by ion chromatography for residual levels of Cl-, F-, SO42-, and NO3-. Extraction with boiling DI water revealed levels up to several hundred ng/cm2 for these anions. When these components were placed 2 mm under a suspended cobalt-nickel-phosphorous disk in an environmental chamber at 55°C/85% RH, corrosion of the cobalt disk occurred directly over the chromated component when the total level of anions was greater than 200 ng/cm2. This accelerated corrosion did not occur in a 120-h test when the level was less than 50 ng/cm2. The amount of corrosion was found to be dependent on the separation distance between the disk and the chromated component.

Surface analysis of the disk corrosion area exposed to a chromate conversion coating showed NiO, CoO, and an increased concentration of chlorine.
Accelerated Environmental Testing of Magnetic Recording Disks
Authors: Setchell J
Published: March 2009

An accelerated test of the stability of images recorded on still video floppy disks was conducted by incubating disks at elevated temperatures. Small changes in the playback output were detected at 7 MHz, enabling prediction of storage life at ambient temperature by extrapolation. The test method revealed changes equivalent to a few nanometres in film thickness.
Corrosion Mechanism of Nd-Fe-B Magnets in Humid Environments
Authors: Kim AS , Camp FE , Constantinides S
Published: March 2009

Although Nd-Fe-B magnets exhibit outstanding magnetic properties, they have suffered from poor corrosion resistance in humid air. Because of this, various problems occur such as degradation of magnetic properties and/or contamination of the magnetic circuits. The corrosion rate of Nd-Fe-B magnets was found to vary with O, C, and N contents and to reach its minimum when these elements are within the range of: O = 0.6 to 1.2%; C = 0.05 to 0.15%; and N = 0.05 to 0.10%. To understand the corrosion mechanism of Nd-Fe-B magnets and the effects of these elements on corrosion resistance, various Nd-Fe-B magnets were metallographically examined. The corrosion resistance of Nd-Fe-B magnets is closely related to the microstructure, which is determined by controlling the minor elements such as O, C, and N. When the oxygen or the carbon content is low, the alloy forms a thick Nd-rich phase (α-Nd or α-Nd and Nd-oxide) on the grain boundaries. As oxygen and carbon contents increase, the Nd-rich phase at the grain boundaries becomes thinner and agglomerates into the triple junction. As oxygen content increases, the α-Nd changes to the Nd-oxide. This microstructural change may be responsible for improving the corrosion resistance. A small increase in N content further improves corrosion resistance.
The Influence of a Magnetic Field on Corrosion of Steel
Authors: Yee S-SW , Bradford SA
Published: March 2009

Magnetic influence has been suspected or reported as a contributing factor in many cases of metallic corrosion near magnetic sources. This study has examined the corrosion behavior of a plain-carbon steel in a 3% sodium chloride (NaCl) solution within and without magnetic fields up to 0.245 T at room temperature. Anode and cathode corrosion half-cells were separated so that the magnetic effect on each electrode reaction could be studied by itself. Direction of the field with respect to the electrode surfaces was also varied.

Cathodic polarization of steel is not affected by a magnetic field; neither is anodic polarization up to a current density of approximately 10 mA/cm2. It was also found by cyclic polarization that a magnetic field does not affect the susceptibility of the steel to localized attack in a 3% NaCl solution. However, at extreme corrosion rates (more than 24 cm/year), an applied magnetic field of 0.245 T causes a maximum change of 2.7% in the anodic dissolution rate.
Fracture Mechanics Parameters for Small Fatigue Cracks
Authors: Newman JC
Published: March 2009

The small-crack anomaly, where small cracks tend to grow either faster or slower than large cracks when compared on the basis of linear-elastic stress-intensity factors, has been shown to be significant for some materials and loading conditions. Conventional linear-elastic analyses of small cracks in homogeneous bodies are considered inadequate because of microstructural influences not accounted for in the stress-intensity factor and because of thenonlinear stress-strain behavior at notches and in the crack-front region. In this paper, plasticity effects and crack-closure transients are reviewed and investigated.

This paper presents a review of some common small-crack test specimens, the underlying causes of the small-crack effect, and the fracture-mechanics parameters that have been used to correlate or predict their growth behavior. Although microstructural features are important in the initiation and growth of small cracks, this review concentrates on continuum mechanics concepts and on the nonlinear behavior of small cracks. The paper reviews some stress-intensity factor solutions for small-crack test specimens and develops some simple elastic-plastic J integral and cyclic J integral expressions that include the influence of crack closure. These parameters were applied to small-crack growth data on two aluminum alloys, and a fatigue life prediction methodology is demonstrated. For these materials, the crack-closure transient from the plastic wake was found to be the major factor in causing the small-crack effect. Plasticity effects on small-crack growth rates were found to be small in the near threshold region, in that the elastic stress-intensity factor range and the equivalent value from the cyclic J integral gave nearly the same value.
Measurement of Small Cracks by Photomicroscopy: Experiments and Analysis
Authors: Larsen JM , Jira JR , Ravichandran KS
Published: March 2009

A procedure for using photomicroscopy to record the growth of small fatigue cracks is presented. Using a specially designed fatigue specimen, the method is applicable to both naturally initiated cracks and cracks initiated from a small electro-discharge machined notch. Components of the experimental apparatus, which are low cost and readily available, include a standard metallurgical microscope, a 35-mm camera with bulk film capability, an electronic flash, and a microcomputer to control the fatigue machine and record test data. The photographic record provides a direct measurement of surface crack length and documents crack interactions with microstructural features; measurement precision less than 1 μm is possible. Following a test, the photographs of small cracks are projected on a computer digitizing tablet for convenient measurement of crack length. The crack length data are then combined with fatigue cycle-count data and reduced to the form of da/dN versus ΔK (or some other appropriate crack driving force). The capabilities of the photomicroscopic method are illustrated using typical data from specimens of the alloy Ti-6Al-2Sn-4Zr-6Mo, and an assessment is made of the practical advantages and limitations of the technique. Finally, some commonly unrecognized pitfalls that routinely arise in the analysis of small-crack data are discussed, and an alternative procedure for the analysis of such data is presented.
Direct Current Electrical Potential Measurement of the Growth of Small Cracks
Authors: Gangloff RP , Slavik DC , Piascik RS , Van Stone RH
Published: March 2009

This paper demonstrates the successful application of the direct current electrical potential difference method for continuous in-situ monitoring of the growth of short (50 to 5000 μm) fatigue cracks in a variety of metallic alloys exposed to various environments. Crack initiation site and crack shape must be known a priori. Instrumentation includes a constant current power supply, relay circuit, 104 gain amplifier, and computer analog to digital board with controlling software. A closed-form analytical model accurately relates crack size to measured potential, as a function of crack shape and probe position local to the growing crack. Analytical calibrations are experimentally confirmed for small through-thickness edge cracks, surface semielliptical and semicircular cracks, and corner cracks in steels, nickel-based super-alloys, titanium and aluminum alloys. Micron-level average crack advance resolution, long-term stability, continuous measurement, simplicity, compatibility with aggressive environments, and programmed stress intensity loading are attributes of the electrical potential method. Crack length versus load cycles and growth rate versus stress intensity range data demonstrate the power of this method for studies of the effects of microstructure, environment (elevated temperature, high purity gases, vacuum and aqueous solutions) and loading variables on the growth kinetics of small and short fatigue cracks. Additional work is required for electrical potential monitoring of fatigue cracks sized below 75 μm, cracking not associated with a defined initiation site and thermal-mechanical fatigue.
Simulation of Short Crack and Other Low Closure Loading Conditions Utilizing Constant Kmax ΔK-Decreasing Fatigue Crack Growth Procedures
Authors: Hertzberg R , Herman WA , Clark T , Jaccard R
Published: March 2009

Small cracks can grow at ΔK levels previously thought safe (that is, ΔK < ΔKΔKth). It follows that ΔKth values, measured according to ASTM Test Method for Measurements of Fatigue Crack Growth Rates (E 647) procedures and associated with long-crack test specimens, may lead to nonconservative lifetime estimates of a component that contains very small cracks. One major reason for this discrepancy may be traced to the fact that different closure levels are found at the same applied ΔK level for long and short cracks, respectively. Much effort has been given to the generation of conservative fatigue crack propagation (FCP) data through the development of large quantities of short-crack data. Unfortunately, the generation of these data are time-consuming, tedious, and subject to considerable amounts of scatter.

An alternative test method, based on the use of standard-sized samples and employing established automated data acquisition procedures, has been identified. This method is based on maintaining a constant maximum stress intensity level (Kcmax) during the ΔK-decreasing test procedure. By maintaining a constant Kmax value, mean stress and associated R levels are found to increase markedly as ΔK decreases. The development of these high levels of mean stress and R ratios produces a long crack with negligible crack closure, which provides an upper bound estimate for the behavior of short cracks. Kcmax FCP data for several aluminum, iron, and nickel-based alloys have been generated and demonstrate the utility of the Kcmax test methodology as a simple and convenient means to obtain upper bound estimates of short-crack behavior in these structural alloy systems.
Contribution of Individual Spectrum Load Cycles to Damage in Notch Root Crack Initiation, Short and Long Crack Growth
Authors: Sunder R
Published: March 2009

A unified cumulative damage picture of spectrum load notch fatigue is presented, covering all its three stages: crack initiation, short crack growth, and long crack growth. The analysis uses rainflow cycle count, low-cycle fatigue, and crack growth constants as inputs. While cumulative damage in crack initiation is handled using conventional methods, recent observations of a substantial difference between notch root crack closure and crack opening stress serve as the inputs for modeling sequence sensitive damage accumulation in short and long crack growth. Range/damage-exceedance (RDE) diagrams showing contribution of individual load cycles to spectrum load cumulative damage are presented for an Al-Cu alloy (2014-T6511) subject to FALSTAFF and TWIST load spectra. These diagrams include information on load sequence sensitivity of damage in crack initiation and growth. RDE diagrams can be used to assess effects of load sequence reconstitution from rainflow cycle count and of load spectrum editing on cumulative damage.
A Creep Cavity Growth Model for Creep-Fatigue Life Prediction of a Unidirectional W/Cu Composite
Authors: Kim Y-S , Verrilli MJ , Halford GR
Published: March 2009

A microstructural model has been developed to predict creep-fatigue life in a [0°]4, 9 volume percent tungsten fiber-reinforced copper matrix composite at the temperature of 833 K. The mechanism of failure of the composite is assumed to be governed by the growth of quasi-equilibrium cavities in the copper matrix of the composite, based on the microscopically observed failure mechanisms. The methodology uses a cavity growth model developed for prediction of creep fracture. Instantaneous values of strain rate and stress in the copper matrix during fatigue cycles were calculated and incorporated in the model to predict cyclic life. The stress in the copper matrix was determined by use of a simple two-bar model for the fiber and matrix during cyclic loading. The model successfully predicted the composite creepfatigue life under tension-tension cyclic loading through the use of this instantaneous matrix stress level. Inclusion of additional mechanisms such as cavity nucleation, grain boundary sliding, and the effect of fibers on matrix-stress level would result in more generalized predictions of creep-fatigue life.
The Cumulative Fatigue Damage Behavior of MAR-M 247 in Air and High-Pressure Hydrogen
Authors: McGaw MA , Kalluri S , Moore D , Heine J
Published: March 2009

The cumulative fatigue damage behavior of the cast nickel-base superalloy, MAR-M 247 (in a fine grain form), was characterized experimentally at room temperature in ambient air and in high-pressure gaseous hydrogen. The material was first characterized in simple fatigue, consisting of fully reversed strain- and load-controlled axial fatigue tests. These data were used to establish the reference life behavior for use in the cumulative fatigue experiments. The principal tool used to study the cumulative damage behavior was the two-level loading test (single-block) wherein low-cycle fatigue loading is applied initially to the specimen for various low-cycle fatigue life fractions, and the specimen is subsequently cycled under high-cycle fatigue conditions to failure. MAR-M 247 was found to exhibit a strongly nonlinear cumulative interaction behavior, with the remaining high-cycle fatigue life capability in the second load level reduced by as much as a factor of 30 over results predicted from a linear damage rule. The Damage Curve Approach of Manson and Halford was used to predict this nonlinear cumulative behavior.

Similar experiments were performed at room temperature in a high-pressure gaseous hydrogen environment to assess the cumulative damage behavior and applicability of the Damage Curve Approach under differing environmental conditions. To study the applicability to more general loading cases, multi-block loading experiments were performed in air, in which each block consisted of both low-cycle and high-cycle fatigue cycling in prescribed amounts, ranging from one low-cycle fatigue cycle plus a fixed number of high-cycle fatigue cycles, to many low-cycle fatigue cycles plus a fixed number of high-cycle fatigue cycles. This basic loading block was then repeated until failure of the specimen. The results of these experiments and the predictions using the Damage Curve Approach are presented and discussed.
Crack Density and Fatigue Lifetime of Metals Under Variable Amplitude Loading
Authors: Bomas H , Mayr P
Published: March 2009

During cycling with constant and variable amplitudes, the density of the fatigue cracks of a copper alloy, a mild steel, and an austenitic steel was measured. Cycling with constant amplitude showed that the crack density increases with rising stress amplitude and advancing number of cycles. Cycling with variable amplitudes showed that the lifetime depends on the sequence of loading. Numerous changes from high to low amplitudes lead to a short lifetime. This effect depends on the material and is very pronounced for the copper alloy. The crack density was calculated by an addition model, which is based on the constant amplitude data, and was compared with the measured crack density. Lifetime was calculated after the method of Miner. For all metals investigated there exists a unique relation between the ratio measured crack density/calculated crack density and the realized damage sum after Miner. The crack density is always higher than expected when the realized damage sum is lower than the theoretical value of 1. This indicates that the influence of the sequence of loading on the lifetime bases on an alteration of the crack initiation rate in variable amplitude loading compared with constant amplitude loading and that crack coalescence plays an important role in failure of the materials investigated.
Determination of Fatigue Limit Between 105 and 109 Cycles Using an Ultrasonic Fatigue Device
Authors: Bathias C , Jingang N
Published: March 2009

High-power ultrasound was used to study fatigue behavior of three alloys (Udimet 500, 17-4PH, and Ti-6A1-4V) since the loading frequency is very high (f = 20 kHz) and a great number of cycles can be obtained in a short time. The specimens were excited in resonance at the first longitudinal mode in such a way that maximum strain occurred in the middle and maximum displacement amplitude was obtained at the end of the specimen. Fatigue life behavior of the three alloys were investigated at ultrasonic (/ = 20 kHz) and conventional loading frequency (f = 20 to 50 Hz). Miniature dynamic strain gages were employed to measure maximum strain values during ultrasonic fatigue tests. The effect of high strain rate on fatigue crack initiation and propagation failure mechanism was examined. Fracture surface of specimens was observed by means of a scanning electron microscope. Fatigue resistance was found to depend upon ultrasonic and mechanical characteristics and microstructure of the material, loading regime, and surface condition of the specimen.
Novel Methodology for Fatigue Lifetime Prediction
Authors: Puskar A
Published: March 2009

A unique fatigue lifetime equation offers the possibility of predicting residual fatigue lifetime of a material. Derivation of this lifetime equation is based on a correlation between the total and plastic strain amplitude during cyclic loading. A ratio determined as the reciprocal value of the elasticity modulus degradation is evaluated from the continuous measurements of the specimen resonance frequency in a resonance testing machine. For successful application of this procedure, coefficients and exponents for a fatigue curve, Coffin/Manson curve and cyclic stress-strain curve, or coefficient and exponent for the new fatigue lifetime equation are needed. This new equation and the well-known equations of cyclic plastic response are fully compatible. Measurements of dynamic elastic (secant) modulus of a virgin specimen and of a specimen after previous cyclic loading then provide an opportunity to evaluate the residual fatigue life.

The verification of the proposed procedure was performed using a carbon steel with 0.485 mass % carbon at a test frequency of 20.2 kHz, using three different rates of increase for the total strain amplitude before starting ordinary fatigue testing (loading ramps).
Impact of U.S. NRC Reactor Vessel Embrittlement Research on Regulation of Nuclear Power Plants
Authors: Serpan CZ
Published: March 2009

The embrittlement of nuclear reactor pressure vessels caused by neutron radiation is the chief aging mechanism of these critical components. It is so critical to safe operation of reactors that the U.S. Nuclear Regulatory Commission (USNRC) and its predecessor, the Atomic Energy Commission, have spent millions of dollars over the past three decades to document the effects. Furthermore, USNRC regulations and associated regulatory documents derived from that research basis now provide guidance on establishing operational conditions to assure that the pressure vessels of U.S. reactors will behave in a ductile manner, and that vessel materials will exhibit a sufficiently high resistance to ductile fracture.

The regulations and regulatory guidance that pertain to assuring the integrity of U.S. reactor pressure vessels are described and related to the operational constraints. Areas where additional data are needed, and where the analyses may be improved are noted and described. Finally, a prognosis is offered for future improvements in the technical basis for U.S. regulations.
Technical Issues Relevant to ASTM LWR Surveillance Standards
Authors: McElroy WN , McElroy RJ , Gold R
Published: March 2009

Specific physics-dosimetry-metallurgy issues are considered that will impact on current and future revisions of the thirty-four standards that have been identified for the next revision of the ASTM E706 Master Matrix of LWR Surveillance Standards, E706(0). Two major classes of technical issues are identified, namely those associated with 1) physics dosimetry methodology and 2) physically based modelling methodology. Symposium papers were reviewed to select the more important issues that exist in these two classes. A synthesis of these issues is presented.
VVER-440 Reactor Vessel Exposure Monitoring in the Czech Republic
Authors: Ošmera B , Svoboda č , Hep J , Holman M , Hógel J , Valenta V
Published: March 2009

A short review of the current reactor pressure vessel fast neutron fluence monitoring and evaluation methodology is presented. The reactor pressure vessel exposure must be evaluated for every fuel cycle combining calculations, surveillance monitors, exvessel (cavity) monitors and other experimental results. The series of WER-440 mock-up experiments were carried out in the LR-0 experimental reactor in the Nuclear Research Institute. The results were revised and the authorized data files were completed. A recommended practice for the use of mock-up results in fluence monitoring system was prepared. A database consisting of the surveillance and exvessel fast monitor data and the mock-up results is used to increase the reliability of the reactor pressure vessel exposure evaluation.
Revision of Neutron Dosimetry for the Loviisa WER-440 Reactors: Principles and Application
Authors: Baers LB , Serén T0
Published: March 2009

A comprehensive re-evaluation of all neutron dosimetry results for the Loviisa reactors (WER-440, 2 units) started in 1992.

Revised and better validated neutron spectra, a refined semiem pirical approach utilizing the computer program PREVIEW [4], unified and up-dated nuclear data and refined and extended calibrations of the applied Liquid Scintillation Counter (LSC) for 93mNb activity measurements are applied. The kernel-based computer program PREVIEW makes it possible to take into account the detailed local irradiation history e.g. the influence of the calculated time-dependent axial flux changes during a cycle.

The main features of this re-evaluation are reported together with updated results for two IAEA-CRP surveillance chains (SC) denoted “4K” and “5K”.

The fluence (>1 MeV) estimates increased by +16 to +20% and the dpa estimates by +24 to +27% as compared to previous results. The largest changes, about +20% in the >1 MeV fluences, arose from the revised and better validated neutron spectra. The refined treatment in PREVIEW of time-dependent variations fin the axial flux distribution accounted for 0 to −5% in the fluence estimates based on 54Fe (0 to −16% for 58Co).

Nb, Fe, Ni, Ti and Cu dosimeters were used for validations of the calculated neutron spectra built into the PREVIEW kernel library.

The quality of the neutron fluence estimates could be signifi cantly improved by applying revised and better validated neutron spectrum shapes and by applying a refined semiempirical approach taking into account calculated axial flux changes during an operating cycle. The remaining significant uncertainties in the neutron spec trum shapes and cross sections emphasise the development and applica tion of additional and more reliable dosimeters for the lower fast neutron energies.
Pressure Vessel Fluence Monitoring at NPP with VVER: Routine Technique and New Approaches
Authors: Borodkin GI , Kovalevich OM , Lomakin SS , Sycheva NV
Published: March 2009

For Reactor Pressure Vessel (RPV) neutron fluence monitoring at a nuclear power plant (NPP) with VVER-type reactor the Russian Nuclear Regulatory Body (GOSATOMNADZOR) has recently recommended use of the ex-vessel cavity dosimetry combined with neutron transport calculations. By using activation dosimeters with long-lived reaction products on the base of 54Fe, 58Ni, 46Ti, 60Ni, 63Cu, 55Mn and 59Co the routine experiments are now carried out. The WER cavity dosimetry methodology has been developed on the basis of experimental results obtained at Rovno NPP Unit 1 (Ro-1), Unit 2 (Ro-2), Unit 3 (Ro-3), Novovoronezh NPP Unit 5 (NV-5), Yuzhno-Ukrainsk NPP Unit 1 (YuU-1) with VVER-440 and VVER-1000. The set of experiments has been performed during the various reactor fuel cycles including the low leakage core. More accurate evaluation of integral neutron spectrum in cavity is possible by using short-time irradiations of the dosimeters with short-lived reaction products. Now performing cavity experiments also include 237Np, 238U and 93Nb detectors. The track detectors, similar SSTR, may be used for cavity dosimetry, and now such set dosimeters are available for installation in cavities of some VVERs. Also the “on-line” fluence monitoring technique is now developed. A summary of experimental and calculational results is presented.
Investigations into Power Reactor Fluence Uncertainties and their Implications for Trend Curves
Authors: Helm JL
Published: March 2009

The analysis of surveillance capsule dosimetry has played a defining role in the database used for trend curve development. The extent to which calculational meth ods and nuclear data have improved over the past 30 years suggests that early analyses may be significantly biased. Accordingly, the comparability of fast fluence estimates (f) made nowadays to those of the late 1960s and early 1970s may be affected, with possible consequences for the trend curves fit to these data.
Special Dosimetry at Saint Laurent BL Mox-Loaded Unit
Authors: Bevilacqua A , Lloret R , Nimal JC , Zengh S , Rieg C
Published: March 2009

A special highly instrumented surveillance capsule and dosimetry in the cavity along three generating lines were installed in the Electricité de France Saint Laurent Bl 900 MWe pressurized water reactor, and irradiated during the 1991–92 cycle. This instrumentation was presented at the Seventh ASTM-EURATOM Symposium in Strasbourg.

The reactor and cycle were chosen to be representative of balanced loading with mixed oxides UO2-PuO2 fuel (MOX), with the following objectives: -studying the effects of the MOX loading on the exposure of the capsule and vessel, -checking the disturbances of the axial distribution of the reaction rates of the dosimeters, caused by the barrel formers, -validation of the exposure calculations in the thickness of th vessel, -testing the Solid State Track Recorder technique compared with a dosimetry by activation in the cavity, -determining the maximum irradiation temperature of the capsule

Using the mean fission sources over the irradiation period, a TRIPOLI computation (with the ENDF-BVI library and the probability tables method) evaluates the measured parameters and the neutron spectra accompanied by their variance-covariance matrices, with a viev to adjusting these spectra.

The devices implemented are described. The instrumentation loading and unloading operations were performed successfully. The available results are given: MOX loading in the second periphery seems not to significantly change the exposure parameters, calculations slightly underestimate the reaction rates in the capsule and tend to overestimate them in the cavity.
Neutron Dosimetry with the 58Ni(n,p)58Co Reaction at Short-Decay Times
Authors: Griffin PJ , Holm CV , Vehar DW , Kelly JJG
Published: March 2009

The 58Ni(n,p)58Co reaction is a very attractive choice for use as a neutron fluence monitor foil in reactor irradiations. The most important drawback to this reaction is the interference from the 9.15-h half-life 58mCo metastable state. A methodology is presented in this paper to allow the 58gCo ground-state activity to be read at short decay times and to be converted into the total 58tCo activity with no significant increase in the measurement uncertainty. This methodology involves modeling the 58mCo/58tCo population ratio. Both theoretical and experimental estimates of the energy dependence of this ratio are presented. A method is presented to accurately measure this ratio. An empirical model of the energy dependence of this ratio is presented to allow simple estimates of the ratio to be made prior to a measurement.
Applicability of Superheated Drop (Bubble) Detectors to Reactor Dosimetry
Authors: d'Errico F , Alberts WG , Apfel RE , Curzio G , Guldbakke S
Published: March 2009

The characteristics of superheated drop (bubble) detectors (SDD's) have been reviewed with respect to the possible application of these devices in reactor dosimetry. In particular, their ability to measure neutrons in the presence of a high noise level, elevated temperatures and intense γ background has been investigated. Based on these studies, the use of SDD's is proposed for the monitoring and analysis of neutron emission from spent fuel assemblies. Finally, the possibility to employ these detectors in radiation protection dosimetry around power plants is discussed.
Dynamic Compensation of the Silver Self-Powered Neutron Detector in the Ramp Program at the Osiris Reactor
Authors: Moulin DJ
Published: March 2009

The Silver self-powered neutron detector (SPND) is a common detector used in the ramp program at the OSIRIS reactor. The Silver SPND signal is a reference during steady states, but its response is too slow for monitoring transient tests. In order to compensate for the inherent time delay a mathematical processing method of the Silver SPND signal was developed. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation algorithm can be used for transient conditions as well as steady state conditions. A computer program reconstructs, in real-time, the dynamic neutron flux sensed by the Silver detector from the current measured between the emitter and the collector of the SPND. Although this method decreases slightly the signal-to-noise ratio, it maintains the SPND's characteristics and reduces the response time from about 10 minutes to less than 4 seconds for a step change in flux. This provides for prompt and accurate measurement of fuel rod power during ramp experiments in the OSIRIS reactor. This development makes the Silver SPND very suitable for many on-line monitoring applications.
The Modified Notched Neutron Spectrum Technique for the Surveillance of Hydrogen in Reactor Materials
Authors: Miller WH
Published: March 2009

The determination of hydrogen content in power reactor materials is a component of an overall surveillance program. Unfortunately, techniques that accurately and nondestructively measure hydrogen content at parts-per-million (ppm) levels have not previously been available. One possible hydrogen measuring technique is based upon the “notched neutron spectrum technique” that has been implemented by EG&G Idaho to characterize hydrogen content in reactor materials for the U.S. Navy. Modifications to this technique have already been demonstrated that increase its sensitivity and utility, allowing for a current accuracy of 0.4 micrograms of hydrogen per gram of steel or an equivalent weight of 6 micrograms in a typical steel sample. This paper describes further tests that have increased the dynamic range of these measurements and provided verification of the absolute calibration of the method.
Accurate Neutron Fluence Measurements Using Bonner Spheres
Authors: Alevra AV
Published: March 2009

Results from measurements performed with the PTB Bonner sphere spectrometer are reported in order to illustrate the capability of the instrument to accurately determine neutron fluence in cases where enough a priori information is available. An example is then given where the a priori information is very limited or even nonexistent, and for such cases the limitations of the method are shown.
Proton Recoil Spectroscopy 400 Meters from a Fission Neutron Source
Authors: Stanka MB
Published: March 2009

Neutron kerma and spectrum measurements have been made at the U.S. Army Pulse Radiation Facility (APRF) to 400m in an air-over-ground geometry from a fission neutron source and have been compared to Monte Carlo transport calculations. The neutron spectra measurements were made using a rotating neutron spectrometer. This spectrometer consists of four spherical proton-recoil detectors mounted on a common rotating base. Detector radius, gas composition, and pressure have been varied to allow sensitivity over a neutron range of 50 keV to 4.5 MeV. Neutron kerma was determined by using the Kerr soft-tissue kerma factors. Measured neutron kerma agreed with the calculated neutron kerma to within 5%. Comparisons with other neutron spectrometers such as NE213 and Bonner Spheres are presented and agreement between the different spectrometers is better than 20%.
Analysis of the Venus Out-Of-Core Activation Measurements Using MCBEND
Authors: Blaise PD , de Wouters RM , Abderrahim HA
Published: March 2009

The VENUS-1 critical assembly simulates a U02-loaded core and the internal structures from the core baffle to the neutron pad, in a generic 3-loop plant. VENUS-2 contained mixed oxide PuO2 — UO2 fuel pins in the eight outer rows. Several out-of-core threshold activation rates (In-115, Np-237, Zn-64, Al-27, Ni-58, U-238, Rh-103) have been measured at numerous locations in water zones and steel structures. From these measurements, “equivalent fission fluxes” have been established by dividing the activation rates by cross-sections weighted by the U-235 fission spectrum. Calculations of equivalent fission fluxes have been carried out with the Monte Carlo transport programme MCBEND. The average ratio calculation/experiment (C/E) in the core barrel, with 80 dosimeters and 7 types of reactions, is 1.02 ± 0.03 (δ). Further away from the core, the average C/E in the neutron pad from 13 measurements is 1.01 ± 0.08 (δ).
Test Calculations of the Neutron Flux on VVER-1000 Reactor Pressure Vessel
Authors: Ilieva KD , Belousov SI , Antonov SY , Zaritsky SM , Brodkin EB
Published: March 2009

A three dimensional test for calculation of the neutron fluence onto the VVER-1000 reactor pressure vessel (RPV) is presented. The test is based on the commercial VVER-1000 reactor design data. The flux results obtained by different authors are in good agreement.
Analysis of Pulsed-Sphere Experiments as Integral Data for Cross-Section Adjustment
Authors: Perel RL , Wagschall JJ , Yeivin Y
Published: March 2009

Cross-section adjustment by integral measurements in fission-reactor assemblies affects cross sections in the energy range around the peak of the fission spectrum and below. For fusion-based systems, the important energy range is much higher (up to 15 MeV). Pulsed-sphere experiments were analyzed in view of their potential use as integral data by which the relevant cross sections in the latter energy range could be adjusted. These analyses were carried out by means of Monte Carlo calculations. The differential operator approach was applied to the calculation of sensitivities. The Monte Carlo code MCNP was modified to facilitate this approach.
Neutron Leakage Benchmarks for Water Moderators
Authors: Gilliam DM , Briesmeister JF
Published: March 2009

Fission reaction rates for four nuclides were measured in the leakage spectrum outside spherical water moderators of various radii surrounding a 252Cf neutron source. Using the MCNP transport code, matching calculations were made with highly detailed modeling of the measurement apparatus. The calculations predicted significantly higher leakage of neutrons in the epicadmium energy range than was found in the measurements. A discrepancy of the same sign but weaker magnitude was found for thermal neutrons. These discrepancies may be relevant to problems with criticality calculations in special cases.
Validation of Neutron Transport Calculations on Magnox Power Plant
Authors: Mossop JR , Thornton DA , Lewis TA
Published: March 2009

This paper describes the validation of fast neutron transport calculations for Magnox RPVs operated by Nuclear Electric. The work compares calculated data, derived using the Monte Carlo code McBEND, with extensive plant measurements. Agreement is typically within ∼20%.

This level of agreement has been achieved using individual models for the top, side and bottom of each reactor in which almost explicit representation of all the major components was required. The modelling detail included the geometric effects of temperature and irradiation on reactor components as well as source data derived from reactor measurements and operating histories. Such effort is justified because the reduced uncertainty in the pressure vessel dose can be translated into increased operating margins and extended plant lifetimes.
“Concrete Benchmark” Experiment as Support to Ex-Vessel LWR Surveillance Dosimetry
Authors: Abderrahim HA , D‘hondt PJ , Risch P , Oeyen JB , Bioux P
Published: March 2009

The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several benchmark experiments (PCA, PSF, VENUS ...) that used the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The “Concrete Benchmark” experiment has been designed to judge the ability of this calculational method to treat the backscattering.

This paper describes the “Concrete Benchmark” experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations.
SILENE, a New Radiation Reference Source
Authors: Rozain JP , Barbry F , Medioni R
Published: March 2009

Originally designed to study criticality accidents, the SILENE reactor today is also a reference source able to meet the expectation of a great number of researchers in the fields of physics, neutronics, biology or dosimetry. The latest technical developments and its multiple operating possibilities enable this CEA Nuclear Protection and Safety Institute experimental reactor to provide a reproducible range of calibrated neutron and gamma radiation.
Development of an Alanine Dosimeter for Gamma Dosimetry in Mixed Environments
Authors: Vehar DW , Griffin PJ
Published: March 2009

L-α-Alanine, a nontoxic polycrystalline amino acid, has been investigated for use in high-precision, high-level absorbed-dose measurements in mixed neutron/photon environments such as research and test reactors. The technique is based on the use of electron paramagnetic resonance (EPR) spectroscopy to determine the extent of free radical production in a sample exposed to ionizing radiation, and has been successfully used for photon absorbed-dose measurements at levels exceeding 105 Gy with high measurement precision. Application of the technique to mixed environments requires knowledge of the energy-dependent response of the dosimeter for both photons and neutrons. Determination of the dosimeter response to photons is accomplished by irradiations in 60Co and bremsstrahlung sources and by calculations of energy-dependent photon kerma. Neutron response is determined by irradiations in conjunction with CaF2:Mn thermoluminescence dosimeters and by calculations of energy-dependent neutron kerma. Several neutron environments are used, including the ACRR and SPR-III reactors.
Reusability of CaF2:Mn Thermoluminescence Dosimeters for Photon Irradiations at High Absorbed-Dose Levels
Authors: Vehar DW
Published: March 2009

The effects of high absorbed dose on the reusability of CaF2:Mn thermoluminescence dosimeters (TLDs) are investigated to determine a recommended upper limit on absorbed dose for TLDs that are to be reused. This investigation examines degradation in the uniformity of response and changes in sensitivity of a batch of TLDs when exposed to gamma-radiation doses ranging from 1 Gy to 1000 Gy, and confirms earlier work suggesting that CaF2:Mn TLDs should not be reused in applications where cumulative absorbed doses are likely to exceed 100 Gy.
Measurement and Analysis of Gamma-Ray Distributions in Kyoto University Critical Assembly, KUCA
Authors: Kimura I , Kagehira K , Sakurai Y , Kimura J , Kanno I , Sato T , Shiroya S , Kanda K
Published: March 2009

Gamma-ray distributions in an undermoderated core of the Kyoto University Critical Assembly, KUCA-B, were measured with four thermoluminescence dosimeters (TLDs), magnesium, strontium and barium orthosilicate (Mg2SiO4(Tb), Sr2SiO4(Tb) and Ba2SiO4(Tb)); and calcium sulfate (CaSO4(Tm)). In advance of the measurements, the response of these TLDs to gamma-rays was calibrated with a 60Co source and that of Mg2SiO4(Tb) to neutrons was tested.

The neutron and gamma-ray flux distributions in the core of KUCA were simultaneously calculated by a transport code ANISN-JR, using a combined cross section set generated by RADHEAT-V3. The distribution of the gamma-rays from fission products was independently calculated. In order to obtain the absorbed dose of each TLD, the following three components were summed up: (1) prompt gamma-rays, (2) fission product gamma-rays and (3) neutrons. All data were normalized by the reaction rate of 58Ni(n,p)58Co at the core center.

In general, the absolute values and the shape of the absorbed dose distribution measured with the four TLDs agree with the calculated ones very well. However the measured values in the reflector region with Sr2SiO4(Tb), Ba2SiO4(Tb) and CaSO4(Tm) are considerably larger than the calculated results.
Reactor Vessel Dosimetry Assessment: Perspectives of the Materials Engineer
Authors: Steele LE , Beaudoin BF , Biemiller EC , Van Konynenburg RA , Server WL
Published: March 2009

Neutron exposure of the reactor pressure vessel is one of the factors that must be quantified when assessing safe vessel lifetime. As a result, the procedures for determining a neutron exposure level for the entire vessel or a critical location should have the same level of accuracy as other factors such as fracture potential based on known or projected mechanical properties. This paper considers the factors affecting neutron dosimetry and associated uncertainties that must be assessed prior to making realistic projections of the vessel condition. In addition, an attempt to describe in simple terms some complex, but important, areas of reactor vessel neutron dosimetry analyses will be made. Assessment of these issues will enhance the mutual understanding between nuclear physicists and chemists analyzing vessel dosimetry and the materials specialist evaluating the mechanical property changes.
Calculated Damage Rates for FFTF Structural Components and Surveillance Assemblies
Authors: Simons RL
Published: March 2009

The Fast Flux Test Facility (FFTF) surveillance program provides coupon surveillance materials that are irradiated to the expected lifetime damage dose that the represented component will experience. This methodology requires a knowledge of the damage dose rates to the surveillance assemblies and to the critical locations of the structural components. This analysis updates the predicted exposures from a total fluence to a displacement per atom (dpa) basis using the Monte Carlo computer code for neutron photon transport, MCNP. The MCNP calculation improves the relative consistency and lowers the predicted damage rates uncertainty in a number of out-of-core locations. The results were used as part of the evaluation to extend the lifetime of the in-vessel components to 30 years in support of multiple missions for FFTF.
Dosimetry Issues for an Ultra-High Flux Beam and Multipurpose Research Reactor Design
Authors: West CD
Published: March 2009

The Advanced Neutron Source is a new user facility for all fields of neutron research, including neutron beam experiments, materials analysis, materials testing, and isotope production. The complement and layout of the experimental facilities have been determined sufficiently, at a conceptual design level, to make reliable cost and schedule estimates. The source of neutrons will be a heavy water reactor, constructed largely of aluminum, with an available thermal neutron flux 5–10 times higher than existing research reactors.

Among the dosimetry issues to be faced are damage prediction and surveillance for component life attainment; measurement of fluence and spectra in regions where both change substantially over a distance of a few centimeters; and characterization and measurement of the radiation field in the research areas around the neutron beam experiments.