bluuebunny/crossref_metadata_2025_split huggingface dataset:
Neutron Flux Estimations Based on Niobium Impurities in Reactor Pressure Vessel Steel
Authors: Baers LB , Hasanen EK 
Published: March 2009

The use of (ppm level) niobium impurities in reactor pressure vessel (RPV) steel for neutron flux estimations based on the reaction 93Nb (n,n′) 93mNb has been reported previously [1].
The method has now been further investigated and refined.
Small niobium fractions in RPV steel (≈ ppm) and plating (≈1 %) materials have been separated by ion exchange chromatography in one to three steps. The measured Nb fractions in samples from some four pressure vessel (RPV) base materials were 1 to 3 ppm. The purification of tens of milligrams of RPV material provides sufficient amounts of niobium for mass determination with a highly sensitive (10−5 ppm) Inductively Coupled Plasma Mass Spectrometer (ICP-MS).
The 93mNb and small remaining 54Mn activities were measured with a calibrated Liquid Scintillation Counter (LSC) based on dual label technique and almost 100 per cent efficiency to 93mNb. One purification is needed for plating materials (≈1 % Nb) and two purifications of about one gram of steel with Nb impurities in order to resolve the needed activities (≈10 Bq 93mNb/ug Nb).
Other neutron induced reactions in Mo produce interfering 93mNb, 92mNb and 95Nb. With about 0.5 per cent Mo in the RPV steel, several tens of ppm of Nb (Nb/Mo &gt; 0.01) are typically needed in order to obtain a clearly dominant 93mNb activity from the 93Nb (n,n′) 93mNb reaction of interest. RPV plating materials with 0.5...1 % of alloyed Nb have insignificant degradation from Mo on the measured 93mNb activities.
Thus, a sufficient chemical purification of Nb was achieved and the specific 93mNb activities could be measured for all types of available RPV and plating samples. However, other competing and 93mNb producing reactions such as 92Mo(n,γ)93Mo (3500 y), disintegrating to 93mNb, strongly degraded the results in samples with low Nb/Mo ratios. Samples with sufficiently high Nb/Mo ratios may be eventually selected for dosimetric purposes since the Nb and Mo mass fractions vary with position and type of material.
The achieved accuracy of the measured specific 93mNb activities was about ±3 % (lσ) in irradiated RPV plating materials and about ±4 % for Nb ppm impurities.
Materials Prepared at the Institute for Reference Materials and Measurements (IRMM) for Reactor Dosimetry Applications
Authors: Ingelbrecht CD , Peetermans FW , Palmeri S , Robouch PB 
Published: March 2009

Thirteen reference materials prepared under the Euratom Working Group on Reactor Dosimetry (EWGRD) monitor materials programme are now available. The determination of trace impurities has recently been completed for rhodium metal (iridium 26.0 ± 0.6 mg·kg−1, platinum &lt; 5 mg·kg−1) and these values will be proposed for certification. Certification analyses of titanium metal, which are continuing, indicate scandium &lt; 0.1 mg·kg−1.
The preparation of alloys for neutron fluence rate measurements are discussed. Satisfactory homogeneity in aluminium alloys can be difficult to achieve on a milligramme scale due to the formation of coarse intermetallic particles, and suitable processing with a high solidification rate, cold working and annealing is essential. The example is given of the alloys Al-15 wt% U and Al-14 wt% Pu, where the inhomogeneity (in terms of relative standard deviation of alloying element) was determined by isotope dilution mass spectrometry as &lt; 0.2% in each case for 20 mg samples.
Some recent examples of dosimeter encapsulation are given including NpO2 and UO2 microspheres in high purity vanadium and Al-Pu and Al-U fission foils in cold-pressed aluminium containers.
Signal Anticipation of Rhodium Self Powered Neutron Detectors for On-Line Applications
Authors: Lacaille R , Jegat A 
Published: March 2009

rhodium self powered neutron detectors (SPND) have been used for a long time in the experimental devices at the SILOE reactor.
In a time-dependant neutron flux, the rhodium SPND behaves as a first order linear system. Consequently, the response of the SPND is not similar to the incident flux evolution because time constants are involved. In order to draw-out the true flux variation, a convolution type method is to be used.
As a result of the time discretisation of the linear differential system that couples the SPND current to the neutron flux, one has to use differential equations. The resolution of these equations is equivalent to an approximate inversion of the initial system.
The numerical method described above has been applied in a microcomputer system, in order to treat the SPND signal by a computer-driven millivolmeter.
Nowadays, devices for transient experiments in the SILOE reactor are driven by a P.C. system, based on this numerical approach and using an independent data acquisition; the reliability and the ergonomics aspects have been taken into account. The modular conception of the system allows easy further developments for flux monitoring in various surroundings.
Characterization of the Gamma Field in the B%W Owners Group Cavity Dosimetry Benchmark Experiment
Authors: King SQ , Gilliam DM 
Published: March 2009

The B%W Owners Group (BWOG) Cavity Dosimetry Benchmark Experiment included, in addition to various types of neutron dosimeters, lithium-7 fluoride (LiF) optical absorbance gamma dosimeters placed at eight cavity locations. The purpose of these dosimeters was to indirectly measure the gamma field in the vicinity of the neutron dosimeters. These measured data were used in a semi-empirical procedure to characterize the gamma-ray dose rate distribution in the power reactor cavity.
The LiF chip dosimeters were irradiated for one cycle in the high-level mixed neutron-gamma field that existed in the power reactor cavity. The 380.3-day cycle length was sufficient to ensure absorbed doses in the tenths of megagrays. After being removed from the reactor, the LiF chips were shipped to the National Institute of Standards and Technology (NIST) for measurement of the optical absorbance, annealing, and re-calibration in a known gamma ray field. The neutron flux spectrum at each LiF chip location was calculated using a semi-empirical methodology that reconciles measured neutron dosimeter data to calculated neutron spectra. Appropriate energy-dependent normalization factors were developed and applied to the a priori flux calculated by the DOT4.3 discrete ordinates transport code. The neutron flux spectra were then used to determine the total neutron dose absorbed in each LiF chip. Finally, the calculated neutron dose was subtracted from the measured (total) dose, yielding the integrated gamma-ray dose. Calculated-to-measured (C/M) gamma dose ratios were then established for each LiF chip and are reported herein.
Overview of the Activities in Spain on Irradiation Embrittlement of RPV Steel
Authors: Bros J , Ballesteros A 
Published: March 2009

Specific problems such as the embrittlement of the WER reactors or the integrity evaluation of the Yankee Rowe reactor pressure vessel, show the necessity of continuing in the understanding of the mechanisms responsible for the embrittlement. Besides, in order to increase the safety and performance of the nuclear power plants, it is important to optimize the surveillance programmes and develop the mitigation techniques of the damage caused by neutron radiation.
The main activities carried out in Spain are related to the surveillance programmes and to phase III of the IAEA Coordinated Research Programme, although other smaller projects also show the interest of the utilities in the embrittlement issues. The extension of the work performed as well as some of the most relevant results are shown in this paper. Some new projects are being considered. The participation in the recently created European Action Group for RPV Materials Irradiation Effects and Studies (AMES) is of huge value for Spain. This Group try mainly give a European response in the field of the standards. In spite of the nuclear moratorium that has been in vigour in Spain for several years and the present economic crisis, there is a general interest in participation in international projects. That is the case of the Spanish contribution to the certification of Avanced Light-Water Reactors. In particular, regarding the embrittlement problem, in the development of a reactor pressure vessel and support skirt material surveillance programme for the Simplified Boiling Water Reactor (SBWR).
A Plant-Specific Upper Shelf Energy Drop Methodology
Authors: Manahan MP , Cuddy LJ , Peterson AJ 
Published: March 2009

Appendix G to 10 CFR 50 requires that licensees evaluate all reactor beltline materials to determine whether the Upper Shelf Energy (USE) screening criterion will be exceeded prior to the end-of-license (EOL). Measurement of the unirradiated USE and estimation of the EOL irradiated USE level are key elements of the screening criterion evaluation. The current Regulatory Guide 1.99 (Revision 2) USE drop model does not provide adequate representation of the actual shelf drop behavior for some reactor materials. Therefore, research was conducted to identify the important factors which should be included in the development of a physically-based USE trend curve model and to test the validity of the material-specific modelling approach.
As a result of this investigation, it has been concluded that there are three radiation damage mechanisms which may have significant influence on the ductile fracture process: precipitation and particle coarsening due to radiation enhanced diffusivity; cascade induced matrix damage; and possibly surface active element transport to grain boundaries and particle interfaces. At present, it is believed that the largest micromechanical effect is due to the impact of precipitates and matrix damage on the hardening of the material bridges between voids. However, additional experimental data are needed to confirm the dominant mechanisms which contribute to the USE drop.
In the absence of an experimentally verified microstructural model, work has been performed to test the efficacy of the material-specific approach. The validity of the material-specific approach has been demonstrated by reducing the Light Water Reactor (LWR) database to produce a data set with comparable chemical composition, heat treatment, and neutron flux spectra. Based on the work reported herein, it has been observed that a drop in the unirradiated USE requires an incubation dose. The reduced data set correlates well with the square root of the dose, indicating the importance of the fine scale precipitates and the matrix damage component.
Calculated and Measured Gas Formation in Beryllium Samples Irradiated in the High Flux Materials Testing Reactor BR2
Authors: De Raedt CM , Sannen LF , Vanmechelen PJ , Oliver BM 
Published: March 2009

Beryllium samples have been irradiated in BR2, the materials testing reactor of the Nuclear Research Centre SCK/CEN at Mol, Belgium, up to fission fluence values of 5.2 1022 n/cm2 at low temperature.
The gas formation (helium, tritium), as measured by SCK/CEN, as well as the induced swelling of the beryllium samples and the enhancement of the swelling due to annealing have been presented at the 17th SOFT Conference (Rome, 14–18 Sept., 1992).
Since this conference, helium measurements on the same samples have been carried out at RI and calculations of the gas production have been performed, taking into account the various formation schemes.
The experimental results from SCK/CEN and from RI are compared with the calculated gas formations.
The Spallation Neutron Source SINQ and Related Dosimetry Problems
Authors: Janett A , Atchison F , Bauer GS 
Published: March 2009

The spallation neutron source SINQ, presently under construction at Switzerland's Paul Scherrer Institut, will handle the highest proton current of any comparable facility in the world: a continuous beam of 1.5 mA, 590 MeV protons from an isochronous ring cyclotron. For the users, SINQ as a neutron source should resemble closely a medium flux research reactor; the presence of high and medium energy particles creates new technical problems in design and operation. The engineering design for the major components is based on the results of neutronic calculations, using a code package built around the HETC program.
At present, opportunities to verify the theoretical calculations experimentally are very limited. Safety factors have to be built in which conflict with the optimation of SINQ as a neutron source. To benchmark the calculational methods, a wide ranging diagnostic system will be required.
Two-Dimensional Neutron and Gamma Calculations for Devices Irradiated in BR2 Reflector Positions
Authors: De Raedt CM , Bodart SF , Ponsard BJ , Wéber MJ , Maldague T 
Published: March 2009

For two typical LWR fuel rod irradiation devices introduced into the BR2 reflector two-dimensional neutron and gamma calculations are being carried out.
In the first case a single fuel rod surrounded by several pressure tubes and by a time-variable neutron screen (He−3 gas at various pressures) is studied. The calculations are carried out in R,θ geometry, centered on the irradiated fuel rod.
In the second case a bundle of nine fuel rods in a 3×3 square lattice is irradiated. Here the calculations are performed in X,Y geometry.
The necessity is discussed of performing two-dimensional calculations for these irradiations in the BR2 reflector, to improve the modelling of the fuel behaviour.
Measurements of Thermal Neutron Cross Section and Resonance Integral for the 237Np(n,γ)238Np Reaction
Authors: Kobayashi K , Yamanaka A , Kimura I 
Published: March 2009

Making use of a standard neutron spectrum field with a pure Maxwellian distribution at the heavy water thermal neutron facility of the Kyoto University Reactor (KUR), the thermal neutron cross section for the 237Np(n,γ)238Np reaction was measured by the activation method, using a high purity Ge detector. The result is 158±3 b, which is obtained relative to the reference value of 98.65±0.09 b for the 197Au(n,γ)198Au reaction. The present value is lower by about 13% than that of the ENDF/B-VI data. The data given by Mughabghab and obtained from JENDL-3 are larger by 11 to 14% than the present measurement.
The resonance integral for the 237Np(n,γ)238Np reaction was measured with a 1/E standard neutron spectrum in the cavity of the central graphite reflector region between the two-divided cores of the Kinki University Reactor (UTR-KINKI), relative to the reference value of 1550±28 b for the 197Au(n,γ)198Au reaction. By defining the Cd cut-off energy as 0.5 eV, the present resonance integral is 652±24 b, which is in good agreement with the JENDL-3, ENDF/B-VI and Mughabghab data. However, most of the old experimental data are, in general, larger by 24 to 38% than the present measurement.
Effect of Gadolinium Covers on Fast Dosimeter Response
Authors: Rutherford MA , King SQ 
Published: March 2009

Gadolinium (Gd) has been used extensively at BSW Nuclear Technologies (BWNT) as a dosimetry shielding material for several years. Gadolinium has a significant cross section throughout the energy range of interest (i.e. above 1 MeV) in vessel fluence calculations. While Gd has a number of advantages as a dosimetry cover, it also has the potential for introducing a bias of unknown magnitude into the measured activities of covered fast response dosimeters such as U238 and iron. The Gd cover causes a reduction in the measured dosimeter activity, which means that the calculated fluxes normalized to those activities would be under-predicted. This is, of course, in the non-conservative direction. Since dosimeter covers are generally not included in the typical transport analysis model, corrections for this effect are not generally made. The calculated activities would not include the effects of the Gd while, of course, the measured activities would. The resulting calculation to measurement ratios, (C/M), which are used to remove biases in the calculated fluxes, would therefore be in error, in the non-conservative direction, by some undetermined amount.
This study has determined the magnitude of the bias in the fast response dosimeters, U238 and iron caused by the absorption of fast (E&gt;1.0 MeV) neutrons in 2 mm (0.002 m) Gd covers. The results indicate that the bias is in the 2.9–3.4% range and that it is only a weak function of dosimeter material.
Variability of Irradiation Growth in Zr-2.5Nb Pressure Tubes
Authors: Fleck RG , Elder JE , Causey AR , Holt RA 
Published: March 2009

Pressure tubes in CANDU reactors are manufactured by extrusion and cold work from the Zr-2.5Nb alloy. Historically, the details of the pre-extrusion production of the tubes have changed since the first tubes were produced (1969) for Ontario Hydro's Pickering A station. The pressure tubes exhibit variations in elongation rate during service that may partially result from these changes. Irradiation growth is a component of the elongation. This paper describes the longitudinal irradiation growth of specimens prepared from pressure tubes with a range of iron contents and in which billets were extruded either in the as-forged condition or after β-quenching.
The irradiations are performed in the Osiris reactor at Saclay, France, in two inserts with nominal operating temperatures of 280°C (Trillium 2) and 310°C (Trillium 3). Results are reported for fast fluences up to 4.3 × 1025 n/m2, E &gt; 1 MeV. Throughout the irradiation, the temperature of each tier of six specimens is recorded and this shows that the actual range of temperatures is 262 to 309°C. In agreement with previous studies, the temperature dependence of growth is negative, that is, an increase in temperature decreases the growth rate.
When this experimental variable is taken into account, there is still considerable variation among specimens from different pressure tubes. There is no effect of β-quenching. However, an increase in iron in the range 367 to 1060 ppm (by weight used exclusively in this paper) reduces the growth rate by about 2 × 10-29 m2/n, E &gt; 1 MeV at both 280 and 310°C. If this difference extends to high fluences, it accounts for about half of the variability in elongation rates observed in pressure tubes in-service.
Optimization of PWR Behavior of Stress-Relieved Zircaloy-4 Cladding Tubes by Improving the Manufacturing and Inspection Process
Authors: Mardon J-P , Charquet D , Senevat J 
Published: March 2009

With the aim of optimizing the basic properties of stress-relieved Zircaloy-4 cladding tubes, particularly those that make it possible to push back the initial technological limits that may be encountered, and of reducing the scatter of those properties and enhancing tube quality, the role of the main parameters involved in manufacturing the ingot, Trex, and cladding tube has been evaluated on an industrial scale. A series of large-sized tube lots were produced under controlled manufacturing conditions, then characterized by out-of-pile test results (short-and long-term corrosion, stress corrosion cracking (SCC), creep, mechanical, and structural properties) on finished tubes. For the investigated parameters (chemical composition, number of melt, quench rate, accumulated annealing parameter, the ΣA factor, surface condition (outside and inside diameters), and finished tube quality), this role is indeed important but complex due to the highly interactive nature of the variables investigated.
Adjustment of the chemical composition within ASTM limits enables generalized corrosion resistance to be enhanced and irradiation growth to be minimized. A significant decrease of the observed scatter in corrosion and mechanical properties is obtained by optimization of the ΣA range, the quenching rate, and the final heat treatment. The optimum seems to be reached for a final treatment at the highest possible temperature compatible with the stress-relieved state, corresponding to an average precipitate size and ΣA. Moreover, by adding anneals upstream in the process, a further increase in this ΣA no longer seems to have a significant effect on generalized corrosion.
Finally, extensive efforts have been employed in the pilgering, surface preparation (outside diameter polishing, flush-pickling), and examination method (UT, EC) leading to a sizable improvement in SCC resistance and to a reduction in scatter for finished tubes.
The result of these optimizations has been implemented in the current AFA-2G, that shows that under irradiation a 30% corrosion gain is reached after three cycles, without degrading creep strength or growth. The intrinsic effect of tin on generalized corrosion resistance under irradiation was also confirmed on this occasion.
Experimental and Theoretical Studies of Parameters that Influence Corrosion of Zircaloy-4
Authors: Billot P , Robin J-C , Giordano A , Peybernès J , Thomazet J , Amanrich H 
Published: March 2009

Waterside corrosion of Zircaloy cladding in pressurized water reactors (PWRs) is largely dependent upon the operating parameters and microstructure of the zirconium alloys. The impact of these parameters on the corrosion kinetics of Zircaloys is investigated on the basis of empirical data and experiences that can be interpreted using existing corrosion models.
The influence of thermo-hydraulic data, heat flux, local boiling conditions, and of the growing oxide films has been studied from corrosion tests performed in static autoclaves or in out-of-pile loops. These parametric investigations are described as well as the models that were developed.
The impact of microstructure is studied from the comparison of the corrosion behavior of different Zircaloy-4 specimens corroded in out-of-pile tests. In particular, a poor corrosion resistance of an experimental Zircaloy-4 material is analyzed as a function of the microstructure close to the metal/oxide interface.
The impact of the alloy composition and primary coolant chemistry on the corrosion kinetics of Zircaloy-4 is modeled empirically or uses a mechanistic approach that proposes a series of chemical equations with a mathematical representation of the kinetics.
These proposed models are then used to investigate the corrosion behavior of Zircaloy-4 cladding in 17 by 17 plants for rods irradiated at high burnups. Higher PWR operating cycles, core average coolant temperature, power, and elevated primary coolant lithium concentrations (3.5 to 4 ppm) are then simulated and discussed in terms of Zircaloy corrosion resistance considerations.