bluuebunny/crossref_metadata_2025 huggingface dataset:
Integral Assessment of the Revised JENDL Dosimetry File
Authors: Kobayashi K , Iguchi T , Iwasaki S , Aoyama T , Shimakawa S , Ikeda Y , Odano N , Sakurai K , Shibata K , Nakagawa T , Nakazawa M
Published: April 2008

The JENDL Dosimetry File, which contains 67 dosimetry reactions, has been revised by the Dosimetry Integral Test Working Group of the Japanese Nuclear Data Committee. Thirty-one cross sections and their covariance data were simultaneously evaluated mainly through a generalized least-squares code using the experimental data in EXFOR. In order to confirm the reliability of the present revised data, integral tests have been carried out by comparing the calculated spectrum-averaged cross sections with the measured ones in fission neutron fields, fast/thermal reactor neutron fields, D-T and D-Li fusion-like neutron fields. In most of the dosimetry cross sections, it has been found that the ratios of calculated to experimental results on their integral tests show reasonable agreement within predicted uncertainty, while unreasonable discrepancies are still observed in some revised JENDL dosimetry cross sections.
High Energy Neutron Activation Cross Sections
Authors: Nakamura T , Kim E , Uwamino Y , Nakao N , Tanaka S
Published: April 2008

The cross section data of 12C(n,2n)11C and 209Bi(n,xn)(x=3,12) reactions and the cross section data which produce 56Mn, 54Mn, 52Mn, 58Co, 57Co, 56Co, 55Co, 59Fe from 59Co reactions and 56Mn, 59Fe, 60Co, 58Co, 57Co, 56Co, 64Cu, 61Cu, 60Cu from natCu reactions have been measured in the quasi-monoenergetic p-7Li neutron fields in the 20- and 150-MeV energy range. The irradiation experiments were performed at four cyclotron facilities:(a) the Institute for Nuclear Study (INS), University of Tokyo;(b)the Cyclotron and Radioisotope Center, Tohoku University; (c) the Takasaki Research Establishment (TIARA), Japan Atomic Energy Research Institute; and (d) the Institute of Physical and Chemical Research (RIKEN). The neutron spectrum has been measured with the time-of-flight method using an organic liquid scintillator, and the absolute value of peak neutron fluence has been estimated from the activation method of lithium target at INS, RIKEN, and the proton recoil counter telescope at TIARA.
SPALLDOS: New Neutron Metrology Cross-Section Library for Use at Spallation Neutron Sources
Authors: Szondi EJ , Zsolnay ÉM , Hegedüs F , Tobler L
Published: April 2008

Neutron spectrum measurements have been performed by the multiple-foil activation method at the spallation neutron source SINQ (PSI, Switzerland). The energy region of interest ranges from thermal energies to 590 MeV. A new high-energy neutron cross-section library has been prepared for use at spallation neutron sources. First a master library, called SPALLDOS, was produced, containing cross-sections for 239 neutron activation and radiation damage characteristic reactions. From the master library, problem dependent cross-section libraries — corresponding to the interesting measurement positions at SINQ — with 138 reactions were derived. Part of the cross-section data was tested by experiments.
New Measurements of the H(n,n)H Angular Distribution
Authors: Bateman FB , Boukharouba N , Brient CE , Carlson AD , Grimes SM , Haight RC , Massey TN , Wasson OA
Published: April 2008

Measurements have been made of the shape of the hydrogen differential neutron scattering cross section standard at 10 MeV neutron energy. The data were obtained at the Institute of Nuclear and Particle Physics at Ohio University using the D(d,n) reaction as the neutron source. Two different thin films of polypropylene mounted on tantalum backings were used as the hydrogenous samples in this experiment. Preliminary data obtained with the thicker film will be reported here. Two independent computer-based data acquisition systems were used to obtain the data which were stored in a multi-parameter mode so that analysis could be done under different conditions after the experiment was finished. The differential cross section was measured simultaneously at 11 laboratory angles (0°, ± 12°, ± 24°, ± 36°, ± 48° and ± 60°) using ΔE-E solid state detector telescopes in an evacuated scattering chamber. The data are in better agreement with the Arndt evaluation than the ENDF/B-VI and ENDF/B-V evaluations.
Fission Cross Section Measurements of Th-229 and Pa-231 Using Linac-Driven Lead Slowing-Down Spectrometer
Authors: Kobayashi K , Yamamoto S , Kai T , Cho HJ , Yamana H , Fujita Y , Mitsugashira T , Kimura I
Published: April 2008

Making use of a back-to-back type double fission chamber and Kyoto University Lead slowing-down Spectrometer (KULS) driven by a 46 MeV electron linear accelerator (linac) at the Research Reactor Institute, Kyoto University (KURRI), the neutron-induced fission cross sections of Th-229 and Pa-231 have been measured relative to the 235U(n,f) reaction from ∼0.1 eV to 10 keV with energy resolution (FWHM) of about 40%. Each of the measured results has been compared with the evaluated data in ENDF/B-VI, JENDL-3.2 and JEF-2.2, whose data were broadened by the energy resolution of the KULS. The experimental data for the fission cross sections of Th-229 and Pa-231 have scarcely been obtained before in the relevant energy region.
Activity Determination with High Precision for Component Disposal
Authors: Polke E
Published: April 2008

The conditioning and disposal of radioactive components is subject to limits not only in terms of weight and volume but also in terms of radioactivity due to transport and storage conditions. It is reasonable to make the maximum use of all limits in order to minimize the disposal costs. This means that the radioactivity of the components should be determined as exactly as possible with no conservative estimates. The radioactivity of components to be disposed of comprises the radioactivity generated by the interaction of neutrons with the material and deposits of radioactive particles on the component surface (crud). While crud determinations are performed by sampling and analysis, the radioactivity of the base material is calculated nuclide-specific based on the material composition, the spectral neutron flux density distribution at the irradiated site, the irradiation time and the decay time. Material samples are taken from the components for individual verification of the radioactivity calculations.
Improved Covariance Analysis and Spectrum Adjustment for VVER-1000 Pressure Vessel Fluences
Authors: Boehmer B , Borodkin GI , Manturov GN
Published: April 2008

The assessment of the uncertainties of neutron transport calculation results caused by uncertainties of the input data is important for the direct use of the calculated fluences and even indispensable for the further improvement of the data by spectrum adjustment on the basis of experimental results. A former work of two of the authors has approached this topic first for a VVER-1000 using a one-dimensional reactor model and applying spectrum adjustment in a single spectrum approximation. In the present work 2D/lD-SN-synthesis calculations have been used to determine the neutron group spectrum covariances including cross-correlations between interesting positions. As possible sources of uncertainties neutron data and reactor model parameters were considered, whereas the uncertainties caused by the calculational method were considered negligible in view of the now available precision of Monte Carlo simulations. The input data uncertainties have been updated including newer information.

The calculated absolute group fluence spectra have been adjusted to measured exvessel cavity activation rates with the help of the COSA3 generalized least squares multi spectrum adjustment code. Adjusted fluences for different detector positions have been compared with corresponding fluences obtained with single spectrum adjustments. The results are discussed and conclusions have been drawn. Results obtained with Russian LUND- and ENDF/B-VI-based cross section and fission spectrum covariance data have been compared.
BWR Neutron Fluence Computations using MF3D
Authors: Sitaraman S , Chiang R-T , Kruger RM , Jenkins AL , Asano K , Koyabu K
Published: April 2008

A full three dimensional model in a quadrant of an operating BWR has been modeled using MCNP to calculate flux spectrum and fluence levels at various locations in the reactor system. The model is being benchmarked against test data obtained over a complete fuel cycle of the host BWR. Preliminary comparisons of the calculated data to measured data are very encouraging.
The PKA Energy Spectrum Analysis of Defect Structures in Neutron Irradiated Metals
Authors: Yoshiie T , Xu X , Xu Q , Yanagita S , Satoh Y
Published: April 2008

The concept of displacement per atom (DPA) is not sufficient to account for the irradiation damage in metals. It ignores the distribution of displaced atoms in a cascade and subsequent point defect processes. PKA energy spectrum analysis has been proposed for the understanding of the defect evolution during cascade damage. In the analysis, special attention was paid to the production rate of subcascades as an important factor. DPA rate and the subcascade production rate in iron were compared with five irradiation fields in the KUR, the JMTR and the JOYO. The fast neutron flux was proportional to DPA rate in these irradiation fields. As the threshold energy for subcascade formation increases, the subcascade production rate in the irradiation fields with soft neutron energy spectra decreases more than those with hard neutron energy spectra. The importance of the analysis was discussed and some examples for application were demonstrated.
U.S. NRC Embrittlement Data Base (EDB)
Authors: Wang JA , Pace JV , Rosseel TM
Published: April 2008

Large amounts of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) [1] is such a comprehensive collection of data resulting from the merging of the Power Reactor Embrittlement Data Base (PR-EDB) [2] and the Test Reactor Embrittlement Data Base (TR-EDB) [3]. Fracture toughness data were also integrated into Version 1 of the EDB. The EDB data files are in dBASE format and can be accessed with a personal computer using the DOS or WINDOWS operating system. A utility program has been written to investigate radiation embrittlement using this data base. The utility program is used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data.
Selective and Prompt Self-Powered Neutron Detectors for Characterization of Mixed Radiation Fields in Reactors
Authors: Blandin C , Breaud S
Published: April 2008

This paper presents new experimental developments of prompt self-powered neutron detectors (platinum, hafnium, gadolinium). In particular, we covered the Gd and Hf emitters with a small stainless steel tube called “the lining of the emitter,” to reach three goals: to decrease the deferred noise due to the sheath, reduce the current due to low energy external gamma-rays and stop a large part of the delayed β- rays, which are produced in the emitter. A large set of SPNDs with different thickness of lining was tested in the SILOE research reactor by means of two special dosimetry devices to study their neutron and gamma sensitivities, and their dynamic characteristics. Finally, we optimized three physically prompt neutron detectors (response times better than 0.03 s) with known sensitivities. We are now able to measure instantaneously in-line the thermal and epithermal neutron fluxes and the gamma field.
A Laser Compton-Scattered Photon Source for Potential Calibration of Gamma-Ray Detectors in the Energy Range from 2 to 22 MeV
Authors: Kudo K , Takeda N , Ohgaki H , Toyokawa H
Published: April 2008

A laser-Compton-scattered (LCS) photon beam was used as the source of quasimonoenergetic photons in the energy range from 2 to 22 MeV. Second-harmonic light (527 nm) from a Nd:YLF laser was guided into the interaction region in a electron storage ring, TERAS (800 MeV, 300 mA). High energy photons back-scattered by relativistic electrons were collimated by a lead collimator with a reentrant hole 2 mm in diameter. The photon fluence was measured by a large coaxial Ge detector. To determine the energy distribution of a photon beam and the response function of a Ge detector, the EGS4/PRESTA-CG code was used taking into account the precise overall computational model. Results showed that, at a photon energy of 5.965 MeV, the calculated pulse height spectra agreed very well with experimental data.
Qualifying C8/F9 Spectrum Index Measurements by Using Standard Neutron Fields
Authors: van der Meer K , D'hondt P , Vandebroek R
Published: April 2008

A measurement method for the determination of the spectrum index C8/F9 and, indirectly the F8/F9 spectrum index in LWR fuel assemblies has been qualified. With this method U- and Pu-foils are placed between fuel pellets and irradiated in the VENUS test reactor. After irradiation the foils are dismantled and the produced activation and fission products are measured with the help of gamma-spectrometry. The amounts of activation and fission products are a measure of the reaction rates. A qualification of this method has been performed in the BR1 reactor in a well-known reference neutron spectrum (thermal). The observed deviation from the theoretical value (1.7%) falls within the uncertainty range (1.9%, 1σ). In this paper the qualification of the measurement technique, together with the uncertainty analysis, will be reported.
Metal Discs as Very Low Thermal Neutron Flux Monitors in Reactor Environment
Authors: Martínez Canet MJ , Hult M , Köhler M , Aït Abderrahim H , Marloye D
Published: April 2008

Metal discs of pure materials like Au, W, Ta, Dy, Sm, In, Re and Mn can be used as monitors of very low thermal neutron fluxes. In order to reach low detection limits, the discs were activated for about 3 half-lives and measured with ultra low-level gamma-ray spectrometry for another 3 half-lives in an underground laboratory located at a depth of 225 m. The minimum detectable activity for these discs is in the range 0.5 mBq to 50 mBq, which was found through calibration to correspond to thermal neutron fluxes in the range 10-4 to 10-2 cm-2 s-1.

The discs were calibrated in the well defined neutron flux of the 1-m diameter large cavity facility located in the thermal column of the BR1 reactor at SCK•CEN. Thermal and 235U fission spectrum neutron standard fields were used in the calibration procedure.

The objectives of this technique, the calibration results as well as the first monitoring of thermal neutron fluxes at work places outside the BR1-reactor are presented.
Development of a Neutron and Gamma-Ray Flux Distribution Measurement System with Scintillator and Optical Fiber Combination
Authors: Mori C , Uritani A , Iguchi T , Hayashi S , Kobayashi H , Takami Y , Kimura I , Katagiri M , Kakuta T
Published: April 2008

for the real time measurement of neutron flux and gamma-ray intensity distribution in reactor cores, a system composed of scintillator-optical fiber detector and fiber driver was newly tried. A pulse counting system that we successfully developed before was difficult to use in such an intense radiation field in the core, because there were very intense output of both scintillation light from the scintillator and Chernkov light in the optical fiber itself. The current from the photomultiplier was therefore measured this time. To distinguish the scintillation light component with longer wave length, 690 nm, from the Cherenkov light component with shorter wave length, an optical filter was used which had the cutoff wave length, 680 nm, just shorter than the scintillation wave length. However it was not possible to completely eliminate the Cherenkov component, therefore the difference between the signal from a fiber with the scintillator on the tip and the signal from the fiber without the scintillator was taken.
Neutron Dosimetry Using Diallyl Phthalate Resin
Authors: Tsuruta T
Published: April 2008

Diallyl phthalate resin plates showed favorable characteristics as fission track detectors. The detection efficiency of fission fragments was about 100% for both perpendicular and random incidence. DAP plates were insensitive to alpha particles and fast neutrons. The plates possessing fission tracks were etched after heat treatment or after gamma-ray irradiation. Heat treatment above 160°C lasting for 1 hour and irradiation above 10kGy increased the etch-pit diameter. A reduction of the track detection efficiency was observed in the plates exposed to heat treatment of 260°C lasting for 1 hour and irradiations of 2MGy and above. These phenomena were observed in the study of plates in which fission tracks were made after the above- mentioned heat treatment or irradiation. These results suggest not only the conditions under which the plates may be used, but also the possibility of estimation of a detector's thermal history or gamma-ray dose.
Development and Modeling of Neutron Detectors for In-Core Measurement Requirements in Nuclear Reactors
Authors: Blandin C , Bignan G , Lebrun A , Poujade O , Recroix H , Eymery S , Trapp J-P , Breaud S
Published: April 2008

The Reactor Measurement Systems Laboratory (LSMR) of the CEA/Cadarache, is specialized in the development, design, modeling and manufacturing of neutron detectors, especially miniature fission chambers.

This laboratory can define the characteristics of the detector for any requirements : - type of chamber (cylindrical or plate), - diameter (1.5 mm, 4 mm, 8 mm, …), - nature of fissile product (uranium, plutonium, neptunium, curium,…), - mass of fissile product, and - nature and pressure of gas (argon, helium).

The laboratory has also developed modeling codes and in particular, a new theoretical method to foresee the current-voltage characteristics of a chamber that operates in current mode (sensitivity, saturation plateau).
A System Design for Measuring Boron Concentration in Pressurized Light Water Reactor Cooling Water
Authors: Oda K , Kojima T , Nakagawa K , Yamauchi T , Ohashi S
Published: April 2008

A new technique has been proposed for measurement of the boron concentration in primary cooling water in Pressurized Light Water Reactors (PLWRs). This designed system consists of two types of photon detectors and a compact external neutron source. The boron concentration is, in principle, determined by the count rule of 0.48 MeV prompt γ-rays emitted after 10B(n,α)7Li* reactions induced by thermal neutron irradiation, diving by that of 2.2 MeV capture γ-rays from 1H(n,γ) reactions. In the first design, a cadmium zinc telluride (CZT) semiconductor and a bismuth germante (BGO) scintillator were selected as promising detectors and their response functions were evaluated by Monte Carlo calculations. Pulse height distributions have been simulated considering the background photons emitted from a neutron source and radioactive impurities in cooling water. The error in evaluating the boron concentration was also estimated; for example, the concentration of 100 ppm with 1% error might be measured within 20 seconds in a thermal neutron field of 5×104 cm-2s-1.
Synthetic Bioabsorbable Polymers for Implants
Authors: Agrawal CM , Parr J , Lin S
Published: April 2008

Description

12 papers provide an overview of the use of synthetics bioabsorbable polymers and address the latest issues surrounding their growing use including: new applications, test and characterization techniques for raw materials and devices, tissue engineering, efficacy, and long-term effects.

A variety of papers examine: basic polymer properties and characterization, tissue engineering strategies such as implantation of a porous biodegradable scaffold at the defect site in the tissue, and use of bioabsorbable polymers as replacements for metallic fracture fixation devices.
Irradiation Creep and Growth Behavior, and Microstructural Evolution of Advanced Zr-Base Alloys
Authors: Gilbon D , Soniak A , Doriot S , Mardon J-P
Published: April 2008

This paper deals with the irradiation-induced changes in the microstructure of FRAMATOME advanced Zr-base alloys and the correlations with their irradiation creep and growth behavior. The first part is dedicated to experimental irradiations performed at 280 and 350°C in a CEA metallurgical test reactor (Siloé, 1 to 1018 n/m2 s, E > 1 MeV) on stress-relieved (SRA) and recrystallized (RXA) low-tin Zircaloy-4 and two advanced RXA materials (Alloy 4 (M4): Zr-SnFeV, and Alloy 5 (M5): Zr-NbO), which are proposed for fuel rod cladding applications in PWRs. The irradiation creep results confirm the improved behavior of RXA Zy-4, M4, and M5 in comparison to that of SRA Zy-4. Similar trends are observed for irradiation growth, the SRA Zy-4 exhibiting a quasi-linear behavior with increasing fluence while RXA alloys undergo an early saturation phenomenon. Among RXA materials, M5 has the higher irradiation growth resistance. These creep and growth results at moderate neutron fluences (<5 × 1025 n/m2) are corroborated by the total elongation measurements of SRA Zy-4 and RXA M4 and M5 claddings irradiated up to 5 cycles (≈10 X 1025 n/m2) in a PWR.

The second part relates to the microstructural evolution of these Zr-base alloys irradiated as growth specimens in MTR or claddings (SRA Zy-4, M4, and M5) and guide tubes (RXA Zy-4) in PWRs. TEM examinations were performed with special emphasis on the evolution of dislocation loop microstructures, Fe- and Nb-rich phases, and alloying element redistribution with irradiation temperature and neutron fluence. Moreover, the comparison between samples taken from the higher part of a RXA Zy-4 guide tube irradiated five cycles and the lower parts of M4 and M5 claddings irradiated four cycles provides relevant information for similar irradiation temperatures (≈320 and 330°C) and fluences (≈9.5 and 7.2 × 1025 n/m2, respectively). Thus, in M4 the Zr(Fe,V)2 Laves phases undergo a partial crystal-to-amorphous transformation that is slower and coupled to a significantly lower Fe-redissolution than the Zr(Fe,Cr)2 Laves phases in Zy-4. Consequently, basal ⟨c⟩-component loops are detected only in the vicinity of intermetallic particles in M4, while they are distributed uniformly throughout the matrix in RXA Zy-4. As for M5, the β-Nb particles undergo neither radiation-induced amorphization nor dissolution, and the Fe content is too low to stabilize a significant density of basal ⟨c⟩-component loops. This excludes the possibility of reaching an accelerated growth regime for the fuel rod burnup achievable in PWR.
Irradiation-Enhanced Deformation of Zr-2.5Nb Tubes at High Neutron Fluences
Authors: Causey AR , Holt RA , Christodoulou N , Ho ETC
Published: April 2008

The effects of neutron flux and temperature on irradiation creep of 10-mm- diameter Zr-2.5Nb tubes are being assessed in end-of-life experiments in the Osiris reactor in France. The tubes have crystallographic texture, grain shape, and dislocation densities that are similar to those of full-size CANDU reactor pressure tubes. The experiments are performed at nominal operating temperatures of 553 and 583 K in fast neutron fluxes up to 2.0 × 1018 n · m-2 · s-1 (E > 1 MeV). Transverse and axial strains are reported as functions of fluence for specimens internally pressurized to hoop stresses from 0 to 160 MPa and irradiated to fluences up to 1.5 × 1026 and 1.3 × 1026 n · m-2 at 553 and 583 K, respectively. The measured transverse and axial strain rates in these high-fluence experiments are shown to be in good agreement with a creep model relating crystallographic texture to the in-reactor deformation of CANDU pressure tubes. The anisotropy of irradiation creep does not change with fluence, but the temperature dependencies of creep and growth are changing; the activation temperature for creep is increasing, while the magnitude of the activation temperature for irradiation growth (a negative value) is becoming less negative with fluence.
Effect of Long-Term Irradiation on the Fracture Properties of Zr-2.5Nb Pressure Tubes
Authors: Hosbons RR , Davies PH , Griffiths M , Sagat S , Coleman CE
Published: April 2008

Results from fracture toughness and tensile and delayed hydride cracking (DHC) tests on Zr-2.5Nb pressure tubes removed from CANDU power reactors in the 1970s and 80s for surveillance showed considerable scatter. At that time, the cause of the scatter was unknown and prediction of fracture toughness to the end of the design life of a CANDU reactor using the surveillance data was difficult. To eliminate the heat-to-heat variability and to determine end-of-life mechanical properties, a program was undertaken to irradiate, in a high-flux reactor, fracture toughness, DHC, and transverse tensile specimens from a single “typical” pressure tube. Two inserts were placed in the OSIRIS reactor at CEA, Saclay, in 1988. Each insert held 16 of each type of specimen. The first insert, ERABLE 1, was designed so that half the specimens could be replaced at intervals and the properties could be measured as a function of fluence. All the specimens would be removed after a total fluence of 15 × 1025 n · m-2, E > 1 MeV. The second insert, ERABLE 2, was designed to run without interruption to a fluence of 30 × 1025 n · m-2, the fluence corresponding to 30 years' operation of a CANDU reactor at 90% capacity factor. The irradiation temperature was chosen to be 250°C, the inlet temperature of early CANDU reactors. The irradiation of ERABLE 1 has been completed and sets of specimens have been removed and tested with maximum fluences of approximately 0.7, 1.7, 2.8, 12, and 17 × 1025 n · m-2, E > 1 MeV. X-ray and TEM examinations have been performed on the material from fractured specimens to characterize the irradiation damage. Results showed that there is, initially, a large change in the mechanical properties before a fluence of 0.6 × 1025 n · m-2, E > 1 MeV (corresponding to an initial rapid increase in a-type dislocation density), followed by a gradual change. As expected, the fracture toughness decreased with fluence, whereas the yield strength, UTS, and DHC crack velocities all increased. Z-ray analysis showed that, although the a-type dislocation density remained constant after the initial increase, the number of c-component dislocations showed a steady increase, agreeing with the behavior seen in the mechanical specimens. Because the flux in OSIRIS is different from that in a CANDU reactor, specimens were also irradiated in NRU, a heavy water moderated test reactor with approximately the same flux as a CANDU reactor, to fluences of 0.3, 0.6, and 1.0 × 1025 n·m-2, E > 1 MeV for comparison. These initial results show that, once past the initial transient, one can have confidence that there will be little further degradation with fluence, with the results from the NRU specimens being similar to those from OSIRIS.
The Long-Range Migration of Hydrogen Through Zircaloy in Response to Tensile and Compressive Stress Gradients
Authors: Kammenzind BF , Berquist BM , Bajaj R , Kreyns PH , Franklin DG
Published: April 2008

Zircaloy-4, which is used widely as a core structural material in pressurized water reactors (PWRs), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and zirconium hydride phases precipitate after the Zircaloy-4 lattice becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4, degrading its mechanical performance as a structural material. Because hydrogen can move rapidly through the Zircaloy-4 lattice, the potential exists for large concentrations of hydride to accumulate in local regions of a Zircaloy component remote from its point of entry into the component. Much has been reported in the literature regarding the long-range migration of hydrogen through Zircaloy under concentration gradients and temperature gradients. Relatively little has been reported, however, regarding the long-range migration of hydrogen under stress gradients. This paper presents experimental results regarding the long-range migration of hydrogen through Zircaloy in response to both tensile and compressive stress gradients. The importance of this driving force for hydrogen migration relative to concentration and thermal gradients is discussed.
Influence of Cladding Microstructure on the Low Enthalpy Failures in RIA Simulation Tests
Authors: Garde AM
Published: April 2008

Welding used during the preparation of specimens for Reactivity Initiated Accident (RIA) simulation testing from fuel rods previously irradiated in reactors introduces the following four microstructural changes in the specimen: (a) annealing of irradiation damage, (b) a change from an α-phase structure of Zircaloy to a transformed β structure in the cladding, (c) dissolution of the hydride rim formed under the oxide on the cladding tube outer surface during normal irradiation and possible radial-oriented hydride reprecipitation at the transformed β platelet boundaries, and (d) reprecipitation of second-phase particles previously dissolved due to radiation damage. A fifth factor, the change in the texture of Zircaloy, is also introduced due to the welding operation. The possible effect of these five changes on the specimen fracture toughness and failure enthalpy is evaluated in this paper.

The published data on the mechanical properties of irradiated and unirradiated transformed beta structure of Zircaloy charged with hydrogen are reviewed to evaluate the impact of the five anticipated microstructural changes on the failure enthalpy. The available RIA simulation test fracture data with low-failure enthalpy are reviewed. The limited failure path information available appears to indicate that microstructural factors have contributed to the low-enthalpy failures.

The applicability of the results from the low-enthalpy RIA test failures to Light Water Reactor (LWR) nuclear fuel should be based on the representativeness of the RIA specimen microstructure to that of the LWR fuel cladding.
Size, Geometry, and Material Effects in Fracture Toughness Testing of Irradiated Zr-2.5Nb Pressure Tube Material
Authors: Davies PH , Shewfelt RSW
Published: April 2008

The effect of initial crack size on the crack growth resistance (J-R) curves determined from burst tests on irradiated Zr-2.5Nb pressure tubes has been studied and the results compared with those obtained from matched curved compact specimens. The study used sections from three different tubes representative of material of low, intermediate, and high toughness. In each case a series of burst tests was conducted at 250°C with different starting crack sizes (from 35 to 85 mm) followed by small specimen testing. The toughness was characterized by means of deformation J-R curves using the d-c potential drop method to measure stable crack growth. Fractographic studies were also conducted in support of the J-R curve results.

For tubes of low to intermediate toughness there is little evidence of a crack size effect on the J-R curves from the burst tests. However, for burst tests on tubes of higher toughness there is an increase in out-of-plane bending associated with the wider crack openings (bulging) that can promote earlier failure by slant/shear instability and a lowering of the J-R curve. Thus the J-R curves from such tubes exhibit more variability in toughness as well as an apparent crack size dependence. Such crack growth behavior overrides any increase in material toughness with irradiation temperature as revealed by the small specimens.
In-Process Investigation of Precipitate Growth in Zirconium Alloys
Authors: Herb B , Ruhmann H , König A
Published: April 2008

The nucleation and growth of second-phase precipitates was investigated in ELS (Extra Low Sn-Sn below, Fe and Cr above ASTM specification) and Zircaloy-4 alloys used for DUPLEX cladding. DUPLEX cladding is typically used for the most demanding irradiation conditions in reactors with high power densities, coolant temperatures, and discharge burnups. Extensive in-process sampling was performed after the β-quenching, hot working, and α-annealing steps during manufacturing. SEM and TEM evaluations were performed to characterize the precipitate distribution and composition.

The β-quench step produces a significant number of fine precipitates along the grain boundaries during the β → α+β phase transformation. Intergrain nucleation of irregular-shaped precipitates was also observed. The subsequent hot-working steps in the α-phase tend to increase the uniformity and size of the precipitates. Zircaloy-4 materials processed with a forge-Quench-Forge (FQF) process generally have a log-normal precipitate distribution during the early processing steps. However, the log-normal distribution transforms into a bimodal distribution after α-annealing. The low-Sn ELS alloy shows broader distributions very early in the process with two or even three Gaussian peaks needed to describe the distribution curve. TEM evaluations suggest that the small particles in the bimodal distribution are a Zr-Si-Fe phase with very slow growth kinetics. The larger particles are the typical Laves phase Zr(Fe,Cr)2 with significantly faster growth rates. Separation of these two precipitate classes allows for a more accurate correlation to the Particle Growth Parameter (PGP) annealing parameter concept.

The precipitate distributions from FQF processing were also compared to a similar Zircaloy-4 alloy fabricated with a Forge-Forge-Quench (FFQ) process. This material was β-quenched and then extruded directly without α-phase forging.
E110 Alloy Cladding Tube Properties and Their Interrelation with Alloy Structure-Phase Condition and Impurity Content
Authors: Shebaldov PV , Peregud MM , Nikulina AV , Bibilashvili YK , Lositski AF , Kuz'menko NV , Belov VI , Novoselov AE
Published: April 2008

The E110 alloy (Zr-1Nb) designed in Russia has been in successful operation in the VVER and RBMK reactors since 1960. The evolution of the Russian reactors during this period of time, their numbers and operating parameters, and, hence, the relevant production of the cladding tubes required much effort from the researchers and producers to optimize the technological processes resulting in the high operating reliability of the fuel rods clad in the E110 alloy.

In the Zr-Nb system to which the E110 alloy pertains, highly sophisticated structure-phase transformations proceed depending on the temperature-rate conditions of processing and the impurity content of the alloy. The alloy structure formed during its processing affects substantially the processability and performance of semi-products and finished tubes. The paper demonstrates the interrelation between the structure-phase transformations of the E110 alloy and the temperature-rate conditions of processing as well as the oxygen content of the alloy. The phase composition and microstructure are shown to play an essential part in providing the tubes with the requisite properties and their stability. The properties of cladding tubes having different structures and fuel rod claddings as in-pile irradiated are analyzed. The corrosion behavior of E110 alloy claddings is shown to be strongly dependent upon the material structure and reactor operating conditions.
The Effect of Microstructure on the Corrosion Behavior of Zircaloy-2 in BWRs
Authors: Etoh Y , Nonaka Y , Kubo T , Nomata T , Hayashi H , Kitamura M
Published: April 2008

For improved Zircaloy-2 cladding tubes used in Japanese BWRs, α+β heat treatment is used on the outer surface of the tubeshell to suppress nodular corrosion. This treatment also provides superior uniform corrosion behavior up to 40 GWd/t. To investigate the effect of α+β heat treatment on corrosion behavior, the size distributions of second-phase particles in improved Zircaloy-2 cladding tubes and their changes during irradiation were observed and measured using an analytical transmission electron microscope. These cladding tubes were found to have smaller particles than previous ones. The dissolution rate of iron and nickel (Fe and Ni) was calculated as a function of particle size, chemical compositions of cladding tubes, and fast neutron flux. The relationship between corrosion behavior and the dissolution rate of Fe and Ni is described. Nodular corrosion occurred when the dissolution rate of Fe and Ni was slow.
Evaluation of Zircaloy-2 Cladding Corrosion Characteristics by Simulated BWR Corrosion Loop Test
Authors: Shimada S , Asahi K , Nonaka Y , Sasaki M , Kogai T , Hayashi H , Kitamura M , Sakamoto M , Yamawaki M
Published: April 2008

The effects of irradiation conditions such as neutron flux, β and γ rays, and of heat flux on the corrosion behavior of Zircaloy-2 cladding tubes were studied by using an inreactor corrosion loop simulating BWR conditions. Platinum and hafnium rings were placed in the rigs as β and γ emitters, respectively, to evaluate local radiation effects on cladding corrosion. The tube materials included an early BWR 8 x 8RJ type and a new, more corrosionresistant BWR 9X9-type cladding. Both tubes contained UO2 and Al2O3 pellets and were irradiated under the same conditions in two rigs in the same loop for about 180 days. Nodular corrosion appeared on the 8X8RJ-type tube, while the 9X9-type tube presented only uniform oxide layer. Metallographic and SEM observations revealed that the β and γ emitters enhanced the uniform oxide formation for the 9X9-type tube, while they increased the nodule density for the 8X8RJ type. Computer analyses on neutron flux and β- and γ-ray intensities were conducted for the 8X8RJ-type and 9X9-type test rods. Those results were utilized to comprehensively discuss and to understand the in-pile corrosion behavior.
Behavior of Lithium and Boron in Irradiated and Unirradiated Oxides Formed on Zircaloy-4 Claddings
Authors: Kido T , Wada S , Takahashi T , Uchida H , Komine I , Inoue Y
Published: April 2008

In order to investigate the effect of lithium and boron on corrosion of Zircaloy-4 fuel cladding, secondary ion mass spectrometry (SIMS) analyses coupled with chemical analyses have been carried out on oxides of specimens corroded in water with lithium hydroxide and boric acid similar to the actual reactor conditions, and of the fuel claddings irradiated in a commercial reactor.

The quantitative analyses on lithium and boron by SIMS and chemical analyses showed that the concentration of lithium in the oxides formed ex-reactor was 2 to 50 ppm when the water contained 2 to 10 ppm lithium and exhibited a strong temperature dependence. The concentration of boron in the unirradiated oxides was approximately 250 ppm when water contained 500 to 1600 ppm boron and showed some increase as the concentration of boron in the water increased. No temperature dependence was detected. The lithium concentration of irradiated oxides was systematically higher than that of the unirradiated oxides, and little difference has been seen in the 11B concentration.

The SIMS analyses of the oxides of the ex-reactor corrosion specimens revealed that the depth profile of lithium concentration was flat in most of the oxide and exhibited a sharp decrease in the dense oxide layer, which corresponds to a thickness of approximately 2μm from the assumed oxide-metal interface. The boron distribution in the oxide was similar in general, but showed an obvious cyclic variation in the outer oxide region. The frequency of the variation corresponded to the rate transitions during the corrosion tests. These features in the profiles of lithium and boron have been also detected on the oxides of the irradiated claddings.

The sharp decrease in the concentrations of lithium in the vicinity of the metal-oxide interface indicates that the inner layer is a barrier against the deteriorating effect of lithium. The hydrogen pickup fraction of the specimens was found to be influenced by the boron concentration in the water.
Contribution to the Understanding of the Effect of the Water Chemistry on the Oxidation Kinetics of Zircaloy-4 Cladding
Authors: Pêcheur D , Godlewski J , Peybernès J , Fayette L , Noé M , Frichet A , Kerrec O
Published: April 2008

Zircaloy-4 cladding has been oxidized in autoclave at 633 K with various chemistry conditions ([Li+] = 0, 10, 70, 700 ppm and [B] = 0, 650 ppm as boric acid) to quantify the effect of lithium hydroxide and boric acid on the oxidation kinetics and to provide oxide films that have been analyzed by SIMS, SEM, and RAMAN spectroscopy to improve our knowledge of the oxidation process in a lithiated environment. Additional specific tests consisting of isotopic exchanges (6Li+/7Li+) have also been conducted to study the incorporation of lithium in the oxide films.

The main results are that: • Depending on the water chemistry, the lithium hydroxide has two effects on the oxidation kinetics: a decrease of the time to transition and a strong enhancement of the post-transition oxidation rate. • The boric acid strongly reduces these two effects of lithium hydroxide. • In the oxide film, the lithium is located mainly in the pores (or adsorbed on their walls). • The lithium has a quick access to the bulk of the oxide film, even in the inner barrier layer (considered as not completely impervious). • The enhancement of the oxidation rate due to lithium hydroxide is linked to the degradual alteration of the inner barrier layer, up to its quasi-disappearance.

These experimental results are described extensively, and several factors involved in the alteration of the inner barrier layer, such as lithium pickup in the ZrO2 film, evolution of the morphology of the oxide grains, tetragonal zirconia phase transformation, and hydriding, are then discussed.
Electrochemical Examinations in 350°C Water with Respect to the Mechanism of Corrosion-Hydrogen Pickup
Authors: Baur K , Garzarolli F , Ruhmann H , Sell H-J
Published: April 2008

During corrosion of Zr alloys in pressurized water at high temperatures a fraction of the corrosion-hydrogen is picked up by the metal. Long-term out-of-pile corrosion experiments have shown that chemical composition of Zr alloys and the size of second-phase particles (SPP) in Zircaloy-4 (Zry-4) affect the corrosion and the corrosion-hydrogen pickup fraction. The mechanism of hydrogen pickup is not well understood, although several influencing parameters were evaluated or discussed in the literature. One of the parameters that might influence hydrogen pickup is the electrical potential gradient that develops over the oxide during corrosion. Long-term electrochemical measurements of Zry-4 samples with different SPP sizes and Fe content and of Zr-2.5Nb in pressurized water at 350°C with and without polarization were used to check this influence.

The potential difference between the reaction interface and the oxide surface is due to the oxidation reaction of the Zr metal resulting in electrons that have to move through the highly resistive oxide to the surface. Tests without polarization showed the potential difference proportional to the corrosion rate and depending on metallurgical aspects as the alloy composition and the SPP size. The lowest potential difference has been found for Zry-type material with large SPP and for Zr-2.5Nb.

A negative polarization voltage of the samples against a Pt-reference electrode increases the H pick up and even leads to an accelerated corrosion at large potential differences. Analysis of H pickup clearly shows that, besides corrosion-H, H from the electrochemical surface reaction is also picked up.

Samples with oxide layers exhibiting high electrical resistance pick up relatively more H than samples exhibiting oxide layers with low resistance. Zr-2.5Nb forming a very low-resistant oxide layer picks up only very little H. The effect of the SPP sizes can, at least partially, be explained by their influence on the electrical resistance of the oxide layer.

The results of this study identify the potential gradient formed over the oxide layer as an important parameter for the relative amount of H pickup.
Role of Intermetallic Precipitates in Hydrogen Transport through Oxide Films on Zircaloy
Authors: Hatano Y , Sugisaki M , Kitano K , Hayashi M
Published: April 2008

The influence of size distribution of intermetallic precipitates on hydrogen uptake behavior of Zircaloy during oxidation in steam has been examined in the pre-transition region to clarify the role of precipitates in the hydrogen transport through the coherent oxide layer. Two different types of specimens containing coarse or fine precipitates are prepared with heat treatments for both Zircaloy-2 and -4. These specimens are oxidized in 0.1-MPa steam at 623 to 723 K. The amount of hydrogen uptake is measured with a hot extraction method. The concentration profiles of hydrogen in the oxide film is measured with secondary ion mass spectroscopy (SIMS). Detailed spatial distribution of hydrogen is observed with tritium microautoradiography.

The results obtained for Zircaloy-2 and -4 are qualitatively similar to each other: The amount of hydrogen uptake of the specimen containing coarse precipitates is larger than that of the specimen containing fine ones, and the difference between the former and the latter increases with decreasing oxidation temperature. The concentration of hydrogen in the oxide film is gradient from the surface to the oxide-metal interface, and the gradient in the oxide film of the former is smaller than that of the latter. The tritium microautoradiographs of a cross section of the oxide film show that hydrogen is concentrated in the intermetallic precipitates.

Summarizing these experimental results, the following can be concluded: 1. The rate of hydrogen uptake is controlled by the transport process through the oxide film. 2. The intermetallic precipitates accommodated in the oxide film play an important role. 3. Hydrogen is transported through not only the zirconium oxide phase but also the intermetallic precipitate phase, in which the latter acts as the supplementary expressway of the hydrogen transport through the oxide film.
What Does a Property Data Sheet Really Tell You?
Authors: Driscoll SB , Shaffer CM
Published: April 2008

Today's engineer, whether a student in training or a practicing professional, is faced with a myriad of design considerations when selecting the optimum material candidate for a functional product. The normal protocol followed includes prioritizing performance credentials, selecting candidate polymeric material, reviewing the properties and processing characteristics of each, and then making a qualified decision based on the available facts.

The problems, however, are both knowing fully what will be required of the product, in a wide range of use/abuse environments, and understanding as thoroughly as possible the true functional behavior of the polymer. What often is not completely understood is the correlation between published data sheets and the relevance to design considerations. How meaningful is the information found in a typical data sheet? How easily, if at all, can the design engineer integrate this information in a series of design iterations, leading to an improved product?

How useful is the information most routinely published? Certainly a great deal of information is readily available in a number of formats, including proprietary databanks maintained and freely distributed by the various material suppliers. This information is equally available via the Internet, and much follows the CAMPUS template containing both single- and multi-point data (International Standards Organization-ISO 10350 and 11403 respectively).

Regardless of the quantity and ready availability of this information, how can published Izod impact behavior be successfully used to make a better, more durable product? How can the Distortion Temperature Under Load (DTUL) of a material translate into practical continuous use temperature? What published data are really useful? And what are meaningless?

Finally, what really is needed — and is usually missing — to help the design engineer predictably produce a safe, reliable, and durable product?