bluuebunny/crossref_metadata_2025 huggingface dataset:
Effect of Freeze-Thaw on the Permeation of Arctic Diesel Through a GCL
Authors: Rowe RK , Mukunoki T , Li MH , Bathurst RJ
Published: April 2008

The hydraulic conductivity of a scrim reinforced nonwoven thermal locked geosynthetic clay liner (GCL) with respect to both de-aired distilled water and arctic diesel (jet fuel) are examined for different numbers of freeze and thaw cycles. The hydraulic conductivity of jet fuel is found to be about one order of magnitude smaller than that for de-aired water. Increasing numbers of freeze and thaw cycles, did not significantly change the hydraulic conductivity although the permittivity with respect to water decreased. The intrinsic permeability of the GCL permeated by jet fuel increased slightly with an increasing number of freeze-thaw cycles. It is shown that under the conditions examined, the GCL performed very well.
Contrast Enhancing Filters in Ski Sports
Authors: Lingelbach B , Jendrusch G
Published: April 2008

The need to visually recognize irregularities in the terrain, such as hummocks or hollows, is necessary for the skier to react appropriately. The state of the art is to use blue attenuator filters or grey filters as sunglasses. But without knowledge of the spectra, e.g., the surface of the slope, it is impossible to choose or develop an optimal filter. We measured the spectra of snow under various conditions. These data served as a base to develop a new kind of filter suitable for skiing. Some physiological evidence is given why this new kind of filter is optimal for skiing. The filters are already used by the German Ski Federation World Cup team (women) in both training and competition.
An Attempt to Predict a Risk of Some of the Knee Region Ski Injuries by Means of Computer Simulations
Authors: Kietlinski K , Rzymkowski C
Published: April 2008

The thesis presented tries to explain some aspects of the phenomenon of ski injuries and also shows the mechanisms through which there is high risk of soft tissue injury. Based on studies about possible injury mechanisms to the soft tissues of the knee joint, two common knee injuries were selected to be studied: “Two-skier impact” and “Boot Induced Anterior Drawer.” A multi-body skier model with flexible elements equipped with a FEM model of the knee joint was developed for the presented work. The solutions obtained confirmed by a majority that for the previously listed mechanisms a high risk of cruciate ligaments injury occurs, especially for the ACL. No damage to the collateral ligaments, the MCL and the LCL, was noticed in any of the cases considered.
A Comparison of East/West Steels for Pressurized Water Reactors
Authors: Davies M , Kryukov A , English C , Nikolaev Y , Server WL
Published: April 2008

Recent studies have shown that for the older East and West water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. In both cases with the identification of and a reduction in the deleterious element concentration the sensitivity of the steels to neutron irradiation was reduced. The deleterious elements which have been identified in the national databases as being responsible for the sensitivity are different. In the ‘Western’ case copper and nickel are identified and used in the current version of the USNRC Regulatory Guide while phosphorus and copper are identified in the Russian Codes. For the comparison neutron fluence is harmonized and the end-of-life fluence differences are considered. Annealing is a mitigating action for reducing the effect of irradiation on these steels and another coincidence in the similarity of the response of these steels is seen. The paper reviews the current understanding on the mechanisms of irradiation embrittlement of these steels and also the role of compositional variations in their irradiation sensitivity. Phosphorus concentration is seen as the key difference between East and West steels. When the phosphorus concentration is greater than about 0.012 wt%, then its effect is a dominating one with regard to irradiation effects.
Effect of Radiation on Material Used for 600MW Chinese Nuclear Power Plant Reactor Pressure Vessel
Authors: Jia X , Xu Y , Yang W , Zhang C
Published: April 2008

Irradiation experiments were conducted to determine the effect of radiation on the properties of 16MnD5 material from a simulate Chinese 600 MW nuclear power plant reactor pressure vessel ring forging. Irradiation experiments were conducted to compare the performance of the 16MnD5 samples to samples of American Society for Testing and Materials Specification A508-class 3 material. The results showed that the radiation effects on the 16MnD5 material were less than those exhibited by the A505-class 3 material.
Characterization of the Decommissioned Novovoronesh-1 Pressure Vessel Wall Materials by Through-Wall Trepans
Authors: Valo MJ , Amaev AD , Ahlstrand R , Chistiakov DA , Krasikov E , Kryukov A , Morosov AM , Nikolaev VA , Platonov P , Rintamaa R , Rybin VV , Shtrombakh Y , Wallin K
Published: April 2008

Material properties of the Novovoronesh-1 pressure vessel were characterized after 20 years of plant operation. The unit was the first commercial PWR unit in Russia producing 210 MW electric power. It was commissioned in 1964 and shut down in 1984. Wall thickness of the vessel is 120 mm, including 20 mm of cladding. The vessel contained only circumferential welds. The coolant temperature was 250°C. Nine through-wall trepans of outer diameter 120 mm were cut out from the vessel both from the weld seams and the base metal forging rings. The trepans were studied by RRC-KI, PROMETEY and VTT. The paper is focusing on the Charpy-V and fracture toughness test data of the critical core weld. Due to the lack of archive material the unirradiated material condition was estimated by a reference annealing procedure.
Supplementary Surveillance Program for Reactor Pressure Vessels of VVER-440/V-213C Type Reactors in NPP Dukovany
Authors: Brumovsky M , Erben O , Novosad P
Published: April 2008

Standard surveillance specimen programs for materials in the reactor pressure vessels of VVER-440/V-213 type reactors were designed with a high lead factor (approx. equal to 12 for base metal and 18 for weld metal). Thus, end-of-life neutron fluence levels are obtained within the first three years of operation. The maximum irradiation time is five years after which there is no active monitoring of neutron fluence inside reactor pressure vessels. Also, the distribution of neutron flux along surveillance specimen chains creates a large difficulty in grouping of specimens for static fracture toughness testing.

A reevaluation of the standard surveillance program for NPP Dukovany has been performed. This reevaluation includes additional irradiations and a very detailed analysis of all results, and provides a basis for a design of a Supplementary Surveillance Program. Its realization started in 1997 by loading the first chains with specimens.

The principles of this Program are explained in this paper. This program will provide: 1) new results under a realistically low lead factor, 2) new data on cladding properties under irradiation as well as after annealing and re-irradiation, and 3) data on the re-irradiation rate of vessel materials. Also the fourth part of the Program will provide a permanent neutron monitoring through the real vessel end-of-life. Within the whole program, the IAEA reference material JRQ will be also used, and this material will serve as a correlation monitor with other irradiation experiments.
Irradiation Behavior of Submerged Arc Welding Materials with Different Copper Content
Authors: Langer R , Bartsch R , Nagel G
Published: April 2008

In Germany there are two first generation commercial nuclear power plants in operation, Obrigheim NPP (KWO) and Stade NPP (KKS). The reactor pressure vessels of both plants were fabricated of forged rings, joined with submerged arc welds. These welds were performed with a flux of the first generation, produced in license of Linde, and with copper-coated wires. When the first results of the irradiation surveillance programs were available, it became obvious that the neutron-induced embrittlement of the welding materials would be higher than expected at design of the power stations in the late sixties.

To evaluate the neutron irradiation response, additional irradiation programs were performed. In the scope of these programs, submerged arc test welds were produced, corresponding in chemical composition, mechanical properties and Charpy toughness in the unirradiated condition of the core-welds of the two NPPs concerned. It was intended to assess also KIc, T-curves of the welding material established on fracture toughness tests of large specimens.

This report presents some results of unirradiated and irradiated tensile, Cv, KIc specimens and E208 drop weight specimens according to ASTM E208 taken from submerged arc welds with different copper content up to 0.42 wt.%. Irradiation was performed in the RPVs concerned and in the VAK RPV with different fluxes and fluences up to about 3.3E19 cm-2 (E > 1MeV).

The irradiation behavior of the welding material in general is shown, especially the influence of the copper content. At a copper content of about 0.30 wt. %. saturation in in the T41 shift was found. The results show no influence of the irradiation facilities and of the different fluxes. With results on irradiated Cv, E208 drop weight, and WOLX specimens the conservatism of the use of the T41 shift according to the RTNDT concept could be confirmed.
Effects of Irradiation on Crack-Arrest Toughness of a Low Upper-Shelf Energy, High-Copper Weld
Authors: Iskander SK , Nanstad RK , McCabe DE , Swain RL
Published: April 2008

The crack-arrest fracture toughness (Ka) of unirradiated and irradiated specimens machined from a low upper-shelf, high-copper weld was determined. The specimens were machined from the beltline weld of the Midland Unit 1 nuclear reactor pressure vessel (RPV). This weld was fabricated from one heat of weld wire and one lot of welding flux and bears the designation WF-70. It is of particular interest because it has been used in several RPVs that may still be in operation. The effect of irradiation has been measured in terms of the shift of the mean and lower-bound crack-arrest fracture toughness versus temperature curves. These shifts are compared to (1) 41-J transition temperature shift from Charpy V-notch (CVN) tests, ΔTT41-J; (2) shift in the Master Curve transition temperature, T0, at the 100 MPa√m level, ΔT0; and (3) shift of the “arrest force” Fa temperature determined from instrumented CVN testing, ΔTFa. Preliminary results indicate that the shift of Ka of this low upper-shelf energy weld is approximately the same as the shift ΔTT41-J, larger than ΔT0, and significantly larger than ΔTFa.
Effect of Location of V-Notch on Neutron Irradiation-Induced Shift in DBTT of HAZ of Welded Pressure Vessel Steel
Authors: Suzuki T , Itoh K , Naruse Y , Matsui H , Kimura A
Published: April 2008

The effect of the location of the V-notch in the heat affected zone (HAZ) on the ductile-brittle transition behavior of neutron-irradiated welded A533B pressure vessel steel has been investigated utilizing the small specimen technique. Standard size, one-third size and 1.5 mm size Charpy V-notch specimens were prepared by changing the distance from the fusion boundary, so that a V-notch was located in the position to be evaluated. Neutron irradiation was performed in JMTR at 290°C up to 0.75 and 1.3 × 1024 n/m2. In both the cases before and after the irradiations, ductile-brittle transition temperature (DBTT) is lower and the upper-shelf energy is higher for so-called HAZ(M1) specimen, which has a V-notch 1mm away from the fusion boundary, than base metal and deposit. However, the DBTT of the HAZ(M4) specimen, with a V-notch 4 mm away from the boundary, is higher than that of base metal. As for the ΔDBTT, the HAZ(M1) and HAZ(M4) specimens showed equivalent and considerably larger ΔDBTT, respectively, than the base metal specimen; hence the HAZ(M4) specimen shows rather higher DBTT, that is, lower toughness. Micro-Vickers hardness measurements revealed that the M1 and M4 site showed the highest and lowest hardness, respectively.
Fracture Toughness Test Results of Precracked Charpy Specimens of Irradiated 73W Weld Material
Authors: Chaouadi R , van Walle E , Fabry A , Scibetta M , Van de Velde J
Published: April 2008

This work investigates the use of precracked Charpy specimens to predict fracture toughness behaviour of reactor pressure vessel steels. Fracture toughness measurements were performed on the 73W weld material previously tested by Oak Ridge National Laboratories (ORNL). Charpy-size specimens were tested in three-point bend and compared with the original data obtained on large compact tension (CT) samples. Fracture toughness tests were performed in both unirradiated and irradiated (Tirrad≈288°C and Φ≈1.5 1019 n/cm2, E>1MeV) conditions. Most of the Charpy size samples were reconstituted from broken CT specimens. Fracture toughness data obtained using Charpy size samples are in good agreement with results from large CT samples. Moreover, the various temperature shifts measured by the Charpy impact test data are in agreement with the fracture toughness shifts. Reconstitution results in a decrease of the upper shelf energy but has little effect on the other transition parameters.
Fracture Toughness of Submerged-Arc Weld Samples from a Decommissioned Magnox Reactor Pressure Vessel: Initial Results
Authors: Bischler PJE , Wootton MR , Bolton CJ
Published: April 2008

BNFL Magnox Generation Group is sampling irradiated submerged-arc weld metal from the reactor pressure vessel at Trawsfynydd Power Station, which is currently being decommissioned. Chemical analysis has shown that the sampled weld is typical of previously sampled submerged-arc weld. There are no significant variations either through the weld thickness or along the weld, indicating high quality and uniformity of construction. Dosimetry measurements are in line with predictions, and show significant attenuation through the vessel thickness. Fracture toughness measurements are being made using pre-cracked Charpy specimens machined from four different depths in the through-wall thickness, with reconstituted material from failed Charpy halves being used to augment the test data. The data from the layer of the first weld furthest from the reactor core have been compared with predictions made by the methodology used in assessing continuing RPV integrity, and support the methodology.
Effects of Irradiation and Thermal Annealing on Fracture Toughness of the Midland Reactor Weld WF-70
Authors: McCabe DE , Nanstad RK , Sokolov MA
Published: April 2008

The influence of irradiation damage and recovery of properties by thermal annealing for low upper-shelf WF-70 weld metal has been evaluated. The test material was obtained from the beltline and nozzle course girth welds of the Midland Unit 1 reactor pressure vessel. The two welds, ostensibly made with the same heat of weld wire and the same lot of submerged-arc flux, were determined to be separable for evaluations by virtue of differing copper contents. Conventional transition temperature methods of drop-weight nil-ductility transition and Charpy V-notch (CVN) transition curves could not distinguish a difference in transition temperature. However, a recently developed master curve method of data evaluation revealed the difference by virtue of achieving better sensitivity using only fracture mechanics-based test data.

Irradiation embrittlement after a fluence of 1.0 × 1019 n/cm2 was evaluated by both the conventional American Society of Mechanical Engineers (ASME) Code method and by the more recent American Society for Testing and Materials (ASTM) master curve method. Again, the Code method rated the postirradiation transition temperatures of the two materials the same. The master curve method indicated that the postirradiated properties of the two materials differed significantly. The U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, reasonably predicted the experimentally obtained CVN transition temperature shift at the 41 J reference level for the beltline weld.

Postirradiation annealing at 454°C (850° F) for 168 h almost completely restored the as-received transition temperature properties. The recovery of the transition temperature compared to unirradiated values after 343°C (650°F) for 168-h recovery anneal was 52% by the 41 J CVN temperature, which conflicts with the theoretical recovery of 17% calculated by NRC Regulatory Guide 1.162. The ΔTo from the master curve method indicated 21% recovery of transition temperature.
Grain Boundary Phosphorus Segregation under Neutron Irradiation and Aging in 2.25Cr1Mo Steel
Authors: Song S , Faulkner RG , Hewitt PEJ , Marmy P , Victoria M
Published: April 2008

Impurity segregation to grain boundaries under irradiation is one of the important radiation effects on materials. For this reason, phosphorus segregation to prior austenite grain boundaries in a P-doped 2.25Cr1Mo steel subjected to neutron irradiation and aging is examined using field emission gun scanning transmission electron microscopy (FEGSTEM) with energy dispersive X-ray microanalysis (EDX). The steel samples are irradiated and aged around 270 °C and 400 °C, respectively. The irradiation dose rate and dose are $#x223C;1.05×10-8 dpa/s and $#x223C;0.042 dpa respectively for 270 °C irradiation, and $#x223C;1.75×10-8 dpa/s and $#x223C;0.13 dpa respectively for 400 °C irradiation. The FEGSTEM results indicate that there is no apparent phosphorus segregation during 270 °C irradiation but there is some during 400 °C irradiation.
Radiation-Induced Grain Boundary Segregation of Phosphorus in ASME SA533B Ferritic Pressure Vessel Steel
Authors: Meade D , Faulkner RG , Ellis D
Published: April 2008

The segregation of phosphorus to grain boundaries is well known to cause embrittlement in a range of ferritic pressure vessel steels. Modelling of segregation is therefore useful in predicting the service life of pressure vessels, particularly in those pressure vessels used in nuclear reactor power generating plant where significant cost savings can be made. In this paper, predictions of the radiation induced grain boundary segregation of phosphorus in ASME SA533B pressure vessel steels are made using an analytical model for impurity segregation in dilute ternary alloys. The model takes into account equilibrium-type segregation, together with a newly adopted approach incorporating the site competition effects that exist between phosphorus and carbon. The site competition effect causes the predicted phosphorus segregation to fall to levels significantly lower than the predictions obtained from models of dilute binary alloys. The predictions are compared to FEGSTEM results of samples irradiated at between 255°C and 315°C, doses between 1.5 and 38mdpa and dose rates ranging from 6×10-9 to 8.9×10-11 dpa/s. The predictions obtained from the model are in close agreement to the experimentally observed segregated phosphorus results.
Comparison of Observed and Predicted Data on Radiation Induced Grain Boundary Phosphorus Segregation in VVER Type Steels
Authors: Nikolaeva AV , Kevorkyan YR , Nikolaev YA
Published: April 2008

The influence of impurity grain boundary segregation on mechanical properties of low-alloy Cr-Ni-Mo steels is discussed. Mechanisms of impurity and alloying effects on grain boundary embrittlement of low-alloy steels are considered. An approach for estimation of radiation enhanced grain boundary segregation kinetics is proposed. The approach takes into account the reduction of dislocation density (from 1×1014 to 1×1013 m-2) due to irradiation that was observed for low-alloy Cr-Ni-Mo steels. The influence of collision cascades on the radiation damage is also considered. Auger spectroscopy is used for grain boundary segregation detection. The contribution of grain boundary embrittlement to mechanical property degradation due to neutron irradiation is assessed.
Annealing of a Low Copper Steel: Hardness, SANS, Atom Probe, and Thermoelectric Power Investigations
Authors: Pareige P , Auger P , Welzel S , Van Duysen J-C , Miloudi S
Published: April 2008

The evolution of the microstructure of the French Chooz A pressure vessel steel (characterized by a low copper content (<0.1 at.% Cu)) has been studied during annealing at 450°C. The irradiated material (fluence: 0.5 to 16×1023 n.m-2, flux : 2×1015 n.m-2.s-1, temperature : 275°C) shows a quick recovery of hardness after 20 h of aging. The intensity of the SANS scattered signal as well, as the thermoelectric power one, decreases with annealing time. In addition, 3D images, at the atomic scale, given by the Tomographic Atom Probe show that during the first hours of annealing, the neutron-induced Cu-Ni-Mn-Si clusters formed during the in-service irradiation are “dissolved” while copper precipitates are formed. Their low number density is in agreement with the full recovery of the hardness.

These results are of primary importance for prediction of recovery of an irradiation embrittled steel and also for prediction of its behavior during reirradiation. This work also shows that the nondestructive technique, thermoelectric power, may be used to “follow” the microstructural evolution of materials during annealing treatments.
Application of the Floating Curve Model for Estimation of Re-Irradiation Embrittlement of VVER-440 RPV Steels
Authors: Nikolaev YA , Nikolaeva AV
Published: April 2008

Radiation embrittlement and its mitigation by annealing of VVER-440 reactor pressure vessel steels are studied. The Russian regulatory approach for prediction of radiation embrittlement of VVER-440 steels is considered. Results of an investigation of materials cut out of operating nuclear power plants are discussed. Different models of re-irradiation embrittlement are compared. A new model for assessment of embrittlement under re-irradiation is developed.
Embrittlement, Hardening, and Grain Boundary Composition in an Fe-P-C Alloy after Irradiation or Thermal Aging
Authors: Jones RB , Cowan JR , Corcoran RC , Walmsley JC
Published: April 2008

An Fe-1200ppmP-30ppmC alloy has been examined after aging and irradiation exposure at 240°C. Thermal aging for 97 days reduced the yield stress by 10% but increased the 40J Charpy transition temperature by 50°C. Irradiation for a similar duration, in a highly thermalised neutron spectrum at 5.8 ± 1.7×10-11dpa/sec, increased both the yield stress by 15% and the transition temperature by 70°C. Longer irradiation, in spectra with or without the high thermal fraction, produced additional hardening but little extra embrittlement. Irradiation and aging gave a fivefold increase in the fraction of brittle intergranular fracture. Auger Spectroscopy showed that the P grain boundary monolayer coverage was virtually unchanged for all start-of-life, aging and irradiation conditions. In contrast, the C coverage reduced by 50% during the initial aging or irradiation, but changed little after longer irradiations. Grain boundary N increased initially but then stabilised. No grain boundary S occurred. Transmission microscopy revealed some dislocation loops and small regions of strain contrast in the grains after irradiation but no P-rich precipitates were seen. From this, and other published work, it is concluded that irradiation hardening arises from dislocation loops and possibly from P-rich cluster formation. Embrittlement developed due to loss of C from the grain boundaries, rather than from changes to the boundary P content. Loss of C reduced the boundary fracture stress but embrittlement was independent of the matrix strength. For the irradiation conditions examined, there was no evidence of irradiation-induced grain boundary P segregation.
Neutron Irradiation Rate Dependence of Damage Structures in Fe-Cu Model Alloys
Authors: Yanagita S , Xu Q , Yoshiie T , Ino H
Published: April 2008

Fe-Cu model alloys and pure iron were irradiated at 573 K under two different neutron flux conditions (9.4 × 10-8 dpa/s and 1.1 × 10-9 dpa/s) up to about 0.008 dpa. Although defect clusters were not detected by TEM in any specimen, the results of positron lifetime measurement and mechanical property tests clearly indicated an irradiation rate dependence on the microvoid formation and mechanical properties. In positron lifetime measurements, the size of microvoids under low flux irradiation was larger than under high flux irradiation. But the number density was lower than high flux irradiation. In mechanical property testing, more hardening occurred by low flux irradiation. The irradiation rate dependence increased with copper content. These results suggest that the dependence is caused by the precipitation of copper. As copper migrates mainly by the vacancy mechanism, the size and the density of precipitates increase with increasing vacancy migration distance, which becomes longer with a decreasing of the irradiation rate.
Creep-Fatigue Response of 20% CW 316 SS Under Irradiation at 60°C
Authors: Murase Y , Nagakawa J , Yamamoto N , Fukuzawa Y
Published: April 2008

An in-beam creep-fatigue testing machine for a miniaturized specimen was developed at National Research Institute for Metals to investigate the synergistic effects of stressing and irradiation damage on the mechanical properties of materials. The loading and displacement measuring system of the machine is outlined in this paper. An in-beam creep-fatigue experiment was carried out with 17 MeV protons for 20% cold-worked 316 stainless steel at 60°C. The specimen size was 4× 10×0.15 mm in gauge with fatigue crack starter notch introduced at the edge. Creep-fatigue (ramp with hold) tests were carried out in stress control (536.4 MPa, 12 s ↔ 236.4 MPa, 1 s; rate = 50 MPa/s). Simple fatigue tests without holding time were also performed in the same condition. Although the lifetime of simple fatigue was substantially prolonged under irradiation, the prolongation was not so much in the creep-fatigue. The holding during irradiation appeared to play some role in reducing the lifetime.
The Question of Saturation of Void Swelling in Fe-Cr-Ni Austenitic Alloys
Authors: Garner FA , Black CA
Published: April 2008

Whereas in earlier publications it appeared that a possible saturation of neutron-induced void swelling might be developing at temperatures of 400 and 427°C in annealed Fe-Cr-Ni model alloys in a manner that was dependent on composition, data obtained at higher neutron exposures in EBR-II have led to the conclusion that such an expectation was erroneous. At all temperatures in the range 400–650°C, the neutron induced swelling eventually reaches ∼1%/dpa and continues to very high swelling levels thereafter without any hint of saturation.
The Effect of Molybdenum on the Grain Boundary Segregation of Chromium Exhibited by Austenitic Stainless Steels after Neutron Irradiation in a BWR
Authors: Goodwin CC , Faulkner RG , Walmsley JC , Spellward P
Published: April 2008

Analytical investigations employing field-emission gun scanning transmission electron microscopy (FEGSTEM) have indicated that the degree of chromium depletion at grain boundaries is less pronounced in AISI type-316 when compared to type-304 stainless steel after neutron irradiation in Boiling Water Reactor (BWR) conditions to a fluence of 1026 n.m-2 (E > 1 MeV). An explanation of this observation is proposed, which is based on the non-equilibrium segregation (NES) theory of grain boundary segregation. NES model predictions for the radiation-induced segregation of chromium and molybdenum in an austenitic matrix are presented together for the first time. The predicted segregation profiles are compared with the FEGSTEM data, whilst the predicted trends as a function of neutron dose are discussed in terms of possible microstructural evolution in a BWR environment. In particular it will be demonstrated how, according to NES theory, the binding of molybdenum to self-interstitial atoms (SIAs) will take precedence over chromium. In turn, the subsequent migration of molybdenum-SIA “complexes” retard the migration of chromium away from the grain boundary, thus moderating the magnitude of radiation-induced chromium depletion.
Void-Induced Swelling and Embrittlement in the Russian EI-847 Stainless Steel at PWR-Relevant End-of-Life Conditions
Authors: Garner FA , Porollo SI , Vorobjev AN , Konobeev YV , Dvoriashin AM , Krigan VM , Budylkin NI , Mironova EG
Published: April 2008

Void swelling has been proposed to eventually reach significant levels in some austenitic internals at PWR end-of-life conditions. A prevailing perception based on previously published data from fast reactors is that swelling will never reach significant levels at temperatures in the 330–372°C range, even at the 80–90 dpa level anticipated in PWR baffle-former components after 40 years of operation. Fast reactor data are presented in this paper that show that this perception is incorrect, at least for a niobium-stabilized 16Cr-15Ni steel used in Russian reactors for applications where 316 would be used in Western reactors. As a result of the large swelling levels reached in some cases, the steel becomes very brittle, as expected from fast reactor experience.
Interaction of Void-Induced Phase Instability and Subsequent Void Growth in AISI 304 Stainless Steel
Authors: Porter DL , Garner FA , Bond GM
Published: April 2008

There are two types of radiation-induced phase instabilities that AISI 304 is prone to develop at temperatures in the 400–550°C range at dpa rates typical of fast reactors, and there is some concern that these instabilities may shift to lower temperatures (≥400°C) at lower dpa rates typical of pressurized water reactors.

One of these instabilities, the void-induced γ→α transformation, is investigated in this study at dpa rates typical of both types of reactors. The mode of nickel segregation shifts somewhat at PWR-relevant conditions, but the γ→α transformation has not yet been observed in the range 380–400°C for swelling levels less than 2%.
Modeling of Helium Bubble Formation During Welding of Irradiated Stainless Steels
Authors: Hashimoto T , Mochizuki M
Published: April 2008

Helium atoms are generated in neutron-irradiated stainless steels. In order to predict the post-irradiation weldability, a model is presented to calculate helium bubble formation along grain boundaries during welding of helium containing stainless steels. It calculates helium diffusion toward the grain boundary, bubble nucleation, coalescence and growth at the grain boundary, using the temperature and stress changes during welding. Annealing experiments of helium implanted stainless steels at constant temperatures are analyzed first to examine the bubble formation without tensile stresses. Good agreement with measurements is obtained for the bubble density and the average radius. Then, bubble formations are calculated for a bead-on-tube welding test on a neutron-irradiated stainless steel. Good correlation is confirmed between bubble formation calculations and the crack formation observed in the welding test.
Microstructural Examination of Irradiated Beryllium Pebbles
Authors: Gelles DS , Dalle Donne M , Kawamura H , Ohnuki S , Scaffidi-Argentina F
Published: April 2008

Commercially available beryllium pebbles have been irradiated to approximately 4×1022 n/cm2 (E > 0.1 MeV) at 380°C in the EBR-II. Microstructural examination has now been performed on pebbles 3 mm in diameter manufactured by Brush Wellman, USA and contained in that experiment. This paper describes optical, scanning electron and transmission electron microscopy of these pebbles. It was found that the major effect of irradiation was formation of small, highly non-equiaxed bubbles. The bubbles were thin disk shapes lying on the basal plane.
In-situ Electrical Resistivity Measurement of MgO•Al2O3 Coating under Neutron Irradiation
Authors: Nakamichi M , Kawamura H
Published: April 2008

Ceramic coatings on the surface of structural materials such as 316SS have been considered for electrical insulators and tritium permeation barriers in fusion reactor designs. MgO•Al2O3 is one of the most promising materials as a coating material from a point of high electrical resistivity and so on. In this report, the electrical resistivity of a MgO•Al2O3 coating on 316SS was investigated under neutron irradiation with the JMTR (Japan Materials Testing Reactor) by in-situ electrical resistivity measurement. The specimen was irradiated at 360°C in a He atmosphere. Radiation Induced Conductivity (RIC) was observed at the beginning of reactor. The electrical resistivity of the coating before neutron irradiation decreased from 1×1011Ω•cm to 2×108Ω•cm with increase reactor power. The electrical resistivity of the coating under neutron irradiation decreased again from 2×108Ω•cm to 5 × 104Ω•cm above 1 × 1019n/cm2 (En> 1 MeV). It did not recover up to unirradiated level. The electrical resistivity of the coating after neutron irradiation was about 2×106Ω•cm. It was clear that Radiation Induced Electrical Degradation (RIED) in the MgO•Al2O3 coating was recognized above a fluence of about 1×1020n/cm2 (En> 1MeV).
Effect of Ionizing and Displacive Components of Irradiation on the Structural Transformations in Pyrolitic Boron Nitride Ceramics
Authors: Stepanov VA , Stepanov PA , Chernov VM , Golovinov AV , Kryukova LM
Published: April 2008

The structure of pyrolitic boron nitride (BN) samples assembled to a package and irradiated with protons has been studied with Raman spectroscopy and X-ray diffraction. Shift and change of shape of X-ray bands and change of Raman band shape have been found to correlate with the calculated values of energy losses along the proton track. Ionizing constituent of irradiation is responsible for a change of the state of grain boundaries and partial recrystallization of pyrolitic BN. Displacive constituent of irradiation is responsible for stacking fault formation in the grains. A mechanism of the effect of ionizing constituent of irradiation has been introduced, where energy release on grain boundaries and break of chemical bonds on the boundaries result in partial recrystallization of the material. According to this mechanism the recrystallization is much more feasible in pre-irradiated materials.
Helium/Hydrogen Measurements in High-Energy Proton-Irradiated Tungsten
Authors: Oliver BM , Hamilton ML , Garner FA , Sommer WF , Maloy SA , Ferguson PD
Published: April 2008

To provide structural material design data for the Accelerator Production of Tritium (APT) project, a 1 mA, 800 MeV proton beam at the Los Alamos Neutron Science Center (LANSCE) was used to irradiate a large number of metal samples, including a tungsten target similar to that being considered as the neutron source for the tritium production. The maximum proton fluence to the tungsten target was ∼1021 protons/cm2. An unavoidable byproduct of spallation reactions is the formation of large amounts of hydrogen and helium. Postulated off normal conditions for APT would result in target temperatures approaching 1200°C. The use of tungsten target rods clad with Alloy 718 raises concerns as to the amount and rate of release of these gases under such off normal conditions, with the major concern being pressurization and possible failure of the cladding. To address these issues, portions of the LANSCE tungsten rods were heated to temperatures up to 1200°C, and the time-dependent evolution of helium and hydrogen gas was measured.

Stepped-anneal and total helium/hydrogen measurements were conducted at Pacific Northwest National Laboratory (PNNL) on multiple samples of the tungsten material. Helium measurements were conducted using a high-sensitivity magnetic-sector isotope-dilution helium analysis system. Stepped-anneal measurements were conducted at temperatures from 25°C to 1600°C in ∼100°C steps. Total helium measurements were conducted by rapid vaporization after completion of the stepped-anneal process, and are compared with Monte Carlo calculations performed at Los Alamos National Laboratory (LANL) using the LAHET code system.

Hydrogen measurements were conducted between 750°C and 1200°C using a high-temperature furnace that had been extensively modified for the application. Hydrogen detection was accomplished by periodic sampling of the furnace gas using a separate quadrupole analyzer. Hydrogen measurements are also compared with LANL calculations.
Effects of Fast Neutron Irradiation at 80K on Titanium Nickelide
Authors: Dubinin SF , Parkhomenko VD , Teploukhov SG
Published: April 2008

The structural state of a Ti49Ni51 single crystal irradiated by fast neutrons at 80K was investigated by the neutron diffraction method. It was found that a relatively low fluence amorphized 40% of sample volume. It was shown that the existence of a critical concentration of point defects provides the stability of positional atom disorder in Ti49Ni51 low temperature irradiation damages of titanium nickelide can be well explained within the Gibbon statistical scheme.
Irradiation Aging of Aluminum Alloys
Authors: Gillemot F , Gillemot L , Uri G
Published: April 2008

In Hungary the VVRSZM-5 research reactor went into operation in 1959 and was operated continuously until 1986 without any failure. In 1986 the reactor was shut down for reconstruction. Since then the primary and secondary cooling circuits were dismantled and replaced. Many of these components were made from precipitation hardened aluminum alloy SZAV-1 similar in composition to AA6005 alloy. To increase knowledge about the aging of the mechanically loaded components few dismantled parts of the reactor were tested or samples were cut and stored for testing.

In the meantime a new aluminum alloy for production of the main components of the reconstructed reactor has been developed and tested. This alloy, R-AlMg2.5 was produced on a clean aluminum base, and its weldment was carefully tested before usage. Sample heats of the SZAV-1 alloy (material of the dismantled reactor) and its weldments were irradiated together with the new R-AlMg2.5 (similar to AA5252) alloy and the properties of the materials compared. The results of the laboratory test series show a good correlation with the data obtained on the out-of-service material and weldments.

The results shows that high dose irradiation (∼1024-1026 n/m2) fast neutron fluence, (E>1MeV) does not decrease the mechanical properties of the R-AlMg2.5 alloy and its weldments, while the toughness of the SZAV-1 weldments is reduced.

The reconstructed reactor (called VVRSZM-10) had also a surveillance testing program, and surveillance results are verify the application of the irradiation resistant R-AlMg2.5 alloy.
Neutron Irradiation Test of Optical Components for Fusion Reactor
Authors: Ishitsuka E , Sagawa H , Nagashima A , Sugie T , Nishitani T , Yamamoto S , Kawamura H
Published: April 2008

The neutron irradiation effects on the window and mirror materials proposed as optical elements for plasma diagnostic components were studied. Specimens were sapphire and KU-quartz as the window material, molybdenum corner cubic reflector, gold-coated mirror and aluminum corner cubic reflector. These materials were irradiated by the Japan Materials Testing Reactor, and transmission and refraction spectrum, change of color and surface roughness were observed.
Correlation of Hardening and Microstructure of Tantalum Irradiated with Heavy Ions
Authors: Muroga T , Yasunaga K , Katoh Y , Watanabe H , Yoshida N , Noda N
Published: April 2008

The microhardness change of tantalum irradiated with heavy ions has been evaluated using a micro-indentation technique. The increase in hardness by irradiation was significant below ∼900K and negligible above ∼1100K. The hardness decreased steeply with temperature at ∼1000K. Correlation of the results with microstructure data showed that the strength of voids and line dislocations as barriers against deformation should be very weak relative to that of loops and TEM-invisible defects. The temperature and fluence dependence of the hardening by TEM-invisible defects, derived by the correlation, implied an accumulation process of cascade vacancy clusters during irradiation.
Improvement of Post Irradiation Ductility of V-Ti-Cr-Si Type Alloys Neutron Irradiated Around 400°C
Authors: Satou M , Chuto T , Hasegawa A , Abe K
Published: April 2008

Mechanical properties after neutron irradiation at temperatures around 400°C of the V-Ti-Cr-Si type alloys were reported. Neutron irradiation was carried out at temperatures 374 °C and 406 °C with fluence of about 7 and 50 dpa, respectively. Tensile tests were conducted using miniaturized specimens at ambient temperature and irradiation temperature. Uniform elongation of the alloy V-5Ti-5Cr-1Si-1Al-1Y (nominal weight percentage) was maintained at about 4.2 and 5.9% after irradiation and testing at 374 and 406 °C, respectively. Increases in yield stress were 240 MPa and 160 MPa, respectively. Although the irradiation hardening was large, apparently smaller than that of V-Ti-Cr type alloys without Si, Al and Y additions, it was possible that the additional elements reduced the concentration of interstitial impurities in solution, which caused high irradiation hardening. It was found that a small addition of Si, Al and Y showed possible improvement in the ductility at lower temperature irradiation of vanadium alloys.
The Influence of Transmutation and Initial Composition on the Microstructural Evolution of Vanadium and V-Cr-Ti Alloys Irradiated in HFIR
Authors: Ohnuki S , Takahashi H , Garner FA
Published: April 2008

Pure vanadium and a series of V-Cr and V-5Ti-Cr alloys were irradiated at 300, 400, 500 and 600°C to doses of 8–10 dpa in the HFIR mixed spectra reactor. Transmutation of vanadium increased the chromium content by 6–7 additional percent. This compositional change had the largest impact on the microstructural and microchemical evolution of the alloys and also dominated the change observed in alloy density. Even more importantly, the results were quite different than would be expected from the results of similar irradiation in FFTF where transmutation is much lower.
Investigations about Oxygen Fires in Electronic Circuits of Medical Devices
Authors: Kohl H-J , Morbach A , Degenhart R
Published: April 2008

For the purpose of designing safe medical devices in the presence of oxygen we investigated two aspects of circuits. First, we examined the ignition of adjacent material by sparks, generated by opening contacts of solder tin in pure oxygen at ambient pressure under different electrical loading conditions. Second, the burning of electronic circuits in an oxygen chamber at 300 or 500 kPa absolute pressure and 21% or 25% oxygen concentration after ignition with a tantal capacitor was studied. Safe values of power for resistors and energy for capacitors and inductances are concluded for circuits with a supply voltage less than 30V. Measures to avoid burning in an oxygen chamber were defined.
Ignitability in Air, Gaseous Oxygen, and Oxygen-Enriched Environments of Polymers Used in Breathing-Air Devices, Final Report
Authors: Bryan CJ , Hirsch DB , Haas J , Beeson HD
Published: April 2008

Commercial breathing-air devices contain gases that may vary in concentration from 20.9 percent to 100 percent oxygen at pressures up to 34.5 megapascal (MPa). Analysis of the limited flammability and reactivity data available on some of the materials in these devices under use conditions raised serious concerns about the risks to National Aeronautics and Space Administration (NASA) and contractor personnel at the John F. Kennedy Space Center (KSC). Therefore, these materials were evaluated using the criteria of the ASTM Test Method for Autogenous Ignition Temperature of Liquids and Solids in a High-Pressure Oxygen-Enriched Environment (G 72), ASTM Test Method for Determining Ignition Sensitivity of Materials to Mechanical Impact in Ambient Liquid Oxygen and Pressurized Liquid and Gaseous Oxygen Environments (G 86), ASTM Test Method for Heat of Combustion of Liquid Hydrocarbon Fuels by Bomb Calorimeter (D 240), and Flammability, Odor, Offgassing, and Compatibility Requirements and Test Procedures for Materials in Environments that Support Combustion (NASA-STD-6001), Test 1, Upward Flame Propagation. Materials typically used in these devices that do not meet the KSC requirements include: polyethylene, Delrin®, Neoprene®, Buna N®, Zytel 42®, ethylene/propylene rubber (EPDM), Tefzel®, Noryl®, Buna S®, and silicone rubber. Comparative data were also generated on Teflon TFE®, Viton A®, Vespel SP-21®, and Kel-F 81®. In mechanical impact tests, polyethylene, Buna N®, Neoprene®, and Zytel 42® passed at higher than expected energy levels while Tefzel® reacted similar to Delrin®. In the autoignition tests (AIT), the AIT for Buna N®, Neoprene®, ethylene/propylene rubber (EPDM), Buna S®, Delrin®, Tefzel®, Viton A®, and Zytel 42® decreased with increasing oxygen concentrations while the effect of oxygen concentration was minimal for Teflon TFE®, Kel-F 81®, Vespel SP-21®, polyethylene, and silicone. The heats of combustion ranged from 6.17 megajoule per kilogram (MJ/kg) for Kel-F 81® to 54.59 MJ/kg for polyethylene. All materials failed the upward flammability test in 20.9 percent oxygen except Teflon TFE®, silicone, Zytel 42®, Viton A®, and Kel-F 81®.
The Presence of Excess Oxygen in Burning Metallic Materials
Authors: Wilson DB , Steinberg TA , DeWit JR
Published: April 2008

Early work on burning of iron rods under conditions of the ASTM/NASA flammability test showed that there was excess oxygen, that is, above stoichiometric requirements for Fe2O3, present in the molten product formed during burning. Since that initial work, this phenomenon has been confirmed for burning under microgravity conditions and has been observed for a wide range of metals under various burning conditions at ambient pressures and under pressurized oxygen-enriched conditions. This paper reviews these experimental observations and discusses the possible thermodynamic analysis for the metals iron, cobalt, and nickel. The excess oxygen in the burning molten iron oxide is represented as combined to form a series of ferrite ions. For cobalt the excess oxygen is represented as a bridging species; that is, the cobalt oxide liquid is oxygenated. There is no definitive evidence for excess oxygen in nickel oxide.
ASTM G 124 Test Data For Selected Al-Si Alloys, Al-Composites, Binary Alloys and Stainless Steels
Authors: De Wit JR , Steinberg TA , Haas JP
Published: April 2008

Promoted-ignition/combustion testing of metallic materials in gaseous oxygen at pressures of up to 68.9 MPa (10,000 psia) has been performed at NASA's White Sands Test Facility (WSTF) since 1986. Recent work within the Department of Mechanical Engineering at The University of Queensland (UQ), Australia, prompted an extensive review of WSTF's promoted-ignition/combustion database. Previously unpublished data were found relating specifically to the present research, as well as other data which may be of interest to industry and researchers. This paper presents data on the following specific metallic systems; Aluminum-Silicon Binary Alloys (10, 20, 30, and 50 wt% Si), A luminum-Silicon Carbide Metal Matrix Composites (MMCs) (10, 20, 30, and 55 volume% SiC) and other related materials, Binary Alloy Combinations of Al, Co, Cu, Fe, Ni, and Si, and Stainless Steels 304, 316, 347, CF8M, and 15-5 PH. A brief discussion of the data and the motivation behind the testing is included. The results indicate that the Al-Si alloys, up to 50 wt% Si, have threshold pressures similar to that of pure aluminium. Results for the Al-SiC MMCs show a significant increase in threshold pressure over that of pure aluminium. This material may provide a useable alternative to conventional oxygen system component materials. Further testing of binary alloys and stainless steels to better understand the influence of different configurations and alloying elements is warranted.