Alloying effect on martensite transformation in stainless steels

International Nuclear Information System (INIS)

Gulyaev, A.P.; Shlyamnev, A.P.; Sorokina, N.A.

1975-01-01

The effect of cobalt, nickel, molybdenum on the martensite transformation kinetics in stainless steels containing 9 to 13% Cr has been studied. Cobalt in Fe-Cr base alloys decreases the temperature of the Msub(in) and Msub(fin) points without a considerable decrease of the martensite phase amount after the transformation. Nickel reduces the martensite transformation temperature range, the nickel effect being enhanced in the presence of cobalt, which is characterized by a change of the linear dependence Msub(in)=f(%Ni) for a quadratic one. Molybdenum decreases the temperature of the Msub(in) and Msub(fin) points intensively, thus, substantially increasing the residual austenite amount. In the steels investigated Ni and Co decrease, whereas Mo increases, to some extent, the temperature of the reverse a-γ-transformation. The reduction of chromium content from 13 to 9% stimulates the martensite transformation initiation, that is why, in alloys containing 9% Cr, the increase in the contents of Ni, Co., Mo with the martensite structure maintained is possible. A further alloying of steel containing 13% Cr with these elements is rather limited due to the inhibition of the martensite transformation

Microstructural evolution of martensitic steels during fast neutron iradiation

International Nuclear Information System (INIS)

Maziasz, P.J.

1989-01-01

Irradiation of martensitic/ferritic steels with fast neutrons (E > 0.1 MeV) to displacement damage levels of 30--50 dpa at temperatures of 300--500 degree C produces significant changes in the as-tempered microstructure. Dislocation loops and networks can be produced, irradiation-induced precipitates can form, the lath/subgrain boundary structure and the thermal precipitates produced during tempering can become unstable, and if helium is present, bubbles and voids can form. These microstructural changes caused by irradiation can have important effects on the properties of this class of steels for both fast breeder reactor (FBR) and magnetic fusion reactor (MFR) applications. The purpose of this paper is to compare reactor-irradiated and long-term thermally aged 9Cr--1MoVNb specimens, in order to distinguish effects due to displacement damage from those caused by elevated-temperature exposure alone. 7 refs., 1 fig

Effects of irradiation on tungsten stabilized martensitic steels*1

Science.gov (United States)

Gelles, D. S.; Hsu, C. Y.; Lechtenberg, T. A.

1988-07-01

Tungsten stabilized martensitic stainless steels are being developed for fusion reactor first wall applications in order to lower retained radioactivity so as to permit shallow land burial after reactor decommissioning. Two such alloys have been designed, fabricated, fast neutron irradiated in FFTF and examined by transmission electron microscopy. The two compositions were Fe-7.5Cr-2.0W-0.17 C and Fe-10.2Cr-1.7W-0.3V-0.02C. Conditions examined included irradiation temperatures of 365, 426, 520 and 600°C to doses as high as 34 dpa. Small amounts of void swelling are found at the two lowest temperatures. It is demonstrated that levels of tungsten on the order of 2 wt% do not result in excessive intermetallic precipitation under these irradiation conditions.

Microchemistry of neutron irradiated 12%CrMoVNb martensitic steel

International Nuclear Information System (INIS)

Little, E.A.; Morgan, T.S.; Faulkner, R.G.

1992-01-01

Non-equilibrium solute segregation has been studied in a 12%CrMoVNb martensitic steel following fast reactor irradiation at 465 C and correlated with the development of M 6 X η-phase. Cr, Ni, Si, Mo, P and Mn are all shown to exhibit positive segregation to lath boundaries and are subsequently incorporated into M 6 X precipitates. The co-segregation of a combination of these elements which include P and Si, and possibly Cr or Mo, appears to promote M 6 X formation

Martensitic/ferritic steels as container materials for liquid mercury target of ESS

International Nuclear Information System (INIS)

Dai, Y.

1996-01-01

In the previous report, the suitability of steels as the ESS liquid mercury target container material was discussed on the basis of the existing database on conventional austenitic and martensitic/ferritic steels, especially on their representatives, solution annealed 316 stainless steel (SA 316) and Sandvik HT-9 martensitic steel (HT-9). Compared to solution annealed austenitic stainless steels, martensitic/ferritic steels have superior properties in terms of strength, thermal conductivity, thermal expansion, mercury corrosion resistance, void swelling and irradiation creep resistance. The main limitation for conventional martensitic/ferritic steels (CMFS) is embrittlement after low temperature (≤380 degrees C) irradiation. The ductile-brittle transition temperature (DBTT) can increase as much as 250 to 300 degrees C and the upper-shelf energy (USE), at the same time, reduce more than 50%. This makes the application temperature range of CMFS is likely between 300 degrees C to 500 degrees C. For the present target design concept, the temperature at the container will be likely controlled in a temperature range between 180 degrees C to 330 degrees C. Hence, CMFS seem to be difficult to apply. However, solution annealed austenitic stainless steels are also difficult to apply as the maximum stress level at the container will be higher than the design stress. The solution to the problem is very likely to use advanced low-activation martensitic/ferritic steels (LAMS) developed by the fusion materials community though the present database on the materials is still very limited

Neutron irradiation effects on magnetic properties of iron-nickel Invar alloys

International Nuclear Information System (INIS)

Morita, H.; Tanji, Y.; Hiroyoshi, H.; Nakagawa, Y.

1983-01-01

The Curie temperature of fcc Fe-Ni containing 30-50% Ni is reaised by neutron irradiation, although no appreciable change is detected in the X-ray diffraction pattern. These results are related to a tendency to two-phase separation of the fcc phase. (orig.)

Temperature Effects on the Mechanical Properties of Candidate SNS Target Container Materials after Proton and Neutron Irradiation; TOPICAL

International Nuclear Information System (INIS)

Byun, T.S.

2001-01-01

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54 to 2.53 dpa. Irradiation temperatures were in the range 30 to 100 C. Tensile testing was performed at room temperature (20 C) and 164 C to study the effects of test temperature on the tensile properties. Test materials displayed significant radiation-induced hardening and loss of ductility due to irradiation. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative strain hardening. In the EC316LN stainless steel, increasing the test temperature from 20 C to 164 C decreased the strength by 13 to 18% and the ductility by 8 to 36%. The tensile data for the EC316LN stainless steel irradiated in spallation conditions were in line with the values in a database for 316 stainless steels for doses up to 1 dpa irradiated in fission reactors at temperatures below 200 C. However, extra strengthening induced by helium and hydrogen contents is evident in some specimens irradiated to above about 1 dpa. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. It was estimated that the 316 stainless steels would retain more than 1% true stains to necking at 164 C after irradiation to 5 dpa. A calculation using reduction of area (RA) measurements and stress-strain data predicted positive strain hardening during plastic instability

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 deg. C

International Nuclear Information System (INIS)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-01-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 deg. C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 deg. C and 300 deg. C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 deg. C (up to 2.6 dpa), and tested between -170 deg. C and 300 deg. C. Irradiation effects at lower irradiation temperatures are more significant

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

Science.gov (United States)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-06-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between -170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.

Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

Science.gov (United States)

Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

2012-12-01

Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

A reassessment of the effects of helium on Charpy impact properties of ferritic/martensitic steels

International Nuclear Information System (INIS)

Gelles, D.S.; Hamilton, M.L.; Hankin, G.L.

1998-01-01

To test the effect of helium on Charpy impact properties of ferritic/martensitic steels, two approaches are reviewed: quantification of results of tests performed on specimens irradiated in reactors with very different neutron spectra, and isotopic tailoring experiments. Data analysis can show that if the differences in reactor response are indeed due to helium effects, then irradiation in a fusion machine at 400 C to 100 dpa and 1000 appm He will result in a ductile to brittle transition temperature shift of over 500 C. However, the response as a function of dose and helium level is unlikely to be simply due to helium based on physical reasoning. Shear punch tests and microstructural examinations also support this conclusion based on irradiated samples of a series of alloys made by adding various isotopes of nickel in order to vary the production of helium during irradiation in HFIR. The addition of nickel at any isotopic balance to the Fe-12Cr base alloy significantly increased the shear yield and maximum strengths of the alloys. However, helium itself, up to 75 appm at over 7 dpa appears to have little effect on the mechanical properties of the alloys. This behavior is instead understood to result from complex precipitation response. The database for effects of helium on embrittlement based on nickel additions is therefore probably misleading and experiments should be redesigned to avoid nickel precipitation

Extended solubility and martensitic hcp nickel formation in antimony implanted nickel

International Nuclear Information System (INIS)

Johnson, E.; Sarholt-Kristensen, L.; Johansen, A.

1982-01-01

Radiation damage microstructure and associated disorder have been investigated in antimony implanted nickel crystals using combined RBS and TEM analyses. In crystals implanted at and below room temperature with 80 keV Sb + ions to a fluence of 5x10 20 m -2 , the retained antimony concentration in the implantation zone is approaching 15 at.%, with nearly all the antimony located substitutionally. The associated disorder as seen in the RBS analysis is insignificant. Annealing up to 600 0 C has little influence on the antimony distribution, whilst the dechanneling level is reduced. TEM and diffraction analysis of room temperature implanted samples show that the radiation damage consists of dense distributions of dislocation clusters and tangles, superimposed on a rather homogeneous background of new phase particles, identified as hcp nickel. The particles have a size 0.1-0.2 μm. The high substitutional antimony concentration at and below room temperature, which exceeds the solubility limit, indicates that its formation is thermally diffusionless and rather an effect of radiation enhanced solubility. The diffusionless nature of the microstructure is also indicated from the presence of martensitic hcp nickel, believed to form due to relief of radiation induced internal stress. (Auth.)

Effect of Ni +-ION bombardment on nickel and binary nickel alloys

Science.gov (United States)

Roarty, K. B.; Sprague, J. A.; Johnson, R. A.; Smidt, F. A.

1981-03-01

Pure nickel and four binary nickel alloys have been subjected to high energy Ni ion bombardment at 675, 625 and 525°C. After irradiation, each specimen was studied by transmission electron microscopy. The pure nickel control was found to swell appreciably (1 to 5%) and the Ni-Al and the Ni-Ti samples were found to swell at all temperatures, but to a lesser degree (0.01 to 0.35%). The Ni-Mo contained a significant density of voids only at 525° C, while swelling was suppressed at all temperatures in the Ni-Si alloy. The dislocation structure progressed from loops to tangles as temperature increased in all materials except the Ni-Ti, in which there was an absence of loops at all temperatures. Dislocation densities decreased as temperature increased in all samples. These results do not correlate well with the relative behavior of the same alloys observed after neutron irradiation at 455°C. The differences between these two sets of data appear to be caused by different mechanisms controlling void nucleation in ion and neutron irradiation of these alloys.

On the reflectivity of nickel neutron mirrors

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Habib, N. (Atomic Energy Establishment, Cairo (Egypt). Reactor and Neutron Physics Dept.); Kenawy, M.A.; Wahba, M.; Ashry, A.H. (Ain Shams Univ., Cairo (Egypt))

1991-02-01

Neutron reflectivities were determined for 300 nm thick films of natural nickel and nickel 58 coated on glass plates. The measurements were performed at glancing angles between 40' and 60'. The incident neutron beam from one of the ET-RR-1 reactor horizontal channels covered neutron wavelengths between 0.55 and 0.80 nm. It was found that nickel 58, because of the high value of its critical glancing angle, is more efficient as a neutron mirror than natural nickel. (orig.).

Precipitate evolution in low-nickel austenitic stainless steels during neutron irradiation at very low dose rates

International Nuclear Information System (INIS)

Isobe, Y.; Sagisaka, M.; Garner, F.; Okita, T.

2007-01-01

Full text of publication follows: Not all components of a fusion reactor will be subjected to high atomic displacement rates. Some components outside the plasma containment may experience relatively low displacement rates but data generated under long-term irradiation at low dpa rates is hard to obtain. In another study the neutron-induced microstructural evolution in response to long term irradiation at very low dose rates was studied for a Russian low-nickel austenitic stainless steel that is analogous to AISI 304. The irradiated samples were obtained from an out-of-core anti-crush support column for the BN-600 fast reactor with doses ranging from 1.5 to 22 dpa generated at 3x10 -9 to 4x10 -8 dpa/s. The irradiation temperatures were in a very narrow range of 370-375 deg. C. Microstructural observation showed that in addition to voids and dislocations, an unexpectedly high density of small carbide precipitates was formed that are not usually observed at higher dpa rates in this temperature range. These results required us to ask if such unexpected precipitation was anomalous or was a general feature of low-flux, long-term irradiation. It is shown in this paper that a similar behavior was observed in a western stainless steel, namely AISI 304 stainless steel, irradiated at similar temperatures and dpa rates in the EBR-II fast reactor, indicating that irradiation at low dpa rates for many years leads to a different precipitate microstructure and therefore different associated changes in matrix composition than are generated at higher dpa rates. One consequence of this precipitation is a reduced lattice parameter of the alloy matrix, leading to densification that increases in strength with increasing temperature and dose. A. non-destructive method to evaluate these precipitates is under development and is also discussed in this paper. (authors)

Influence of impurities on the evolution of vacancy-type defects in neutron-irradiated nickel

International Nuclear Information System (INIS)

Druzhkov, A.P.; Perminov, D.A.; Arbuzov, V.L.

2012-01-01

Highlights: ► We study, by means of PAS, the effects of purity on damage evolution in neutron-irradiated Ni at 330 K. ► Impurity carbon atoms in solution decrease the cascade efficiency during irradiation. ► C–V complexes are formed on the recovery stage III in impure Ni irradiated with 10 −4 dpa. ► The formation of V-loops and SFTs dominate on stage III with increasing dose level. ► The thermal stability of SFTs in impure Ni is similar to that in pure Ni. - Abstract: In order to investigate the effect of impurities on vacancy defect evolution in nickel, specimens with high (5N) and technical (3N) purity were neutron-irradiated at ∼330 K in the IVV-2M reactor (Russia) to fluencies in the range of 1 × 10 21 –1 × 10 23 n/m 2 (E > 0.1 MeV) corresponding to displacement dose levels in the range of about 0.0001–0.01 dpa and subsequently stepwise annealed to about 900 K. The specimens of Ni with different purities were characterized both in as-irradiated state as well as after post-irradiation annealing by positron annihilation spectroscopy. The formation of three-dimensional vacancy clusters (3D-VCs) in cascades was observed under neutron irradiation. The density and size of 3D-VCs depended not only on dose level, but also on purity. The population of 3D-VCs in the technical Ni is lower than that in the high-purity Ni. 3D-VCs collapse into secondary-type clusters (stacking fault tetrahedra (SFTs) and vacancy loops) during stepwise annealing at 350–450 K (stage III in Ni). The suppression of secondary cluster formation in 3N Ni is attributed to an effective vacancy interaction with impurity carbon atoms, which based on a relatively large vacancy–carbon atom binding energy (0.32–0.35 eV). The trapping of vacancies released at the collapse of 3D-VCs by the interstitial impurity atoms dominates at low irradiation dose level (10 −4 dpa). Thus, we found that carbon impurity atoms have strong effects both on the primary vacancy-type defect

Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiang; Miao, Yinbin; Li, Meimei; Kirk, Marquis A.; Maloy, Stuart A.; Stubbins, James F.

2017-07-01

In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 x 10(15) ions/cm(2) at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 x 1015 ions/cm(2), about 51% of the loops were a(0)/2 < 111 > type for the 673 K irradiation, and the dominant loop type was a(0)< 100 > for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100-150 MPa. (C) 2017 Elsevier B.V. All rights reserved.

Radiation induced microstructural evolution in ferritic/martensitic steels

International Nuclear Information System (INIS)

Kohno, Y.; Kohyama, A.; Asakura, K.; Gelles, D.S.

1993-01-01

R and D of ferritic/martensitic steels as structural materials for fusion reactor is one of the most important issues of fusion technology. The efforts to characterize microstructural evolution under irradiation in the conventional Fe-Cr-Mo steels as well as newly developed Fe-Cr-Mn or Fe-Cr-W low activation ferritic/ martensitic steels have been continued. This paper provides some of the recent results of heavy irradiation effects on the microstructural evolution of ferritic/martensitic steels neutron irradiated in the FFTF/MOTA (Fast Flux Test Facility/Materials Open Test Assembly). Materials examined are Fe-10Cr-2Mo dual phase steel (JFMS: Japanese Ferritic/Martensitic Steel), Fe-12Cr-XMn-1Mo manganese stabilized martensitic steels and Fe-8Cr-2W Tungsten stabilized low activation martensitic steel (F82H). JFMS showed excellent void swelling resistance similar to 12Cr martensitic steel such as HT-9, while the manganese stabilized steels and F82H showed less void swelling resistance with small amount of void swelling at 640-700 K (F82H: 0.14% at 678 K). As for irradiation response of precipitate behavior, significant formation of intermetallic χ phase was observed in the manganese stabilized steels along grain boundaries which is though to cause mechanical property degradation. On the other hand, precipitates identified were the same type as those in unirradiated condition in F82H with no recognition of irradiation induced precipitates, which suggested satisfactory mechanical properties of F82H after the irradiation. (author)

Microstructural evolution of neutron-irradiated Ni-Si and Ni-Al alloys

Science.gov (United States)

Takahashi, H.; Garner, F. A.

1992-10-01

Additions of silicon and aluminum suppress the neutron-induced swelling of pure nickel but to different degrees. Silicon is much more effective initially when compared to aluminum on a per atom basis but silicon exhibits a nonmonotonic influence on swelling with increasing concentration. Silicon tends to segregate toward grain boundaries while aluminum segregates away from these boundaries. Whereas the formation of the Ni 3Si phase is frequently observed in charged particle irradiation experiments conducted at much higher displacement rates, it did not occur during neutron irradiation in this study. Precipitation also did not occur in Ni-5Al during neutron irradiation, nor has it been reported to occur during ion irradiation.

Microstructural evolution of neutron-irradiated Ni-Si and Ni-Al alloys

International Nuclear Information System (INIS)

Takahashi, H.; Garner, F.A.

1992-01-01

Additions of silicon and aluminium suppress the neutron-induced swelling of pure nickel but to different degrees. Silicon is much more effective initially when compared to aluminium on a per atom basis but silicon exhibits a nonmonotonic influence on swelling with increasing concentration. Silicon tends to segregate toward grain boundaries while aluminium segregates away from these boundaries. Whereas the formation of the Ni 3 Si phase is frequently observed in charged particle irradiation experiments conducted at much higher displacement rates, it did not occur during neutron irradiation in this study. Precipitation also did not occur in Ni-5Al during neutron irradiation, nor has it been reported to occur during ion irradiation. (orig.)

Twinning and martensitic transformations in nickel-enriched 304 austenitic steel during tensile and indentation deformations

Energy Technology Data Exchange (ETDEWEB)

Gussev, M.N., E-mail: gussevmn@ornl.gov; Busby, J.T.; Byun, T.S.; Parish, C.M.

2013-12-20

Twinning and martensitic transformation have been investigated in nickel-enriched AISI 304 stainless steel subjected to tensile and indentation deformation. Using electron backscatter diffraction (EBSD), the morphology of α- and ε-martensite and the effect of grain orientation to load axis on phase and structure transformations were analyzed in detail. It was found that the twinning occurred less frequently under indentation than under tension; also, twinning was not observed in [001] and [101] grains. In tensile tests, the martensite particles preferably formed at the deformation twins, intersections between twins, or at the twin-grain boundary intersections. Conversely, martensite formation in the indentation tests was not closely associated with twinning; instead, the majority of martensite was concentrated in the dense colonies near grain boundaries. Martensitic transformation seemed to be obstructed in the [001] grains in both tensile and indentation test cases. Under a tensile stress of 800 MPa, both α- and ε-martensites were found in the microstructure, but at 1100 MPa only α-martensite presented in the specimen. Under indentation, α- and ε-martensite were observed in the material regardless of the stress level.

Superconductivity degradation in Gd-containing high temperature superconductors (HTSC) under thermal neutron irradiation

International Nuclear Information System (INIS)

Petrov, A.; Kudrenitskis, I.; Makletsov, A.; Arhipov, A.; Karklin, N.

1999-01-01

The physical properties of ordered crystals are extremely sensitive to the degree of order in the distribution of the various kinds of atoms over the corresponding sites in the crystal lattice. An increasingly popular means of creating disordered states is to use nuclear radiation. The type of radiation defects which appear and the nature and degree of the structural changes in ordered crystals depend on the kind of radiation and the fluence level, the irradiation temperature, the type of crystal structure, the composition and initial disorder of the material, the character of the interatomic forces, etc. There are many such scientific publications where the effects of fast neutron irradiation on high temperature superconductors (HTSC) have been studied in both polycrystalline and single crystalline superconductors. It is known also that the role of thermal neutrons in structural defects forming is negligible in comparison with fast neutrons because of their small (∼0.025 eV) energy. But it is evident enough that in superconductors containing isotopes with large thermal neutron cross sections the important results concerning the role of point defects could be obtained. Such point defects are creating due to soft displacements of isotopes having interacted with thermal neutrons. Such the possibility of creating point defects in solids including HTSC is investigating by several groups (Austria, USA, China, Latvia) and these investigations have found the support in the person of IAEA. In this review the authors consider the changes brought about by thermal-neutron irradiation (E∼0.025 eV) in the structure, superconducting and magnetic properties of gadolinium containing ordered HTSC with the structure 123, whose extreme electric and magnetic properties continue to attract both research and practical interest. All of the studies reviewed have been done on bulk polycrystalline samples RBa 2 Cu 3 O 7-δ (where R - natural mixture of Gd isotopes, 155 Gd, 157 Gd, 160

Swelling in several commercial alloys irradiated to very high neutron fluence

International Nuclear Information System (INIS)

Gelles, D.S.; Pintler, J.S.

1984-01-01

Swelling values have been obtained from a set of commercial alloys irradiated in EBR-II to a peak fluence of 2.5 x 10 23 n/cm 2 (E > 0.1 MeV) or approx. 125 dpa covering the range 400 to 650 0 C. The alloys can be ranked for swelling resistance from highest to lowest as follows: the martensitic and ferritic alloys, the niobium based alloys, the precipitation strengthened iron and nickel based alloys, the molybdenum alloys and the austenitic alloys

Study of points defects produced by irradiation of monocrystalline nickel and polycrystalline gadolinium; Etude des defauts ponctuels crees par irradiation dans: - le nickel monocristallin - le gadolinium polycristallin

Energy Technology Data Exchange (ETDEWEB)

Cope, R [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1969-07-01

The work described in this thesis falls into two parts: the first comprises a study of magnetocrystalline nickel by resistivity measurements; the second is a description of resistivity and magnetic after effect measurements on an h.c.p. ferromagnetic crystal other than cobalt, namely gadolinium. For the first part we have demonstrated the existence of a small but definite orientation dependence in the creation of point defects by electron irradiation (20 deg. K) of a nickel single crystal. In particular, the effect is manifested in the form of the stage I{sub C}, II and III in the resistivity recovery. In the second part an important result has emerged: namely that there is no magnetic after effect phenomenon in a neutron irradiated (27 deg. K) ferromagnetic metal. Several considerations are discussed by way of a preliminary interpretation of this important difference between gadolinium and cobalt. (author) [French] Le travail qui a fait l'objet de ce memoire comporte deux parties distinctes: une premiere constituee de l'etude du nickel sous la forme de monocristaux, a l'aide de mesures de resistivite; la seconde partie est composee d'experiences de resistivite et de trainage magnetique sur un cristal h.c.p. ferromagnetique autre que le cobalt, le gadolinium. Dans la premiere partie, nous avons mis en evidence un effet sensible de la direction de l'irradiation electronique (a 20 deg. K) sur la creation des defauts ponctuels dans le nickel monocristallin, en particulier au niveau des stades I{sub C}, II et III des courbes de revenu de la resistivite. Dans la deuxieme partie, un point particulier a ete mis en evidence: l'absence de phenomene de trainage magnetique dans un metal ferromagnetique irradie par des neutrons (a 27 deg. K). Quelques points d'interpretations preliminaires sont avances pour expliquer la difference notable entre le gadolinium et le cobalt. (auteur)

Hardness of AISI type 410 martensitic steels after high temperature irradiation via nanoindentation

Science.gov (United States)

Waseem, Owais Ahmed; Jeong, Jong-Ryul; Park, Byong-Guk; Maeng, Cheol-Soo; Lee, Myoung-Goo; Ryu, Ho Jin

2017-11-01

The hardness of irradiated AISI type 410 martensitic steel, which is utilized in structural and magnetic components of nuclear power plants, is investigated in this study. Proton irradiation of AISI type 410 martensitic steel samples was carried out by exposing the samples to 3 MeV protons up to a 1.0 × 1017 p/cm2 fluence level at a representative nuclear reactor coolant temperature of 350 °C. The assessment of deleterious effects of irradiation on the micro-structure and mechanical behavior of the AISI type 410 martensitic steel samples via transmission electron microscopy-energy dispersive spectroscopy and cross-sectional nano-indentation showed no significant variation in the microscopic or mechanical characteristics. These results ensure the integrity of the structural and magnetic components of nuclear reactors made of AISI type 410 martensitic steel under high-temperature irradiation damage levels up to approximately 5.2 × 10-3 dpa.

Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

Energy Technology Data Exchange (ETDEWEB)

Sato, K., E-mail: ksato@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Xu, Q.; Yoshiie, T. [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Dai, Y. [Spallation Neutron Source Division, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Kikuchi, K. [Frontier Research Center for Applied Atomic Sciences, Ibaraki University, Tokai-mura, Naka-gun, Ibaraki 319-1106 (Japan)

2012-12-15

Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<{approx}0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

International Nuclear Information System (INIS)

Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

2012-01-01

Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<∼0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

Effect of heat treatment and irradiation temperature on mechanical properties and structure of reduced-activation Cr-W-V steels of bainitic, martensitic, and martensitic-ferritic classes

International Nuclear Information System (INIS)

Gorynin, I.V.; Rybin, V.V.; Kursevich, I.P.; Lapin, A.N.; Nesterova, E.V.; Klepikov, E.Yu.

2000-01-01

Effects of molybdenum replacement by tungsten in steels of the bainitic, martensitic, and martensitic-ferritic classes containing 2.5%, 8% and 11% Cr, respectively, were investigated. The phase composition and structure of the bainitic steels were varied by changing the cooling rates from the austenitization temperature (from values typical for normalization up to V=3.3 x 10 -2 deg. C/s) and then tempering. The steels were irradiated to a fluence of 4x10 23 n/m 2 (≥0.5 MeV) at 270 deg. C and to fluences of 1.3x10 23 and 1.2x10 24 n/m 2 (≥0.5 MeV) at 70 deg. C. The 2.5Cr-1.4WV and 8Cr-1.5WV steels have shown lower values of the shifts in ductile-brittle transition temperature (DBTT) under irradiation in comparison with corresponding Cr-Mo steels. Radiation embrittlement at elevated irradiation temperature was lowest in bainitic 2.5Cr-1.4WV steel and martensitic-ferritic 11Cr-1.5WV steel. The positive effect of molybdenum replacement by tungsten at irradiation temperature ∼300 deg. C is reversed at T irr =70 deg. C

The martensitic transformation of Ti–Ni shape memory thin films under proton irradiation

International Nuclear Information System (INIS)

Gao, Z.Y.; Wang, H.Z.; Zhu, Y.Y.; Meng, X.L.; Cai, W.

2015-01-01

The martensitic transformation behavior of a Ti–Ni alloy irradiated by proton with different doses has been investigated. It is found that the samples irradiated by 150 keV protons have a two-step phase transformation during heating and only one-step transformation during cooling. The exothermic peak at higher temperature disappears during the following thermal cycling. A model based on the stress-assisted martensitic transformation was established by the Transport of Ions in Matter (TRIM) calculations in order to explain the transformation behavior. - Highlights: • Martensitic transformation behavior of TiNi alloy under proton irradiation • Two-step transformation appears upon heating for a sample irradiated at 150 keV. • One-step transformation appears upon cooling for a sample irradiated at 150 keV. • In the following thermal cycling, the higher temperature exothermic peak vanishes

Twinning processes in Cu-Al-Ni martensite single crystals investigated by neutron single crystal diffraction method

International Nuclear Information System (INIS)

Molnar, P.; Sittner, P.; Novak, V.; Lukas, P.

2008-01-01

A neutron single crystal diffraction method for inspecting the quality of martensite single crystals is introduced. True interface-free martensite single crystals are indispensable for, e.g. measurement of elastic constants of phases by ultrasonic techniques. The neutron diffraction method was used to detect and distinguish the presence of individual lattice correspondence variants of the 2H orthorhombic martensite phase in Cu-Al-Ni as well as to follow the activity of twinning processes during the deformation test on the martensite variant single crystals. When preparing the martensite single variant prism-shaped crystals by compression deformation method, typically a small fraction of second unwanted martensitic variant (compound twin) remains in the prism samples. Due to the very low stress (∼1 MPa) for the compound twinning in many shape memory alloys, it is quite difficult not only to deplete the martensite prisms of all internal interfaces but mainly to keep them in the martensite single variant state for a long time needed for further investigations

Characterization of the martensite phase formed during hydrogen ion irradiation in austenitic stainless steel

Science.gov (United States)

Jin, Hyung-Ha; Lim, Sangyeob; Kwon, Junhyun

2017-10-01

Microstructural changes in austenitic stainless steel caused by hydrogen ion irradiation were investigated using transmission electron microscopy (TEM). It has been confirmed that the irradiation induced the formation of martensite along the grain boundary; the martensite phase exhibited a crystal orientation relationship with the adjacent austenite phase. The results of this study also indicate that the concentration of Cr in the martensite phase is lower compared to that in the austenite matrix. The TEM results showed the development of asymmetric radiation-induced segregation (RIS) near the grain boundary, which leads to local changes in the chemical composition such as reduction of Cr near the grain boundary. The asymmetric RIS serves as a prerequisite for the formation of the martensite under hydrogen irradiation.

Swelling in neutron irradiated nickel-base alloys

International Nuclear Information System (INIS)

Brager, H.R.; Bell, W.L.

1972-01-01

Inconel 625, Incoloy 800 and Hastelloy X were neutron irradiated at 500 to 700 0 C. It was found that of the three alloys investigated, Inconel 625 offers the greatest swelling resistance. The superior swelling resistance of Inconel 625 relative to that of Hastelloy-X is probably related to differences in the concentrations of the minor rather than major alloy constituents, and can involve (a) enhanced recombination of defects in the Inconel 625 and (b) preferential attraction of vacancies to incoherent precipitates. (U.S.)

Container for irradiated fuel

International Nuclear Information System (INIS)

Guy, R.

1978-01-01

The transport container for irradiated or used nuclear fuel is provided with an identical heat shield against fires on the top and bottom sides. Each heat shield consists of two inner nickel plates, whose contact surfaces are polished to a mirror finish and an outer plate of stainless steel. The nickel plate on the box is spot welded to it while the second nickel plate is spot welded to the steel plate. Both together are in turn welded so as to be leaktight to the edges of the box. For extreme heat effects and based on the different (bimetal) coefficients of expansion, the steel plate with the nickel plate attached to it bulges away from the box. The second nickel plate remains at the box, so that a subpressure space is formed with the mirror nickel surfaces. The heat radiation and heat conduction to the box are greatly reduced by this. (DG) [de

Neutron irradiation of seeds 2

Energy Technology Data Exchange (ETDEWEB)

1968-10-01

The irradiation of seeds with the fast neutron of research reactors has been hampered by difficulties in accurately measuring dose and in obtaining repeatable and comparable results. Co-ordinated research under an international program organized by the FAO and IAEA has already resulted in significant improvements in methods of exposing seeds in research reactors and in obtaining accurate dosimetry. This has been accomplished by the development of a standard reactor facility for the neutron irradiation of seeds and standard methods for determining fast-neutron dose and the biological response after irradiation. In this program various divisions of the IAEA and the Joint FAO/IAEA Division co-operate with a number of research institutes and reactor centres throughout the world. Results of the preliminary experiments were reported in Technical Reports Series No. 76, ''Neutron Irradiation of Seeds''. This volume contains the proceedings of a meeting of co-operators in the FAO/IAEA Neutron Seed Irradiation Program and other active scientists in this field. The meeting was held in Vienna from 11 to 15 December 1967. Refs, figs and tabs.

Neutron irradiation of seeds 2

International Nuclear Information System (INIS)

1968-01-01

The irradiation of seeds with the fast neutron of research reactors has been hampered by difficulties in accurately measuring dose and in obtaining repeatable and comparable results. Co-ordinated research under an international program organized by the FAO and IAEA has already resulted in significant improvements in methods of exposing seeds in research reactors and in obtaining accurate dosimetry. This has been accomplished by the development of a standard reactor facility for the neutron irradiation of seeds and standard methods for determining fast-neutron dose and the biological response after irradiation. In this program various divisions of the IAEA and the Joint FAO/IAEA Division co-operate with a number of research institutes and reactor centres throughout the world. Results of the preliminary experiments were reported in Technical Reports Series No. 76, ''Neutron Irradiation of Seeds''. This volume contains the proceedings of a meeting of co-operators in the FAO/IAEA Neutron Seed Irradiation Program and other active scientists in this field. The meeting was held in Vienna from 11 to 15 December 1967. Refs, figs and tabs

Microstructural defects in EUROFER 97 after different neutron irradiation conditions

Directory of Open Access Journals (Sweden)

Christian Dethloff

2016-12-01

Full Text Available Characterization of irradiation induced microstructural evolution is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM is used to determine the defect microstructure after different neutron irradiation conditions. In particular dislocation loops, voids and precipitates are analyzed concerning defect nature, density and size distribution after irradiation to 15 dpa at 300 °C in the mixed spectrum High Flux Reactor (HFR. New results are combined with previously obtained data from irradiation in the fast spectrum BOR-60 reactor (15 and 32 dpa, 330 °C, which allows for assessment of dose and dose rate effects on the aforementioned irradiation induced defects and microstructural characteristics.

Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

International Nuclear Information System (INIS)

Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk; Jeong, Jong Ryul; Maeng, Cheol Soo; Lee, Myoung Goo

2016-01-01

As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10"-"5 dpa.

Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

Energy Technology Data Exchange (ETDEWEB)

Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk [KAIST, Daejeon (Korea, Republic of); Jeong, Jong Ryul [Chungnam University, Daejeon (Korea, Republic of); Maeng, Cheol Soo; Lee, Myoung Goo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10{sup -5} dpa.

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

2007-07-01

Full text of publication follows: Neutron irradiation of 9-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects that cause an increase in yield stress and ultimate tensile strength. This irradiation hardening causes embrittlement, which is observed in Charpy impact and toughness tests as an increase in ductile-brittle transition temperature (DBTT). Based on observations that show little change in strength in these steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above this irradiation-hardening temperature regime. In a recent study of F82H steel irradiated at 300, 380, and 500 deg. C, irradiation hardening-an increase in yield stress-was observed in tensile specimens irradiated at the two lower temperatures, but no change was observed for the specimens irradiated at 500 deg. C. As expected, an increase in DBTT occurred for the Charpy specimens irradiated at 300 and 380 deg. C. However, there was an unexpected increase in the DBTT of the specimens irradiated at 500 deg. C. The observed embrittlement was attributed to the irradiation-accelerated precipitation of Laves phase. This conclusion was based on results from a detailed thermal aging study of F82H, in which tensile and Charpy specimens were aged at 500, 550, 600, and 650 deg. C to 30,000 h. These studies indicated that there was a decrease in yield stress at the two highest temperatures and essentially no change at the two lowest temperatures. Despite the strength decrease or no change, the DBTT increased for Charpy specimens irradiated at all four temperatures. Precipitates were extracted from thermally aged specimens, and the amount of precipitate was correlated with the increase in transition temperature. Laves phase was identified in the extracted precipitates by X-ray diffraction. Earlier studies on conventional elevated-temperature steels also showed embrittlement effects above the irradiation-hardening temperature

IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

International Nuclear Information System (INIS)

Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

2002-01-01

Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

Microstructural examination of 12% Cr martensitic stainless steel after irradiation at elevated temperatures in FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Hsu, Chen-Yih; Gelles, D.S.; Lechtenberg, T.A.

1986-06-01

A remelted 12% Cr martensitic stainless steel (HT-9) has been examined by transmission electron microscopy before and after irradiation in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The irradiation temperatures were 365,420, 520, and 600 degree C with the fluences as high as 7.3 x 10 22 n/cm 2 (E > 0.1 MeV) or 34 dpa. The extracted precipitates from each specimen were identified using x-ray microanalysis and selected area diffraction. The precipitates in the unirradiated condition were primarily M 23 C 6 carbides, which formed at martensite lath and prior austenite grain boundaries. During irradiation at elevated temperatures, small amounts of other phases formed, which were tentatively identified as the chromium-rich α', the nickel-silicon rich G-phase, and the intermetallic Chi phase. Irradiation-induced voids were observed only in specimens irradiated at 420 degree C to a dose of 34 dpa; no voids were found for specimens irradiated at 365, 520, and 600 degree C (∼11, ∼34, and ∼34 dpa). These results are not in agreement with previous experiments in that voids have not been reported in this alloy at relatively high fluence level (∼67 dpa) following irradiation in another fast-spectrum reactor (EBR.II). This is, however, the first observation following FFTF irradiation. The present results indicate that cavities can form in HT-9 at modest fluence levels even without significant generation of helium. Hence, the cavity formation in this class of ferritic alloys is not simply caused by helium generation but rather more complex mechanisms. 12 refs., 2 figs., 3 tabs

Martensitic transformation in SUS304 steels with the same Ni equivalent

International Nuclear Information System (INIS)

Ueda, T.; Okino, Y.; Takahashi, S.; Echigoya, J.; Kamada, Y.

2003-01-01

The behavior of martensitic transformation due to plastic deformation at room temperature was investigated in SUS304 austenitic stainless steels with the same nickel equivalent. The absolute volume of the martensitic phase was obtained by saturation magnetization. We discuss the shapes of the martensitic phase caused by different values of coercive force. Martensitic transformation depends on the applied stress but is independent of nickel content with same nickel equivalent. We investigated applications to nondestructive testing on the basis of the present study. (author)

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

Science.gov (United States)

Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

Use of double and triple-ion irradiation to study the influence of high levels of helium and hydrogen on void swelling of 8-12% Cr ferritic-martensitic steels

Science.gov (United States)

Kupriiyanova, Y. E.; Bryk, V. V.; Borodin, O. V.; Kalchenko, A. S.; Voyevodin, V. N.; Tolstolutskaya, G. D.; Garner, F. A.

2016-01-01

In accelerator-driven spallation (ADS) devices, some of the structural materials will be exposed to intense fluxes of very high energy protons and neutrons, producing not only displacement damage, but very high levels of helium and hydrogen. Unlike fission flux-spectra where most helium and hydrogen are generated by transmutation in nickel and only secondarily in iron or chromium, gas production in ADS flux-spectra are rather insensitive to alloy composition, such that Fe-Cr base ferritic alloys also generate very large gas levels. While ferritic alloys are known to swell less than austenitic alloys in fission spectra, there is a concern that high gas levels in fusion and especially ADS facilities may strongly accelerate void swelling in ferritic alloys. In this study of void swelling in response to helium and hydrogen generation, irradiation was conducted on three ferritic-martensitic steels using the Electrostatic Accelerator with External Injector (ESUVI) facility that can easily produce any combination of helium to dpa and/or hydrogen to dpa ratios. Irradiation was conducted under single, dual and triple beam modes using 1.8 MeV Cr+3, 40 keV He+, and 20 keV H+. In the first part of this study we investigated the response of dual-phase EP-450 to variations in He/dpa and H/dpa ratio, focusing first on dual ion studies and then triple ion studies, showing that there is a diminishing influence on swelling with increasing total gas content. In the second part we investigated the relative response of three alloys spanning a range of starting microstructure and composition. In addition to observing various synergisms between He and H, the most important conclusion was that the tempered martensite phase, known to lag behind the ferrite phase in swelling in the absence of gases, loses much of its resistance to void nucleation when irradiated at large gas/dpa levels.

Evaluation of neutron flux in the WWR-SM reactor channel and in the irradiating zone of U-150 cyclotron

International Nuclear Information System (INIS)

Sadikov, I.I.; Zinov'ev, V.G.; Sadikova, Z.O.; Salimov, M.I.

2006-01-01

Full text: For effective work of a reactor, and correct planning of experiments related to the reactor irradiation of various materials it is required to control a neutron flux in the given irradiation point for a long irradiation period. For realization of research works on topazes ennobling under irradiation by reactor neutrons as well as by secondary neutrons produced in a cyclotron it is necessary to know the total neutron flux and spectra. To resolve the problem a technique for registration of neutrons with different energy and calculation of a neutrons spectrum in the given irradiation points in reactor channels and in cyclotron behind the nickel target has been developed. Neutron flux density and energy spectra were monitored by use of the following nuclear reactions: 59 Co(n,γ) 60 Co, 197 Au(n,γ) 198 Au, 58 Ni(n,p) 58 Co, 24 Mg(n,p) 24 Na, 48 Ti(n,p) 48 Sc, 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 89 Y(n,2n) 88 Y, 60 Ni(np) 60 Co. Gamma spectrometer composed of HPGe detector (Rel. Eff. - 15%) and Digital Spectra Analyzer DSA-1000 (Canberra Ind., USA) was used to measure gamma activity of irradiated samples. Acquired gamma spectra were processed by means of Genie 2000 standard software package. The σ(E) functions and neutron spectra were calculated by using the least squares method and approximating the tabular and experimental data with power polynomials. The developed technique was applied for the adjustment of the topazes irradiation regimes in the reactor core and under secondary neutrons flux from a nickel target in the cyclotron. The given technique allows to calculate a logarithmic spectrum of neutrons in a energy range from 0,025 eV up to 12 MeV with the uncertainty of about 10 %. (author)

Radiological risks from irradiation of cargo contents with EURITRACK neutron inspection systems

International Nuclear Information System (INIS)

Giroletti, E.; Bonomi, G.; Donzella, A.; Viesti, G.; Zenoni, A.

2012-01-01

The radiological risk for the population related to the neutron irradiation of cargo containers with a tagged neutron inspection system has been studied. Two possible effects on the public health have been assessed: the modification of the nutritional and organoleptic properties of the irradiated materials, in particular foodstuff, and the neutron activation of consumer products (i.e. food and pharmaceuticals). The result of this study is that irradiation of food and foodstuff, pharmaceutical and medical devices in container cargoes would neither modify the properties of the irradiated material nor produce effective doses of concern for public health. Furthermore, the dose received by possible stowaways present inside the container during the inspection is less than the annual effective dose limit defined by European Legislation for the public. - Highlights: ► Neutron irradiation of cargo containers implies a radiological risk. ► The risk is about the modification of food properties and the products activation. ► Assessment is made about the EURITRACK neutron irradiation system. ► Results show that the EURITRACK scanning is not dangerous for the population.

DAMAGE IN MOLYBDENUM ASSOCIATED WITH NEUTRON IRRADIATION AND SUBSEQUENT POST-IRRADIATION ANNEALING

Energy Technology Data Exchange (ETDEWEB)

Mastel, B.

1963-07-23

Molybdemum containing carbon was studied in an attempt to establish the combined effect of impurity content and neutron irradiation on the properties and structure of specific metals. Molybdenum foils were punched into discs and heat treated in vacuum. They were then slow-cooled and irradiated. After irradiation and subsequent decay of radioactivity to a low level the foils were subjected to x-ray diffraction measurements. Cold-worked foils with less than 10 ppm carbon showed no change in microstructure due to irradiation. Molybdenum foils that were annealed prior to irradiation showed spot defects. In foils containing up to 500 ppm carbon, it was concluded that the small loops present after irradiation are due to the clustering of point defects at interstitial carbon atoms, followed by collapse to form a dislocation loop. The amount of lattice expansion after irradiation was strongly dependent on impurity content. Neutron irradiation was found to reduce the number of active slip systems. (M.C.G.)

Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

International Nuclear Information System (INIS)

Kim, S.W.; Tanigawa, H.; Hirose, T.; Kohyama, A.

2007-01-01

Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

Energy Technology Data Exchange (ETDEWEB)

Kim, S.W. [Kyoto Univ., Graduate School of Energy Science (Japan); Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan)

2007-07-01

Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

Biomedical irradiation system for boron neutron capture therapy at the Kyoto University Reactor

International Nuclear Information System (INIS)

Kobayashi, T.; Kanda, K.; Ujeno, Y.; Ishida, M.R.

1990-01-01

Physics studies related to radiation source, spectroscopy, beam quality, dosimetry, and biomedical applications using the Kyoto University Reactor Heavy Water Facility are described. Also, described are a Nickel Mirror Neutron Guide Tube and a Super Mirror Neutron Guide Tube that are used both for the measurement of boron concentration in phantom and living tissue and for precise measurements of neutron flux in phantom in the presence of both light and heavy water. Discussed are: (1) spectrum measurements using the time of flight technique, (2) the elimination of gamma rays and fast neutrons from a thermal neutron irradiation field, (3) neutron collimation without producing secondary gamma rays, (4) precise neutron flux measurements, dose estimation, and the measurement of boron concentration in tumor and its periphery using guide tubes, (5) the dose estimation of boron-10 for the first melanoma patient, and (6) special-purpose biological irradiation equipment. Other related subjects are also described

Biomedical irradiation system for boron neutron capture therapy at the Kyoto University reactor

International Nuclear Information System (INIS)

Kobayashi, T.; Kanda, K.; Ujeno, Y.; Ishida, M.R.

1990-01-01

Physics studies related to radiation source, spectroscopy, beam quality, dosimetry, and biomedical applications using the Kyoto University Reactor Heavy Water Facility are described. Also, described are a Nickel Mirror Neutron Guide Tube and a Super Mirror Neutron Guide Tube that are used both for the measurement of boron concentration in phantom and living tissue and for precise measurements of neutron flux in phantom in the presence of both light and heavy water. Discussed are: (1) spectrum measurements using the time of flight technique, (2) the elimination of gamma rays and fast neutrons from a thermal neutron irradiation field, (3) neutron collimation without producing secondary gamma rays, (4) precise neutron flux measurements, dose estimation, and the measurement of boron concentration in tumor and its periphery using guide tubes, (5) the dose estimation of boron-10 for the first melanoma patient, and (6) special-purpose biological irradiation equipment. Other related subjects are also described

An assessment of magnetic effects in ferromagnetic martensitic steels for use in fusion machines

International Nuclear Information System (INIS)

Lechtenberg, T.; Dahms, C.; Attaya, H.

1984-01-01

Interest in the 9-12%Cr class of martensitic stainless steels has accelerated since these materials were included in the U.S. Alloy Development for Irradiation Performance (ADIP) task funded by the Office of Fusion Energy in 1979. This program is focused on developing structural materials for fusion reactor first wall/breeding blanket components where the neutron damage is most severe. This area of a fusion reactor will be required to tolerate damage levels on the order of 110 dpa( 1 ). As a part of ADIP, study of the martensitic steels is focused on establishing the feasibility of using these materials. The interest in martensitic steels stems from their potential to tolerate high levels of radiation damage without significant degradation of material properties. Martensitic steels have a body-centered-cubic crystal structure that, unlike face-centered-cubic structure of austenitic steels, exhibits very little swelling under neutron irradiation( 2 ). One of the outstanding issues with martensitic steels is the possible parasitic stresses associated with ferromagnetic interaction with the magnetic fields. This paper is divided into two parts, the first reviews previous work on magnetic effects to the structure and plasma; the second presents new calculations of stresses on a coolant pipe in a Starfire model assumed to be made of 12Cr-1Mo steel(HT-9)

Effects of post-irradiation annealing on the transformation behavior of Ti-Ni alloys

International Nuclear Information System (INIS)

Kimura, A.; Tsuruga, H.; Morimura, T.; Misawa, T.; Miyazaki, S.

1993-01-01

Recovery processes of martensitic transformation of neutron irradiated Ti-50.0, 50.5 and 51.0 at.%Ni alloys during post-irradiation annealing were investigated by means of differential scanning calorimetry (DSC), tensile tests and transmission electron microscope (TEM) observations. Neutron irradiation up to a fluence of 1.2x10 24 n/cm 2 at 333 K suppressed the martensitic transformation as well as the stress-induced martensitic transformation of these alloys above 150 K. The TEM observations revealed that the disordered zones containing small defect clusters in high density were formed in the neutron irradiated Ti-Ni alloys. The DSC measurements also showed that the post-irradiation annealing caused recovery of the transformation of which the progress depended on the annealing temperature and period. A significant retardation of the recovery was recognized in the Ti-51.0 at.%Ni alloy in comparison with the Ti-50.0 at.%Ni alloy. From the shifts in the transformation temperature upon isothermal annealing at various annealing temperatures, the activation energies of the recovery process of the transformation in the neutron irradiated Ti-50.0 and 51.0 at.%Ni alloys were evaluated by a cross-cut method to be 1.2 eV and 1.5 eV, respectively. The recovery of the transformation was ascribed to the re-ordering resulting from decomposition of vacancy clusters, and those obtained values of the activation energy were considered to be the sum of the migration energy of vacancy and the binding energy of vacancy-vacancy cluster. The retardation of the recovery in the Ti-51.0 at%Ni alloy was interpreted in terms of large binding energy in this alloy due to the off-stoichiometry. (author)

Void formation in irradiated binary nickel alloys

International Nuclear Information System (INIS)

Shaikh, M.A.; Ahmed, M.; Akhter, J.I.

1994-01-01

In this work a computer program has been used to compute void radius, void density and swelling parameter for nickel and binary nickel-carbon alloys irradiated with nickel ions of 100 keV. The aim is to compare the computed results with experimental results already reported

Study of points defects produced by irradiation of monocrystalline nickel and polycrystalline gadolinium

International Nuclear Information System (INIS)

Cope, R.

1969-07-01

The work described in this thesis falls into two parts: the first comprises a study of magnetocrystalline nickel by resistivity measurements; the second is a description of resistivity and magnetic after effect measurements on an h.c.p. ferromagnetic crystal other than cobalt, namely gadolinium. For the first part we have demonstrated the existence of a small but definite orientation dependence in the creation of point defects by electron irradiation (20 deg. K) of a nickel single crystal. In particular, the effect is manifested in the form of the stage I C , II and III in the resistivity recovery. In the second part an important result has emerged: namely that there is no magnetic after effect phenomenon in a neutron irradiated (27 deg. K) ferromagnetic metal. Several considerations are discussed by way of a preliminary interpretation of this important difference between gadolinium and cobalt. (author) [fr

Concerning the problem of the plastic deformation mechanism changeover in neutron-irradiated metals and alloys

International Nuclear Information System (INIS)

Kolesnikov, A.N.; Krasnoselov, V.A.; Prokhorov, V.I.

1982-01-01

With a phenomenological model of plastic deformation instability as a basis, an analysis was made of the neutron irradition effects on the characteristics of strength and plasticity vs. structural parameters and radiation damage morphology. It was demonstrated that the enchanced plasticity in the initial stage of neutron irradiation has to do with the solid solution disintegration. Introduction of indestructible strengthening barriers enhances the stress-resistance of the neck-formation by 1.22 times. The ''big grain'' effect is observable during the deformation channel production only. Both the deformation twinning and deformation-induced martensite transformation raise the plastic flow stability

Martensite phase stress and the strengthening mechanism in TRIP steel by neutron diffraction.

Science.gov (United States)

Harjo, Stefanus; Tsuchida, Noriyuki; Abe, Jun; Gong, Wu

2017-11-09

Two TRIP-aided multiphase steels with different carbon contents (0.2 and 0.4 mass%) were analyzed in situ during tensile deformation by time-of-flight neutron diffraction to clarify the deformation induced martensitic transformation behavior and its role on the strengthening mechanism. The difference in the carbon content affected mainly the difference in the phase fractions before deformation, where the higher carbon content increased the phase fraction of retained austenite (γ). However, the changes in the relative fraction of martensitic transformation with respect to the applied strain were found to be similar in both steels since the carbon concentrations in γ were similar regardless of different carbon contents. The phase stress of martensite was found much larger than that of γ or bainitic ferrite since the martensite was generated at the beginning of plastic deformation. Stress contributions to the flow stress were evaluated by multiplying the phase stresses and their phase fractions. The stress contribution from martensite was observed increasing during plastic deformation while that from bainitic ferrite hardly changing and that from γ decreasing.

The influence of nickel content on microstructures of Fe-Cr-Ni austenitic alloys irradiated with nickel ions

International Nuclear Information System (INIS)

Muroga, T.; Yoshida, N.; Garner, F.A.

1990-11-01

The objectives of this effort is to identify the mechanisms involved in the radiation-induced evolution of microstructure in materials intended for fusion applications. The results of this study are useful in interpreting the results of several other ongoing experiments involving either spectral or isotopic tailoring to study the effects of helium on microstructure evolution. Ion-irradiated Fe-15Cr-XNi (X = 20, 35, 45, 60, 75) ternary alloys and a 15Cr-85Ni binary alloy were examined after bombardment at 675 degree C and compared to earlier observations made on these same alloys after irradiation in EBR-II at 510 or 538 degree C. The response of the ion-irradiated microstructures to nickel content appears to be very consistent with that of neutron irradiation even though there are four orders of magnitude difference in displacement rate and over 200 degree C difference in temperature. It appears that the transition to higher rates of swelling during both types of irradiation is related to the operation of some mechanisms that is not directly associated with void nucleation. 6 refs., 8 figs

A comparison of dilatometry and in-situ neutron diffraction in tracking bulk phase transformations in a martensitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Christien, F., E-mail: frederic.christien@univ-nantes.fr [Institut des Matériaux Jean Rouxel (IMN), Université de Nantes, CNRS, Rue Christian Pauc, BP 50609, 44306 Nantes Cedex 3 (France); Telling, M.T.F. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, OX11 0QX (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford (United Kingdom); Knight, K.S. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, OX11 0QX (United Kingdom); Department of Earth Sciences, The Natural History Museum, Cromwell Road, London (United Kingdom)

2013-08-15

Phase transformations in the 17-4PH martensitic stainless steel have been studied using different in-situ techniques, including dilatometry and high resolution neutron diffraction. Neutron diffraction patterns were quantitatively processed using the Rietveld refinement method, allowing the determination of the temperature-dependence of martensite (α′, bcc) and austenite (γ, fcc) phase fractions and lattice parameters on heating to 1000 °C and then cooling to room temperature. It is demonstrated in this work that dilatometry doesn't permit an accurate determination of the end temperature (Ac3) of the α′ → γ transformation which occurs upon heating to high temperature. The analysis of neutron diffraction data has shown that the respective volumes of the two phases become very close to each other at high temperature, thus making the dilatometric technique almost insensitive in that temperature range. However, there is a very good agreement between neutron diffraction and dilatometry at lower temperature. The martensitic transformation occurring upon cooling has been analysed using the Koistinen–Marburger equation. The thermal expansion coefficients of the two phases have been determined in addition. A comparison of the results obtained in this work with data from literature is presented. - Highlights: • Martensite is still present at very high temperature (> 930 °C) upon heating. • The end of austenitisation cannot be accurately monitored by dilatometry. • The martensite and austenite volumes become similar at high temperature (> ∼ 850 °C)

Behavior of implanted hydrogen in ferritic/martensitic steels under irradiation

Science.gov (United States)

Wan, F.; Takahashi, H.; Ohnuki, S.; Nagasaki, R.

1988-07-01

The aim of this study was to clarify the behavior of hydrogen under irradiation in ferritic/martensitic stainless steel Fe-10Cr-2Mo-1Ni. Hydrogen was implanted into the specimens by ion accelerator or chemical cathodic charging method, followed by electron irradiation in a HVEM at temperatures from room temperature to 773 K. Streaks in the electron diffraction patterns were observed only during electron irradiation at 623-723 K. From these results it is suggested that the occurrence of the streak pattern is due to the formation of radiation-induced complexes of Ni or Cr with hydrogen along directions.

Neutron resistant irradiation alloy and usage thereof

International Nuclear Information System (INIS)

Okada, Osamu; Nakata, Kiyotomo; Kato, Takahiko.

1997-01-01

A neutron irradiation embrittlement-resistant alloy comprising a Ti alloy having an average grain size of 2μm or smaller and containing from 30 to 40wt% of Al is subjected to powder solidification and then to isothermal forging at a forging rate of from 50 to 80% at a temperature range of from 1150 to 1500K. Namely, since the Ti-Al type alloy comprises from 30 to 30wt% of Al, optionally, from 1 to 6% of Mn, from 0.1 to 0.5% of Si, from 4 to 16% of V and the balance of Ti, it has excellent specific strength, high durable temperature and excellent neutron irradiation resistance, and has ductility required as structural materials. Accordingly, if the Ti-Al type alloy excellent in embrittlement resistance to neutron irradiation dimensional stability of materials is applied to constitutional parts of a reactor core of a nuclear reactor and a thermonuclear reactor to be exposed under neutron irradiation, high reliability is provided and the amount of activated materials is reduced by improving the working life of the materials. (N.H.)

Tentative investigation on neutron mirror fabrication with electroless nickel plating

International Nuclear Information System (INIS)

Guo, Jiang; Morita, Shin-ya; Yamagata, Yutaka; Takeda, Shin; Kato, Jun-ichi; Hino, Masahiro; Furusaka, Michihiro

2013-01-01

Neutron optics becomes highly required due to the rapid development of neutron technology. In this paper, we attempt to fabricate the neutron mirror by using a metal substrate made of electroless nickel plating to take place of glass concerning about mirror's optical performance and manufacturing method. A new manufacture process chain of neutron mirror is proposed by following the steps of fast milling and precision cutting of aluminium/stainless, electroless nickel plating, ultra-precision cutting by diamond tools, super-smooth polishing and super mirror coating to obtain high form accuracy and good surface roughness time-efficiently. Some tentative investigations are carried out. A workpiece (□ 50 x 50 mm 2 ) with flat surface made of electroless nickel plating is machined by ultra-precision cutting and polishing. The surface roughness with 0.728 nm rms (0.588 nm Ra) is acquired. According to results of reflectometry, the neutron beam can be reflected effectively with high intensity and little scattering. (author)

The effect of nitrogen on martensite formation in a Cr-Mn-Ni stainless steel

International Nuclear Information System (INIS)

Biggs, T.; Knutsen, R.D.

1995-01-01

The influence of nitrogen (0 to 0.27 wt%) on martensite formation in an experimental low-nickel stainless-steel alloy (Fe-17Cr-7Mn-4Ni) has been investigated. The alloys containing 0.1 wt% or more nitrogen are fully austenitic at room temperature; those containing less nitrogen consist of a mixture of austenite, martensite and δ-ferrite. The alloys containing less than 0.2 wt% nitrogen are metastable and undergo a transformation from austenite to martensite on deformation. Transmission electron microscopy investigations suggest that, within the nitrogen range considered in this investigation, the addition of nitrogen causes an increase in stacking fault energy which in turn inhibits the nucleation of martensite. As the low-nitrogen alloys (less than 0.2 wt% nitrogen) undergo deformation, ε-martensite (with the [ anti 110] γ and [ anti 12 anti 10] ε zone axes parallel) is observed at the intersection of stacking faults. With increasing strain, the presence of α'-martensite is observed in conjunction with the ε-martensite, and only α'-martensite is observed at very high strains. Both the Nishiyama-Wasserman and Kurdjumov-Sachs orientation relationships are observed between austenite and α'-martensite. The transformation to martensite during deformation causes a significant variation in room-temperature mechanical properties, despite the overall narrow range in composition considered. (orig.)

A neutron irradiator applied to cancer treatment

International Nuclear Information System (INIS)

Campos, Tarcisio P.R.; Andrade, Ana P. de

2000-01-01

Cancer and the way of treating it with neutron capture therapy are addressed. This paper discusses also the type of neutron facilities used to treat cancer around the world, as follow: discrete neutron sources, accelerators, and nuclear reactors. The major features of an epithermal neutron irradiation facility applied to BNCT treatment are addressed. The main goal is to give another choice of neutron irradiators to be set in a hospital. The irradiation facility embeds a set of 252 Cf neutron source coupled with a homogeneous mixture of uranium-zirconium hydride alloy containing 8.4 wt % uranium enriched to 20% U 235 . The facility delivers an epithermal neutron beam with low background of fast neutron and gamma rays. The N particle transport code (MCNP-4A) has been used during the simulation in order to achieve the desired configurations and to estimate the multiplication factor, k eff . The present facility loaded with 30 mg of 252 Cf neutron source generates an external beam with an intensity of 10 7 n/cm 2 .s on the spectrum of 4 eV to 40 KeV. The 252 Cf - facility coupled with fissile material was able to amplify the epithermal flux to 10 8 n/cm 2 .s, maintaining the figure-of-merits represented by the ratios of the fast dose and gamma dose in air per epithermal neutron flux closed to those values presented by BMRR, MITR-II and Petten Reactor. The medical irradiation facility loaded with 252 Cf- 235 U can be a choice for BNCT. (author)

Effect of phase instabilities on the correlation of nickel ion and neutron irradiation swelling in solution annealed 316 stainless steel

International Nuclear Information System (INIS)

Rowcliffe, A.F.; Lee, E.H.; Sklad, P.S.

1979-01-01

Annealed 316 stainless steel specimens were neutron irradiated to establish steady-state microstructures and then subjected to further high temperature irradiations with 4 MeV Ni ions. It is shown that void growth under neutron irradiation is simulated in ion irradiations carried out at approx. 180 0 C above reactor temperature. However, the precipitate microstructure developed during neutron irradiation is unstable during subsequent ion irradiation. As a result, the relative swelling rates at various reactor temperatures are not simulated correctly

Neutron irradiation facility and its characteristics

International Nuclear Information System (INIS)

Oyama, Yukio; Noda, Kenji

1995-01-01

A neutron irradiation facility utilizing spallation reactions with high energy protons is conceived as one of the facilities in 'Proton Engineering center (PEC)' proposed at JAERI. Characteristics of neutron irradiation field of the facility for material irradiation studies are described in terms of material damage parameters, influence of the pulse irradiation, irradiation environments other than neutronics features, etc., comparing with the other sorts of neutron irradiation facilities. Some perspectives for materials irradiation studies using PEC are presented. (author)

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

International Nuclear Information System (INIS)

Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

2012-01-01

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different 10 B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

Energy Technology Data Exchange (ETDEWEB)

Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2012-07-15

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Materials irradiation research in neutron science

Energy Technology Data Exchange (ETDEWEB)

Noda, Kenji; Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-11-01

Materials irradiation researches are planned in Neutron Science Research Program. A materials irradiation facility has been conceived as one of facilities in the concept of Neutron Science Research Center at JAERI. The neutron irradiation field of the facility is characterized by high flux of spallation neutrons with very wide energy range up to several hundred MeV, good accessibility to the irradiation field, good controllability of irradiation conditions, etc. Extensive use of such a materials irradiation facility is expected for fundamental materials irradiation researches and R and D of nuclear energy systems such as accelerator-driven incineration plant for long-lifetime nuclear waste. In this paper, outline concept of the materials irradiation facility, characteristics of the irradiation field, preliminary technical evaluation of target to generate spallation neutrons, and materials researches expected for Neutron Science Research program are described. (author)

Impurity effects in neutron-irradiated simple oxides: Implications for fusion devices

International Nuclear Information System (INIS)

Gonzalez, R.; Chen, Y.; Caceres, D.; Vergara, I.

2006-01-01

Radiation damage induced by neutron irradiation was studied in undoped MgO crystals and in MgO doped with either iron, hydrogen or lithium impurities. The oxygen-vacancy concentration produced by irradiation increases with neutron fluence. The net production rates resulting from irradiations with 14.8 MeV neutrons are about twice those produced by fission neutrons. In nominally pure crystals, the oxygen-vacancy concentration incurred by the fission-neutron irradiation is higher in crystals with a larger number of inherent impurities (such as iron) due to trapping of interstitials by impurities. Suppression of these defects is observed in MgO:H crystals and attributed to migration of oxygen vacancies to microcavities filled with H 2 gas. In MgO:Li crystals irradiated with neutron fluences below 10 18 n/cm 2 , most of the oxygen vacancies are camouflaged as hydride ions. Nanoindentation experiments show that hardness increases with neutron fluence and is independent of the presence of lithium in the crystal. Comparison between a neutron-irradiated and a thermochemically reduced crystal containing similar concentrations of oxygen vacancies shows that 70% of the neutron-irradiation hardening is produced by interstitials, 30% by oxygen vacancies and a negligible amount by higher-order point defects

Recovery and purification of nickel-63 from HFIR-irradiated targets

International Nuclear Information System (INIS)

Williams, D.F.; O'Kelley, G.D.; Knauer, J.B.; Porter, C.E.; Wiggins, J.T.

1993-06-01

The production of large quantities of high-specific-activity 63 Ni (>10 Ci/g) requires both a highly enriched 62 Ni target and a long irradiation period at high neutron flux. Trace impurities in the nickel and associated target materials are also activated and account for a significant fraction of the discharged activity and essentially all of the gamma activity. While most of these undesirable activation products can be removed as chloride complexes during anion exchange, chromium, present at 51 Cr, and scandium, present as 46 Sc, are exceptions and require additional processing to achieve the desired purity. Optimized flowsheets are discussed based upon the current development and production experience

Depth distribution of martensite in xenon implanted stainless steels

International Nuclear Information System (INIS)

Johansen, A.; Johnson, E.; Sarholt-Kristensen, L.; Steenstrup, S.; Hayashi, N.; Sakamoto, I.

1989-01-01

The amount of stress-induced martensite and its distribution in depth in xenon implanted austenitic stainless steel poly- and single crystals have been measured by Rutherford backscattering and channeling analysis, depth selective conversion electron Moessbauer spectroscopy, cross-sectional transmission electron microscopy and x-ray diffraction analysis. In low nickel 17/7, 304 and 316 commercial stainless steels and in 17:13 single crystals the martensitic transformation starts at the surface and develops towards greater depth with increasing xenon fluence. The implanted layer is nearly completely transformed, and the interface between martensite and austenite is rather sharp and well defined. In high nickel 310 commercial stainless steel and 15:19 and 20:19 single crystals, on the other hand, only insignificant amounts of martensite are observed. (orig.)

Use of double and triple-ion irradiation to study the influence of high levels of helium and hydrogen on void swelling of 8–12% Cr ferritic-martensitic steels

Energy Technology Data Exchange (ETDEWEB)

Kupriiyanova, Y.E., E-mail: fomenkoj@kipt.kharkov.ua [National Science Centre Kharkov Institute of Physics and Technology, 1, Akademicheskaya St., Kharkov, 61108 (Ukraine); Bryk, V.V.; Borodin, O.V.; Kalchenko, A.S.; Voyevodin, V.N.; Tolstolutskaya, G.D. [National Science Centre Kharkov Institute of Physics and Technology, 1, Akademicheskaya St., Kharkov, 61108 (Ukraine); Garner, F.A. [Radiation Effects Consulting, Richland, WA 99354 (United States)

2016-01-15

In accelerator-driven spallation (ADS) devices, some of the structural materials will be exposed to intense fluxes of very high energy protons and neutrons, producing not only displacement damage, but very high levels of helium and hydrogen. Unlike fission flux-spectra where most helium and hydrogen are generated by transmutation in nickel and only secondarily in iron or chromium, gas production in ADS flux-spectra are rather insensitive to alloy composition, such that Fe–Cr base ferritic alloys also generate very large gas levels. While ferritic alloys are known to swell less than austenitic alloys in fission spectra, there is a concern that high gas levels in fusion and especially ADS facilities may strongly accelerate void swelling in ferritic alloys. In this study of void swelling in response to helium and hydrogen generation, irradiation was conducted on three ferritic-martensitic steels using the Electrostatic Accelerator with External Injector (ESUVI) facility that can easily produce any combination of helium to dpa and/or hydrogen to dpa ratios. Irradiation was conducted under single, dual and triple beam modes using 1.8 MeV Cr{sup +3}, 40 keV He{sup +}, and 20 keV H{sup +}. In the first part of this study we investigated the response of dual-phase EP-450 to variations in He/dpa and H/dpa ratio, focusing first on dual ion studies and then triple ion studies, showing that there is a diminishing influence on swelling with increasing total gas content. In the second part we investigated the relative response of three alloys spanning a range of starting microstructure and composition. In addition to observing various synergisms between He and H, the most important conclusion was that the tempered martensite phase, known to lag behind the ferrite phase in swelling in the absence of gases, loses much of its resistance to void nucleation when irradiated at large gas/dpa levels.

The microstructure of neutron-irradiated Fe-Cr alloys: A small-angle neutron scattering study

International Nuclear Information System (INIS)

Heintze, C.; Bergner, F.; Ulbricht, A.; Eckerlebe, H.

2009-01-01

Ferritic-martensitic chromium steels are candidate materials for future applications in both Gen-IV fission and fusion technology. Experimental investigation of neutron-irradiated Fe-Cr model alloys is important in order to gain a better understanding of the interplay of chromium content and irradiation behaviour. Small-angle neutron scattering (SANS) is particularly well suited to unfold the size distribution of non-planar irradiation-induced nanoscale features such as defect-solute clusters, nanovoids and α'- particles. This size distribution represents a statistically reliable average over a macroscopic volume. Assumptions on the dominant type of features can be checked against the ratio of magnetic and nuclear scattering. The materials investigated in this work are commercial-purity Fe-Cr alloys of nominal compositions Fe-2.5Cr, Fe-5Cr, Fe- 9Cr and Fe-12.5Cr (at %). Neutron irradiation was performed in the reactor BR2 at Mol (Belgium) at a temperature of 300 deg. C and neutron flux of 9 x 10 13 cm -2 s -1 (E > 1 MeV) [Matijasevic, JNM 377 (2008) 147]. The neutron exposures expressed in units of displacements per atom correspond to 0.6 and 1.5 dpa. A wavelength of 0.58 nm and three detector-sample distances of 1, 4 and 16 m were used in the SANS experiments carried out at the SANS-2 facility of GKSS Geesthacht (Germany). The samples were placed in a saturation magnetic field in order to separate magnetic and nuclear contributions. The scattering curves obtained for the unirradiated conditions of the four Fe-Cr alloys were taken as reference. We have found that the volume fraction of scatterers slightly increases with neutron exposure (Fe-9Cr) or exhibits a saturation-like behaviour (Fe-2.5Cr, Fe-5Cr and Fe-12.5Cr) and that the volume fraction at 1.5 dpa is an increasing function of the chromium level with a slight increase up to 9 at%Cr and a steep increase between 9 and 12.5 at%Cr. The radii of irradiation-induced scatterers are essentially less than 8 nm and

Proceedings of neutron irradiation technical meeting on BNCT

International Nuclear Information System (INIS)

2000-10-01

The 'Neutron Irradiation Technical Meeting for Boron Neutron Capture Therapy (BNCT)' was held on March 13, 2000 at Tokai Research Establishment. The Meeting is aimed to introduce the neutron beam facility for medical irradiation at JRR-4 to Japanese researchers widely, as well as providing an opportunity for young researchers, engineers, medical representatives such surgeons and doctors of pharmacology to present their research activities and to exchange valuable information. JAERI researcher presented the performance and the irradiation technology in the JRR-4 neutron beam facility, while external researchers made various and beneficial presentations containing such accelerator-based BNCT, spectrum-shifter, biological effect, pharmacological development and so on. In this meeting, a special lecture titled 'The Dawn of BNCT and Its Development.' was given by MD, Prof. Takashi Minobe, an executive director of Japan Foundation for Emergency Medicine. The 11 of the presented papers are indexed individually. (J.P.N.)

OKMC study of the effect of grain boundaries in martensitic Fe-Cr-C alloys

International Nuclear Information System (INIS)

Chiapetto, M.; Becquart, C.S.; Malerba, L.

2015-01-01

Fe-Cr-C alloys with chromium concentrations in the range from about 2 wt % to 12 wt % form ferritic-martensitic structures by rapid cooling from the austenite state already in the presence of relatively low carbon concentrations. In this process it is possible to obtain different ratios of ferrite and martensite, as well as formation of carbides, by varying the thermal treatment. The presence of ferrite or martensite might have an influence on the nano-structural evolution under irradiation of these alloys. Here, considering a tempered martensite reference alloy with 9% Cr, we make use of an already validated object kinetic Monte Carlo (OKMC) model in order to study the possible effect of the formation of martensite laths on the material nano-structural evolution under neutron irradiation, assuming that the relevant boundaries act as sinks for radiation defects. The results show that the reduction of the grain size (including in this definition the average size of prior austenite grains, blocks and laths) does not play any relevant role until sizes of the order of about 0.5 μm are reached: for smaller grains the number of defects being absorbed by the boundaries becomes dominant. However, this threshold is lower than the experimentally observed martensite lath dimensions, thereby suggesting that what makes the difference in martensitic Fe-Cr-C alloys with respect to ferrite concerning events and mechanisms taking place during irradiation are not the lath boundaries as sinks. Differences between the nano-structural evolution in ferrite and martensite should therefore be ascribed to other factors. This document is composed of an article and the presentation slides. (authors)

Final report on neutron irradiation at low temperature to investigate plastic instability and at high temperature to study caviation

DEFF Research Database (Denmark)

Singh, B.N; Eldrup, Morten Mostgaard; Golubov, D.J.

2005-01-01

Effects of neutron irradiation on defect accumulation and physical and mechanical properties of pure iron and F82H and EUROFER 97 ferritic-martensitic steels have been investigated. Tensile specimens were neutron irradiated to a dose level of 0,23 dpa at333 and 573 K. Electrical resistivity......, based on the production bias model (PBM) were carried out to study the details of evolution of cavitieswith and without helium generation. The phenomena of dislocation decoration and raft formation, which are important for understanding radiation hardening and plastic flow localization, have been...... studied using the Kinetic Monte Carlo (KMC) code during arealistic dynamic irradiation of bcc iron at 300 K. Molecular dynamics (MD) simulations have been carried out to study the stress dependencies of dislocation velocity and drag coefficient for an edge dislocation decorated with small SIA loops...

Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

International Nuclear Information System (INIS)

Zinkle, S.J.

1998-01-01

The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of ∼5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule

A standard fission neutron irradiation facility

International Nuclear Information System (INIS)

Sahasrabudhe, S.G.; Chakraborty, P.P.; Iyer, M.R.; Kirthi, K.N.; Soman, S.D.

1979-01-01

A fission neutron irradiation facility (FISNIF) has been set up at the thermal column of the CIRUS reactor at BARC. The spectrum and the flux have been measured using threshold detectors. The paper describes the setting up of the facility, measurement and application. A concentric cylinder containing UO 2 powder sealed inside surrounds the irradiation point of a pneumatic sample transfer system located in the thermal column of the reactor. Samples are loaded in a standard aluminium capsule with cadmium lining and transported pneumatically. A sample transfer time of 1 s can be achieved in the facility. Typical applications of the facility for studying activation of iron and sodium in fission neutrons are also discussed. (Auth.)

Plastic flow properties and fracture toughness characterization of unirradiated and irradiated tempered martensitic steels

International Nuclear Information System (INIS)

Spaetig, P.; Bonade, R.; Odette, G.R.; Rensman, J.W.; Campitelli, E.N.; Mueller, P.

2007-01-01

We investigate the plastic flow properties at low and high temperature of the tempered martensitic steel Eurofer97. We show that below room temperature, where the Peierls friction on the screw dislocation is active, it is necessary to modify the usual Taylor's equation between the flow stress and the square root of the dislocation density and to include explicitly the Peierls friction stress in the equation. Then, we compare the fracture properties of the Eurofer97 with those of the F82H steel. A clear difference of the fracture toughness-temperature behavior was found in the low transition region. The results indicate a sharper transition for Eurofer97 than for the F82H. Finally, the shift of the median toughness-temperature curve of the F82H steel was determined after two neutron irradiations performed in the High Flux Reactor in Petten

Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

International Nuclear Information System (INIS)

Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

2003-01-01

Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

Proceedings of neutron irradiation technical meeting on BNCT

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-10-01

The 'Neutron Irradiation Technical Meeting for Boron Neutron Capture Therapy (BNCT)' was held on March 13, 2000 at Tokai Research Establishment. The Meeting is aimed to introduce the neutron beam facility for medical irradiation at JRR-4 to Japanese researchers widely, as well as providing an opportunity for young researchers, engineers, medical representatives such surgeons and doctors of pharmacology to present their research activities and to exchange valuable information. JAERI researcher presented the performance and the irradiation technology in the JRR-4 neutron beam facility, while external researchers made various and beneficial presentations containing such accelerator-based BNCT, spectrum-shifter, biological effect, pharmacological development and so on. In this meeting, a special lecture titled 'The Dawn of BNCT and Its Development.' was given by MD, Prof. Takashi Minobe, an executive director of Japan Foundation for Emergency Medicine. The 11 of the presented papers are indexed individually. (J.P.N.)

Study of neutron irradiation on F82H alloys by Mössbauer spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Huang, S.S., E-mail: h.shaosong@ht8.ecs.kyoto-u.ac.jp; Kitao, S.; Kobayashi, Y.; Yoshiie, T.; Xu, Q.; Sato, K.; Seto, M.

2015-01-15

The effects of neutron irradiation on F82H ferritic/martensitic stainless steel and its model alloys were studied by Mössbauer spectroscopy. The microstructural damage mechanisms of these alloys, during the void incubation period were interpreted using the short range order (SRO) parameters. Results show that within Fe–8Cr alloy, the atoms in the nearest neighbor (NN) of the Fe nuclei were inhomogeneous, prior to irradiation. A configuration trapping model of Cr supported the negative average SRO observed for the NN shells in our Fe–Cr alloys. We found that irradiation also accelerated the SRO in Fe–8Cr through a diffusion mechanism, where Cr atom repulsion was concentration dependent. Finally, comparative studies were conducted on F82H model alloys using the present Mössbauer measurements and our previously reported work on positron annihilation spectroscopy, which further established that irradiation of the alloys promoted the growth of a M{sub 23}C{sub 6} complex.

Development of materials for higher burn-up rates

International Nuclear Information System (INIS)

Ehrlich, K.; Anderko, K.

1983-01-01

The results obtained so far concerning both high-nickel austenitic steels (X10CrNiMoTiB10 25) and ferritic-martensitic steels (X18CrMoVNb12 1 corresponds to 1.4914) justify the expectations placed in these materials. Future R and D work will concentrate on high-nickel austenitic steels, in particular on in-pile residence time in the range of maximum cladding mean temperatures (He embrittlement), on confirmation of the fairly good swelling and creep behaviour to be obtained from neutron irradiation experiments as well as on material qualification and technology. As far the development in the field of martensitic steels the major question concerns the influence of neutron irradiation on the transient temperature. It is just as important and interesting to detect and describe the ruling in-pile creep mechanism in the 400 to 600 0 C range. Furthermore, questions on the fabrication and qualification of cladding tube boxes are also preeminent. (orig.) [de

Microstructural evolution of reduced-activation martensitic steel under single and sequential ion irradiations

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Jin, Shuoxue; Li, Tiecheng; Zheng, Zhongcheng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yang, Feng; Xiong, Xuesong; Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2013-07-15

Microstructural evolution of super-clean reduced-activation martensitic steels irradiated with single-beam (Fe{sup +}) and sequential-beam (Fe{sup +} plus He{sup +}) at 350 °C and 550 °C was studied. Sequential-beam irradiation induced smaller size and larger number density of precipitates compared to single-beam irradiation at 350 °C. The largest size of cavities was observed after sequential-beam irradiation at 550 °C. The segregation of Cr and W and depletion of Fe in carbides were observed, and the maximum depletion of Fe and enrichment of Cr occurred under irradiation at 350 °C.

Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

International Nuclear Information System (INIS)

Chakraborty, Pritam; Biner, S.Bulent

2015-01-01

Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

Damage behavior in helium-irradiated reduced-activation martensitic steels at elevated temperatures

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Chen, Jihong; Li, Tiecheng; Zheng, Zhongcheng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-12-15

Dislocation loops induced by helium irradiation at elevated temperatures in reduced-activation martensitic steels were investigated using transmission electron microscopy. Steels were irradiated with 100 keV helium ions to 0.8 dpa between 300 K and 723 K. At irradiation temperatures T{sub irr} ⩽ 573 K, small defects with both Burger vectors b = 1/2〈1 1 1〉 and b = 〈1 0 0〉 were observed, while at T{sub irr} ⩾ 623 K, the microstructure was dominated by large convoluted interstitial dislocation loops with b = 〈1 0 0〉. Only small cavities were found in the steels irradiated at 723 K.

Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

Directory of Open Access Journals (Sweden)

Noble Brooklyn

2012-01-01

Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

Science.gov (United States)

Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

2018-01-01

Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

Energy Technology Data Exchange (ETDEWEB)

Rabenberg, Ellen M.; Jaques, Brian J. [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Sencer, Bulent H. [Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Garner, Frank A. [Radiation Effects Consulting, 2003 Howell Ave., Richland, WA 99354 (United States); Freyer, Paula D. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Okita, Taira [Research Into Artifacts Dept., Center for Engineering, University of Tokyo, Tokyo (Japan); Butt, Darryl P., E-mail: DarrylButt@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States)

2014-05-01

The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. An effective tensile strain hardening exponent was also obtained from the data which shows a relative decrease in ductility of steel with increased irradiation damage. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

Effects of 14 MeV neutron irradiation on creep of nickel and niobium

International Nuclear Information System (INIS)

Barmore, W.; Ruotola, A.; Raymond, E.; Mukherjee, A.

1983-01-01

Flux, stress and temperature effects on the creep strength of nickel and niobium were observed in situ at the RTNS-II 14 MeV neutron source at Lawrence Livermore National Laboratory. Creep test were done on Ni and Nb near 0.3 Tsub(m) with stresses to 280 MPa in a high vacuum test unit using a digital computer for control and data acquisition. Cyclic flux tests produced dramatic changes in creep rate. This creep behavior is attributed to the point defect fluctuations in the crystal structure. Analysis of creep and stress relaxation under steady state flux indicates that an intermediate temperature, thermally activated deformation mechanism is rate controlling. (orig.)

Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

International Nuclear Information System (INIS)

Jones, R.H.; Gelles, D.S.

1993-08-01

Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry

Microstructural probing of ferritic/martensitic steels using internal transmutation-based positron source

Energy Technology Data Exchange (ETDEWEB)

Krsjak, Vladimir, E-mail: vladimir.krsjak@psi.ch; Dai, Yong

2015-10-15

This paper presents the use of an internal {sup 44}Ti/{sup 44}Sc radioisotope source for a direct microstructural characterization of ferritic/martensitic (f/m) steels after irradiation in targets of spallation neutron sources. Gamma spectroscopy measurements show a production of ∼1MBq of {sup 44}Ti per 1 g of f/m steels irradiated at 1 dpa (displaced per atom) in the mixed proton–neutron spectrum at the Swiss spallation neutron source (SINQ). In the decay chain {sup 44}Ti → {sup 44}Sc → {sup 44}Ca, positrons are produced together with prompt gamma rays which enable the application of different positron annihilation spectroscopy (PAS) analyses, including lifetime and Doppler broadening spectroscopy. Due to the high production yield, long half-life and relatively high energy of positrons of {sup 44}Ti, this methodology opens up new potential for simple, effective and inexpensive characterization of radiation induced defects in f/m steels irradiated in a spallation target.

Microstructural probing of ferritic/martensitic steels using internal transmutation-based positron source

Science.gov (United States)

Krsjak, Vladimir; Dai, Yong

2015-10-01

This paper presents the use of an internal 44Ti/44Sc radioisotope source for a direct microstructural characterization of ferritic/martensitic (f/m) steels after irradiation in targets of spallation neutron sources. Gamma spectroscopy measurements show a production of ∼1MBq of 44Ti per 1 g of f/m steels irradiated at 1 dpa (displaced per atom) in the mixed proton-neutron spectrum at the Swiss spallation neutron source (SINQ). In the decay chain 44Ti → 44Sc → 44Ca, positrons are produced together with prompt gamma rays which enable the application of different positron annihilation spectroscopy (PAS) analyses, including lifetime and Doppler broadening spectroscopy. Due to the high production yield, long half-life and relatively high energy of positrons of 44Ti, this methodology opens up new potential for simple, effective and inexpensive characterization of radiation induced defects in f/m steels irradiated in a spallation target.

Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

Energy Technology Data Exchange (ETDEWEB)

Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

1996-10-01

Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600{degrees}C. Each of the four heats exhibited a different stress-free swelling behavior at 400{degrees}C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower ({approx}50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400{degrees}C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600{degrees}C.

Design of a single variable helium effects experiment for irradiation in FFTF [Fast Flux Test Facility] using alloys enriched in nickel 59

International Nuclear Information System (INIS)

Simons, R.L.; Brager, H.R.; Matsumoto, W.Y.

1986-03-01

Nickel enriched in nickel 59 was extracted from the fragments of a fracture toughness specimen of Inconel 600 irradiated in the Engineering Test Reactor (ETR). The nickel contained 2.0% nickel 59. Three heats of austenitic steel doped with nickel-59 were prepared and inserted in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The experiment was single variable in helium effects because chemically identical alloys without nickel-59 were being irradiated side by side with the doped material. The alloys doped with nickel 59 produced 10 to 100 times more helium than the control alloys. The materials included ternary and quaternary alloys in the form of transmission electron microscope (TEM) discs and miniature tensile specimens. The helium to dpa ratio was in the range 5 to 35 and was nearly constant throughout the irradiation. The exposures ranged from 0.25 to 50 displacements per atom (dpa) over the duration of the experiment. The irradiation temperatures covered the range of 360 to 600 0 C

Material science and manufacturing of heat-resistant reduced-activation ferritic-martensitic steels for fusion

International Nuclear Information System (INIS)

Ioltukhovskiy, A.G.; Blokhin, A.I.; Budylkin, N.I.; Chernov, V.M.; Leont'eva-Smirnova, M.V.; Mironova, E.G.; Medvedeva, E.A.; Solonin, M.I.; Porollo, S.I.; Zavyalsky, L.P.

2000-01-01

A number of issues regarding the development and use of 10-12% Cr reduced-activation ferritic-martensitic steels (RAFMS) for fusion are considered. These include: (1) problems of manufacturing and modifying their composition and metallurgical condition; (2) the influence on properties of their composition, purity, δ-ferrite concentration and cooling rates in the final stages of manufacturing; and (3) the effects of neutron irradiation at 320-650 deg. C up to 108 dpa on their mechanical properties. In addition, neutron activation and nuclear accumulation of elements in RAFMS with different initial concentrations of alloying and impurity elements for typical fusion reactor (DEMO) irradiation regimes have been calculated

Neutron irradiation effects of iron alloys and ceramics

International Nuclear Information System (INIS)

Kuramoto, Eiichi; Takenaka, Minoru; Hasegawa, Masayuki.

1991-01-01

Positron annihilation angular correlation measurements have been performed for the neutron irradiated various metals and ceramics in order to obtain the information of the microvoids and positronium formation in them. Positronium (Ps) formation was observed in Nb containing a small amount of oxygen and Fe-15%Cr-16%Ni-0.006%B 10 . In practical steels such as JPCA and JFMS no Ps formation was observed. High temperature deformation might induce microvoids into metals, but the positron annihilation angular correlation measurements could not confirm this. In non-metallic materials neutron irradiated no Ps formation has so far been observed. (author)

Peculiarities of neutron interaction with boron containing semiconductors

International Nuclear Information System (INIS)

Didyk, A.Yu.; ); Hofman, A.; Institute of Atomic Energy, Otwock/Swierk; Vlasukova, L.A.

2009-01-01

The results of point defect creation calculation in B 4 C, BN and BP semiconductor single crystals irradiated in the fast neutron reactor IBR-2 are presented. It has been shown that during the thermal neutron interaction with light isotope boron atoms ( 10 B) the damage creation by means of fission nuclear reaction fragments (alpha particles and 7 Li recoil nuclei) exceeds the damage created by fast neutrons (E n > 0.1 MeV) by more than two orders of value. It has been concluded that such irradiation can create a well developed radiation defect structure in boron-containing crystals with nearly homogeneous vacancy depth distribution. This may be used in technological applications for more effective diffusion of impurities implanted at low energies or deposited onto the semiconductor surface. The developed homogeneous vacancy structure is very suitable for the radiation enhanced diffusion of electrically charged or neutral impurities from the surface into the technological depth of semiconductor devices under post irradiation treatment. (authors)

Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

Directory of Open Access Journals (Sweden)

Yang Seong Woo

2016-01-01

Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

Soft-martensitic stainless Cr-Ni-Mo steel for turbine rotors in geothermic power stations

International Nuclear Information System (INIS)

Schonfeld, K.; Potthast, E.

1986-01-01

Steel Grade X5 Cr-Ni-Mo 12 6 containing 0.05% carbon, 12% chromium, 6% nickel, and 1.50% molybdenum is an advantageous material for turbine rotors in geothermic power stations because of its excellent strength and toughness properties in combination with good erosion and corrosion resistance. In terms of the phase diagram, this soft-martensitic steel has its place at the martensite/austenite/ferrite interface. Therefore, its chemical composition must be chosen so as to have a completely martensitic structure after hardening. The manufacture of and the mechanical properties of a turbine rotor 1200 mm in diameter by 5600 mm in length with a finished weight of approximately 21.5 tons are described in detail

Radiation effects on low cycle fatigue properties of reduced activation ferritic/martensitic steels

International Nuclear Information System (INIS)

Hirose, T.; Tanigawa, H.; Ando, M.; Kohyama, A.; Katoh, Y.; Narui, M.

2002-01-01

The reduced activation ferritic/martensitic steel, RAFs F82H IEA heat has been fatigue-tested at ambient temperature under diametral strain controlled conditions. In order to evaluate the effects of radiation damage and transmutation damage on fatigue characteristics, post-neutron irradiation and post-helium ion implantation fatigue tests were carried out. Fracture surfaces and fatigue crack initiation on the specimen surface were observed by SEM. Low-temperature irradiation caused an increase in stress amplitude and a reduction in fatigue lifetime corresponding to radiation hardening and loss of ductility. Neutron irradiated samples showed brittle fracture surface, and it was significant for large strain tests. On the other hand, helium implantation caused delay of cyclic softening. However, brittle crack initiation and propagation did not depend on the helium concentration profiles

Post-irradiation characterization of PH13-8Mo martensitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Jong, M.; Schmalz, F.; Rensman, J.W. [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands); Luzginova, N.V., E-mail: luzginova@nrg.eu [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands); Wouters, O.; Hegeman, J.B.J.; Laan, J.G. van der [Nuclear Research and consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

2011-10-01

The irradiation response of PH13-8Mo stainless steel was measured up to 2.5 dpa at 200 and 300 deg. C irradiation temperatures. The PH13-8Mo, a martensitic precipitation-hardened steel, was produced by Hot Isostatic Pressing at 1030 deg. C. The fatigue tests (high cycle fatigue and fatigue crack propagation) showed a test temperature dependency but no irradiation effects. Tensile tests showed irradiation hardening (yield stress increase) of approximately 37% for 200 deg. C irradiated material tested at 60 deg. C and approximately 32% for 300 deg. C irradiated material tested at 60 deg. C. This contradicts the shift in reference temperature (T{sub 0}) measured in toughness tests (Master Curve approach), where the {Delta}T{sub 0} for 300 deg. C irradiated is approximately 170 deg. C and the {Delta}T{sub 0} for the 200 deg. C irradiated is approximately 160 deg. C. This means that the irradiation hardening of PH13-8Mo steel is not suitable to predict the shift in the reference temperature for the Master Curve approach.

Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

International Nuclear Information System (INIS)

Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang

2012-01-01

The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated

Review of recent studies on neutron irradiation embrittlement in light water reactor pressure vessel steels

International Nuclear Information System (INIS)

Sudo, Akira; Miyazono, Shohachiro

1983-06-01

Recent studies in foreign countries (USA, France, FRG and UK) on neutron irradiation embrittlement have been reviewed. These studies are classified into four areas, such as 1) effect of chemical composition on irradiation embrittlement sensitivity, 2) postirradiation heat treatment for embrittlement relief, 3) fracture toughness evaluation of irradiated materials based on fracture mechanics analysis, and 4) effect of irradiation on fatigue crack propagation behavior. The first area mainly includes the studies related to the effects of copper, phosphorus impurities and nickel alloying and synergistic effect of these components, and furthermore, evaluation of Regulatory Guide 1.99 Rev.l. Studies in the second area show the effects of annealing condition (temperature and time) and metallugical condition on embrittlement relief, and evaluation of periodic annealing in the period of irradiation as a promising method for embrittlement control. Studies in the third area show the correlation between fracture toughness and Cv notch ductility changes with neutron irradiation, and J-R curves of irradiated materials based on the elasto-plastic fracture mechanics. In the forth area, most of studies are investigated in air condition but a few studies in reactor-grade water at high temperature and pressure. (author)

Enhanced electromagnetic properties of nickel nanoparticiles dispersed carbon fiber via electron beam irradiation

International Nuclear Information System (INIS)

Lee, Yeong Ju; Kim, Hyun Bin; Lee, Seung Jun; Kang, Phil Hyun

2015-01-01

Carbon fiber has received much attention owing to its properties, including a large surface-to-volume ratio, chemical and thermal stability, high thermal and electrical conductivity, and high mechanical strengths. In particular, magnetic nanopowder dispersed carbon fiber has been attractive in technological applications such as the electrochemical capacitor and electromagnetic wave shielding. In this study, the nickel-oxide-nanoparticle dispersed polyacrylonitrile (PAN) fibers were prepared through an electrospinning method. Electron beam irradiation was carried out with a 2.5 MeV beam energy to stabilize the materials. The samples were then heat treated for stabilization and carbonization. The nanofiber surface was analyzed using a field emission scanning electron microscope (FE-SEM). The crystal structures of the carbon matrix and nickel nanopowders were analysed using X-ray diffraction (XRD). In addition, the magnetic and electrical properties were analyzed using a vibrating sample magnetometer (VSM) and 4 point probe. As the irradiation dose increases, the density of the carbon fiber was increased. In addition, the electrical properties of the carbon fiber improved through electron beam irradiation. This is because the amorphous region of the carbon fiber decreases. This electron beam effect of PAN fibers containing nickel nanoparticles confirmed their potential as a high performance carbon material for various applications

Enhanced electromagnetic properties of nickel nanoparticiles dispersed carbon fiber via electron beam irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Yeong Ju; Kim, Hyun Bin; Lee, Seung Jun; Kang, Phil Hyun [Korea Atomic Energy Research Institute, Jeongeup (Korea, Republic of)

2015-02-15

Carbon fiber has received much attention owing to its properties, including a large surface-to-volume ratio, chemical and thermal stability, high thermal and electrical conductivity, and high mechanical strengths. In particular, magnetic nanopowder dispersed carbon fiber has been attractive in technological applications such as the electrochemical capacitor and electromagnetic wave shielding. In this study, the nickel-oxide-nanoparticle dispersed polyacrylonitrile (PAN) fibers were prepared through an electrospinning method. Electron beam irradiation was carried out with a 2.5 MeV beam energy to stabilize the materials. The samples were then heat treated for stabilization and carbonization. The nanofiber surface was analyzed using a field emission scanning electron microscope (FE-SEM). The crystal structures of the carbon matrix and nickel nanopowders were analysed using X-ray diffraction (XRD). In addition, the magnetic and electrical properties were analyzed using a vibrating sample magnetometer (VSM) and 4 point probe. As the irradiation dose increases, the density of the carbon fiber was increased. In addition, the electrical properties of the carbon fiber improved through electron beam irradiation. This is because the amorphous region of the carbon fiber decreases. This electron beam effect of PAN fibers containing nickel nanoparticles confirmed their potential as a high performance carbon material for various applications.

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

Science.gov (United States)

Gelles, D. S.

1990-05-01

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

Neutron irradiation effects on plasma facing materials

Science.gov (United States)

Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

2000-12-01

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

Neutron irradiation effects on plasma facing materials

International Nuclear Information System (INIS)

Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

2000-01-01

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

Neutron irradiation of sapphire for compressive strengthening. II. Physical properties changes

Energy Technology Data Exchange (ETDEWEB)

Regan, Thomas M. E-mail: thomas_regan@uml.edu; Harris, Daniel C. E-mail: harrisdc@navair.navy.mil; Blodgett, David W.; Baldwin, Kevin C.; Miragliotta, Joseph A.; Thomas, Michael E.; Linevsky, Milton J.; Giles, John W.; Kennedy, Thomas A.; Fatemi, Mohammad; Black, David R.; Lagerloef, K. Peter D

2002-01-01

Irradiation of sapphire with fast neutrons (0.8-10 MeV) at a fluence of 10{sup 22}/m{sup 2} increased the c-axis compressive strength and the c-plane biaxial flexure strength at 600 deg. C by a factor of {approx}2.5. Both effects are attributed to inhibition of r-plane twin propagation by damage clusters resulting from neutron impact. The a-plane biaxial flexure strength and four-point flexure strength in the c- and m-directions decreased by 10-23% at 600 deg. C after neutron irradiation. Neutron irradiation had little or no effect on thermal conductivity, infrared absorption, elastic constants, hardness, and fracture toughness. A featureless electron paramagnetic resonance signal at g=2.02 was correlated with the strength increase: This signal grew in amplitude with increasing neutron irradiation, which also increased the compressive strength. Annealing conditions that reversed the strengthening also annihilated the g=2.02 signal. A signal associated with a paramagnetic center containing two Al nuclei was not correlated with strength. Ultraviolet and visible color centers also were not correlated with strength in that they could be removed by annealing at temperatures that were too low to reverse the compressive strengthening effect of neutron irradiation.

Thermal neutron capture cross section of chromium, vanadium, titanium and nickel isotopes

International Nuclear Information System (INIS)

Venturini, L.; Pecequilo, B.R.S.

1990-04-01

The thermal neutron cross section of chromium, vanadium, titanium and nickel can be determined by measuring the pair spectrum of prompt gamma-rays emitted targets of these elements are irradiated by a thermal neutron beam. Such measurements were carried out by irradiating the natural element mixed with a nitrogen standard (melamine) in the tangential beam hole of the IEA-R1 research reactor. The pair spectrometer efficiency calibration curve in the 1.5 to 11 MeV energy range was performed with a melamine plus ammonium chloride mixed target. The cross section was calculated for the most prominent gamma transitions of each isotope, using nitrogen as standard and averaged over the obtained values. The resulting mean cross sections are as follows: (13.4 ± 0.7)b for 50 Cr, (0.79 ± 0,02)b for 52 Cr, (18.1 ± 0,7)b for 53 Cr, (4.9 ± 0.2)b for 51 V, (8.4 ± 0.1)b for 48 Ti, (4.41 ± 0.08)b 58 Ni, (2.54 ± 0.07)b for 60 Ni, (15.2 ± 0.5)b for 62 Ni and (1.6 ± 0.1) for 64 Ni. (author) [pt

Neutron irradiation damage in transition metal carbides

International Nuclear Information System (INIS)

Matsui, Hisayuki; Nesaki, Kouji; Kiritani, Michio

1991-01-01

Effects of neutron irradiation on the physical properties of light transition metal carbides, TiC x , VC x and NbC x , were examined, emphasizing the characterization of irradiation induced defects in the nonstoichiometric composition. TiC x irradiated with 14 MeV (fusion) neutrons showed higher damage rates with increasing C/Ti (x) ratio. A brief discussion is made on 'cascade damage' in TiC x irradiated with fusion neutrons. Two other carbides (VC x and NbC x ) were irradiated with fission reactor neutrons. The irradiation effects on VC x were not so simple, because of the complex irradiation behavior of 'ordered' phases. For instance, complete disordering was revealed in an ordered phase, 'V 8 C 7 ', after an irradiation dose of 10 25 n/m 2 . (orig.)

Influence of PWHT on Toughness of High Chromium and Nickel Containing Martensitic Stainless Steel Weld Metals

Science.gov (United States)

Divya, M.; Das, Chitta Ranjan; Mahadevan, S.; Albert, S. K.; Pandian, R.; Kar, Sujoy Kumar; Bhaduri, A. K.; Jayakumar, T.

2015-06-01

Commonly used 12.5Cr-5Ni consumable specified for welding of martensitic stainless steels is compared with newly designed 14.5Cr-5Ni consumable in terms of their suitability for repair welding of 410 and 414 stainless steels by gas tungsten arc welding process. Changes in microstructure and austenite evolution were investigated using optical, scanning electron microscopy, X-ray diffraction techniques and Thermo-Calc studies. Microstructure of as-welded 12.5Cr-5Ni weld metal revealed only lath martensite, whereas as-welded 14.5Cr-5Ni weld metal revealed delta-ferrite, retained austenite, and lath martensite. Toughness value of as-welded 12.5Cr-5Ni weld metal is found to be significantly higher (216 J) than that of the 14.5Cr-5Ni weld metal (15 J). The welds were subjected to different PWHTs: one at 923 K (650 °C) for 1, 2, 4 hours (single-stage PWHT) and another one at 923 K (650 °C)/4 h followed by 873 K (600 °C)/2 h or 873 K (600 °C)/4 h (two-stage heat treatment). Hardness and impact toughness of the weld metals were measured for these weld metals and correlated with the microstructure. The study demonstrates the importance of avoiding formation of delta-ferrite in the weld metal.

The effect of microstructural change on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS) irradiated in JOYO/MARICO-1

International Nuclear Information System (INIS)

Yano, Yasuhide; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro

2004-03-01

It is well known that the irradiation embrittlement is one of the most important issues to apply ferritic steels for FBR core materials, although ferritic steels have been considered to be candidate core materials of the commercialized FBR core material because of their superior swelling resistance. In order to evaluate the effects of microstructural changes during irradiation on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS), microstructural observations were performed with transmission electron microscopy on ruptured halves of the half-sized Charpy specimens of PNC-FMS irradiated in the JOYO/MARICO-1. The results obtained in this study are as follows: (1) There was remarkable disappearance of the lath of martensite in the samples irradiated at 650degC, although there was no significant change in microstructures, especially the lath of martensite between the samples irradiated at 500degC and unirradiated. The disappearance of martensitic lath in the samples irradiated at 650degC was larger than that of the samples thermally aged at 650degC. (2) The ductile-brittle transition temperature (DBTT) of irradiated PNC-FMS is judged to increase with the disappearance of martensitic lath and to decrease with the recovery in dislocations. (3) The decrease in the upper shelf energy (USE) of irradiated PNC-FMS is significantly accompanied by the change of precipitation behavior. (4) The Charpy impact properties and microstructures of PNC-FMS irradiated at 500degC were superior under these irradiation conditions. In future, it is necessary to establish how to evaluate Charpy impact properties in a high fluence region, based on theoretical methods introduced from the data gained in low fluence experiments, in addition to expanding the data area widely. (author)

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

English, C.A.; Phythian, W.J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

New facility for post irradiation examination of neutron irradiated beryllium

International Nuclear Information System (INIS)

Ishitsuka, Etsuo; Kawamura, Hiroshi

1995-01-01

Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

Saturation volume changes and resistivity changes in nickel

International Nuclear Information System (INIS)

Birtcher, R.C.; Blewitt, T.H.

1976-01-01

Saturation defect concentrations generated by thermal neutron irradiation of 235 U doped nickel at liquid helium temperature were measured by changes in electrical resistivity and volume. The experimental procedure is described

α′ precipitation in neutron-irradiated Fe–Cr alloys

International Nuclear Information System (INIS)

Bachhav, Mukesh; Robert Odette, G.; Marquis, Emmanuelle A.

2014-01-01

Graphical abstract: -- A series of model Fe–Cr alloys containing 3–18 at.% Cr was neutron irradiated at a nominal temperature of 563 K to 1.82 dpa. Solute distributions were analyzed by atom probe tomography, which revealed α′ precipitation for alloys containing more than 9 at.% Cr. Both the Cr concentration dependence of α′ precipitation and the measured matrix compositions are in agreement with the recently published Fe–Cr phase diagrams. An irradiation-accelerated precipitation process is strongly suggested

Comparison of deuterium retention for ion-irradiated and neutron-irradiated tungsten

International Nuclear Information System (INIS)

Oya, Yasuhisa; Kobayashi, Makoto; Okuno, Kenji; Shimada, Masashi; Calderoni, Pattrick; Oda, Takuji; Hara, Masanori; Hatano, Yuji; Watanabe, Hideo

2014-01-01

The behavior of D retentions for Fe 2+ irradiated tungsten with the damage of 0.025-3 dpa was compared with that for neutron irradiated tungsten with 0.025 dpa. The D 2 TDS spectra for Fe 2+ irradiated tungsten consisted of two desorption stages at 450 K and 550 K although that for neutron irradiated tungsten was composed of three stages and addition desorption stage was found around 750 K. The desorption rate of major desorption stage at 550 K increased as the number of dpa by Fe 2+ irradiation increased. In addition, the first desorption stage at 450 K was only found for the damaged samples, indicating that the second stage would be based on intrinsic defects or vacancy produced by Fe 2+ irradiation and the first stage should be the accumulation of D in mono vacancy leading to the lower activation energy, where the dislocation loop and vacancy was produced. The third one was only found for the neutron irradiation, showing the D trapping by void or vacancy cluster and the diffusion effect is also contributed due to high FWHM of TDS spectrum. It can be said that the D 2 TDS spectra for Fe 2+ -irradiated tungsten could not represent that for neutron-irradiated one, showing that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten has a difference from that for ion-irradiated one. (author)

Study on neutron irradiation behavior of beryllium as neutron multiplier

Energy Technology Data Exchange (ETDEWEB)

Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-03-01

More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

Charpy impact behavior of manganese-stabilized martensitic steels

International Nuclear Information System (INIS)

Hu, W.L.; Gelles, D.S.

1986-05-01

Tests were conducted to evaluate the irradiation-induced shift in ductile-to-brittle transition behavior of two manganese stabilized martensitic steels. Miniature Charpy specimens were fabricated from two heats of steel similar in composition to HT-9 but with 0.1% C and Mn contents ranging from 3.3 to 6.6.%. The 3.3% Mn steel showed a transition temperature similar to that of HT-9 in both the unirradiated condition and in specimens irradiated to 11.3 dpa. The steel containing 6.6% Mn exhibited a higher transition temperature after irradiation than the steel containing 3.3% Mn. The upper shelf energy (USE) after irradiation for the manganese stabilized alloys was much higher than for HT-9. 6 refs., 3 figs., 2 tabs

Severe Embrittlement of Neutron Irradiated Austenitic Steels Arising from High Void Swelling

International Nuclear Information System (INIS)

Neustroev, V.S.; Garner, F.

2007-01-01

Full text of publication follows: Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components. Similar loss of ductility is expected when swelling arises in fusion and light water reactor environments. Above 7-16% swelling there is complete loss of ductility, with the onset of ductility loss beginning at lower swelling in ring-pull tensile tests than for flat tensile specimens. For steels that develop extensive precipitation during irradiation, the critical swelling level is even lower. A model is presented to demonstrate the effect of voids acting alone to produce the embrittlement. Although voids are not very effective hardeners, they are very effective to generate stress concentrations between voids. The stress concentration ratio increases strongly when the void diameter exceeds ∼40% of the void-to-void separation distance. When the volume fraction of voids is rather high (about 16 % and higher), a geometric situation develops where it is possible to create an intense field of deformation glide planes residing at an angle of 45 deg. to the void-to-void axis. Significant localized flow then proceeds on these planes for specimen stress levels that are significantly lower than the yield stress. Voids also segregate nickel to their surfaces such that flow localization occurs in the low-nickel inter-void regions to produce strain-induced martensite, which is further accelerated by stress concentrations at the advancing crack tip, leading to catastrophic failure. (authors)

In situ neutron diffraction studies of martensitic transformations in NiTi polycrystals under tension and compression stress

Czech Academy of Sciences Publication Activity Database

Šittner, Petr; Lukáš, Petr; Novák, Václav; Daymond, M. R.; Swallowe, G. M.

2004-01-01

Roč. 378, - (2004), s. 97-104 ISSN 0921-5093 Institutional research plan: CEZ:AV0Z1010914 Keywords : martensitic transformation * shape memory alloy * neutron diffraction * NiTi Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.445, year: 2004

Comparative Study of Hardening Mechanisms During Aging of a 304 Stainless Steel Containing α'-Martensite

Science.gov (United States)

Jeong, S. W.; Kang, U. G.; Choi, J. Y.; Nam, W. J.

2012-09-01

Strain aging and hardening behaviors of a 304 stainless steel containing deformation-induced martensite were investigated by examining mechanical properties and microstructural evolution for different aging temperature and time. Introduced age hardening mechanisms of a cold rolled 304 stainless steel were the additional formation of α'-martensite, hardening of α'-martensite, and hardening of deformed austenite. The increased amount of α'-martensite at an aging temperature of 450 °C confirmed the additional formation of α'-martensite as a hardening mechanism in a cold rolled 304 stainless steel. Additionally, the increased hardness in both α'-martensite and austenite phases with aging temperature proved that hardening of both α'-martensite and austenite phases would be effective as hardening mechanisms in cold rolled and aged 304 stainless steels. The results suggested that among hardening mechanisms, hardening of an α'-martensite phase, including the diffusion of interstitial solute carbon atoms to dislocations and the precipitation of fine carbide particles would become a major hardening mechanism during aging of cold rolled 304 stainless steels.

Influence of neutron irradiation on the tritium retention in beryllium

Energy Technology Data Exchange (ETDEWEB)

Rolli, R.; Ruebel, S.; Werle, H. [Forschungszentrum Karlsruhe, Inst. fuer Neutronenphysik und Reaktortechnik, Karlsruhe (Germany); Wu, C.H.

1998-01-01

Carbon-based materials and beryllium are the candidates for protective layers on the components of fusion reactors facing plasma. In contact with D-T plasma, these materials absorb tritium, and it is anticipated that tritium retention increases with the neutron damage due to neutron-induced traps. Because of the poor data base for beryllium, the work was concentrated on it. Tritium was loaded into the samples from stagnant T{sub 2}/H{sub 2} atmosphere, and afterwards, the quantity of the loaded tritium was determined by purged thermal annealing. The specification of the samples is shown. The samples were analyzed by SEM before and after irradiation. The loading and the annealing equipments are contained in two different glove boxes with N{sub 2} inert atmosphere. The methods of loading and annealing are explained. The separation of neutron-produced and loaded tritium and the determination of loaded tritium in irradiated samples are reported. Also the determination of loaded tritium in unirradiated samples is reported. It is evident that irradiated samples contained much more loaded tritium than unirradiated samples. The main results of this investigation are summarized in the table. (K.I.)

Hyper-thermal neutron irradiation field for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1994-01-01

The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

Polarized Neutron Study of Ni-Mn-Ga Alloys: Site-Specific Spin Density Affected by Martensitic Transformation.

Science.gov (United States)

Lázpita, P; Barandiarán, J M; Gutiérrez, J; Mondelli, C; Sozinov, A; Chernenko, V A

2017-10-13

Polarized neutron scattering has been used to obtain the magnetic moment at specific crystallographic sites of the austenitic and martensitic phases of two nonstoichiometric Ni-Mn-Ga single crystals with close composition. These alloys have been chosen because they exhibit different structures in the paramagnetic state and inverse positions of the respective martensitic transformation and Curie temperature. The diffraction analysis revealed a remarkable result: Despite the similar alloy composition, the magnetic moments of Mn are quite different for the two alloys at the same crystallographic position. Furthermore, such a difference enabled us to assess that the exchange coupling between Mn atoms switches from ferro- to antiferromagnetic at a distance between 2.92 and 3.32 Å in the martensite. These results are of great importance to guide first principles calculations that, up to now, have not been contrasted with experiments at the atomic level.

Quantitative analysis of tensile deformation behavior by in-situ neutron diffraction for ferrite-martensite type dual-phase steels

International Nuclear Information System (INIS)

Morooka, Satoshi; Umezawa, Osamu; Harjo, Stefanus; Hasegawa, Kohei; Toji, Yuki

2012-01-01

The yielding and work-hardening behavior of ferrite-martensite type dual-phase (DP) alloys were clearly analyzed using the in-situ neutron diffraction technique. We successfully established a new method to estimate the stress and strain partitioning between ferrite and martensite phase during loading. Although these phases exhibit the same lattice structure with similar lattice parameters, their lattice strains on (110), (200) and (211) are obviously different from each other under an applied stress. The misfit strains between those phases were clearly accompanied with the phase-scaled internal stream (phase stress). Thus, the martensite phase yielded by higher applied stress than macro-yield stress, which resulted in high work-hardening rate of the DP steel. We also demonstrated that ferrite phase fraction influenced work-hardening behavior. (author)

Applications of neutron irradiation

International Nuclear Information System (INIS)

Ito, Yasuo

1999-01-01

The present state of art of applications of neutron irradiation is overviewed taking neutron activation analysis, prompt gamma-ray analysis, fission/alpha track methods, boron neutron capture therapy as examples. What is common among them is that the technologies are nearly matured for wide use by non- nuclear scientists. But the environment around research reactors is not prospective. These applications should be encouraged by incorporating in the neutron science society. (author)

Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

International Nuclear Information System (INIS)

Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

1978-05-01

Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case

Evaluation of copper, aluminum bronze, and copper-nickel container material for the Yucca mountain project

International Nuclear Information System (INIS)

Kass, J.

1990-01-01

Copper, 70 percent aluminum bronze, and 70/30 copper-nickel were evaluated as potential waste-packaging materials as part of the Yucca Mountain Project. The proposed waste repository site is under a desert mountain in southern Nevada. The expected temperatures at the container surface are higher than at other sites, about 250C at the beginning of the containment period; they could fall below the boiling point of water during this period, but will be exposed to very little water, probably less than 5 l/a. Initial gamma flux will be 10 4 rad/h, and no significant hydrostatic or lithostatic pressure is expected. Packages will contain PWR or BWR fuel, or processed-glass waste. Three copper alloys are being considered for containers: oxygen-free copper (CDA 102); 7 percent aluminum bronze (CDA 613); and 70/30 copper-nickel (CDA 715). Phase separation due to prolonged thermal exposure could be a problem for the two alloys, causing embrittlement. The reduction of internal oxides present in pure copper by hydrogen could cause mechanical degradation. Corrosion and oxidation rates measured for the three materials in well water with and without gamma irradiation at flux rates about ten times higher than those expected were all quite small. The corrosion/oxidation rates for CDA715 show a marked increase under irradiation, but are still acceptable. In the presence of ammonia and other nitrogen-bearing species stress corrosion cracking (SCC) is a concern. Welded U-bend specimens of all three materials have been tested for up to 10000 h in highly irradiated environments, showing no SCC. There was some alloy segregation in the Al bronze specimens. The investigators believe that corrosion and mechanical properties will not present problems for these materials at this site. Further work is needed in the areas of weld inspection, welding techniques, embrittlement of weld metal, the effects of dropping the containers during emplacement, and stress corrosion cracking. Other materials

Polarized Neutron Reflectometry of Nickel Corrosion Inhibitors.

Science.gov (United States)

Wood, Mary H; Welbourn, Rebecca J L; Zarbakhsh, Ali; Gutfreund, Philipp; Clarke, Stuart M

2015-06-30

Polarized neutron reflectometry has been used to investigate the detailed adsorption behavior and corrosion inhibition mechanism of two surfactants on a nickel surface under acidic conditions. Both the corrosion of the nickel surface and the structure of the adsorbed surfactant layer could be monitored in situ by the use of different solvent contrasts. Layer thicknesses and roughnesses were evaluated over a range of pH values, showing distinctly the superior corrosion inhibition of one negatively charged surfactant (sodium dodecyl sulfate) compared to a positively charged example (dodecyl trimethylammonium bromide) due to its stronger binding interaction with the surface. It was found that adequate corrosion inhibition occurs at significantly less than full surface coverage.

Changes in grain boundary composition induced by neutron irradiation of austenitic stainless steels

International Nuclear Information System (INIS)

Asano, K.; Nakata, K.; Fukuya, K.; Kodama, M.

1992-01-01

The radiation induced segregation of solutes to the grain boundary in austenitic stainless steels were studied. Type 304 and type 316 steel samples neutron irradiated at 561K up to 9.2x10 25 n/m 2 were obtained and minute compositional profiles across grain boundaries were examined using an analytical scanning transmission electron microscope equipped with a field emission electron gun. Chromium was slightly enriched at grain boundaries at the lowest irradiation dose but decreased with increasing fluence. Higher fluence irradiation resulted in depletion in chromium and molybdenum, and enrichment in nickel, silicon and phosphorus. These changes in grain boundary chemistry were limited within about 5nm of the boundary. Significant depletion of chromium and enrichment of impurities on the grain boundary occurred at fluences roughly coincidental with that of SCC susceptibility change obtained from another project

Temper-bead repair-welding of neutron-irradiated reactor (pressure) vessel by low-heat-input TIG and YAG laser welding

International Nuclear Information System (INIS)

Nakata, Kiyotomo; Ozawa, Masayoshi; Kamo, Kazuhiko

2006-01-01

Weldability in neutron-irradiated low alloy steel for reactor (pressure) vessel has been studied by temper-bead repair-welding of low-heat-input TIG and YAG laser welding. A low alloy steel and its weld, and stainless steel clad and nickel (Ni)-based alloy clad were irradiated in a materials test reactor (LVR-15, Czech Republic) up to 1.4 x 10 24 n/m 2 (>1 MeV) at 290degC, which approximately corresponds to the maximum neutron fluence of 60-year-operation plants' vessels. The He concentration in the irradiated specimens was estimated to be up to 12.9 appm. The repair-welding was carried out by TIG and YAG laser welding at a heat input from 0.06 to 0.86 MJ/m. The mechanical tests of tensile, impact, side bend and hardness were carried out after the repair-welding. Cracks were not observed in the irradiated low alloy steel and its weld by temper-bead repair-welding. Small porosities were formed in the first and second layers of the repair-welds of low alloy steel (base metal). However, only a few porosities were found in the repair-welds of the weld of low alloy steel. From the results of mechanical tests, the repair-welding could be done in the irradiated weld of low alloy steel containing a He concentration up to 12.9 appm, although repair-welding could be done in base metal of low alloy steel containing up to only 1.7 appmHe. On the other hand, cracks occurred in the heat affected zones of stainless steel and Ni-based alloy clads by repair-welding, except by YAG laser repair-welding at a heat input of 0.06 MJ/m in stainless steel clad containing 1.7 appmHe. Based on these results, the determination processes were proposed for optimum parameters of repair-welding of low alloy steel and clad used for reactor (pressure) vessel. (author)

Alterations in water and electrolyte absorption in the rat colon following neutron irradiation: influence of neutron component and irradiation dose.

Science.gov (United States)

Dublineau, I; Ksas, B; Joubert, C; Aigueperse, J; Gourmelon, P; Griffiths, N M

2002-12-01

To study the absorptive function of rat colon following whole-body exposure to neutron irradiation, either to the same total dose with varying proportion of neutrons or to the same neutron proportion with an increasing irradiation dose. Different proportions of neutron irradiation were produced from the reactor SILENE using a fissile solution of uranium nitrate (8, 47 and 87% neutron). Water and electrolyte fluxes were measured in the rat in vivo under anaesthesia by insertion into the descending colon of an agarose gel cylinder simulating the faeces. Functional studies were completed by histological analyses. In the first set of experiments, rats received 3.8 Gy with various neutron percentages and were studied from 1 to 14 days after exposure. In the second set of experiments, rats were exposed to increasing doses of irradiation (1-4Gy) with a high neutron percentage (87%n) and were studied at 4 days after exposure. The absorptive capacity of rat colon was diminished by irradiation at 3-5 days, with a nadir at 4 days. The results demonstrate that an increase in the neutron proportion is associated with an amplification of the effects. Furthermore, a delay in the re-establishment of normal absorption was observed with the high neutron proportion (87%n). A dose-dependent reduction of water absorption by rat colon was also observed following neutron irradiation (87%n), with a 50% reduction at 3 Gy. Comparison of this dose-effect curve with the curve obtained following gamma (60)Co-irradiation indicates an RBE of 2.2 for absorptive colonic function in rat calculated at 4 days after exposure.

Progress of research on the mechanism of microstructure evolution during irradiation with collision cascades

International Nuclear Information System (INIS)

Kiritani, Michio

1997-01-01

Progress of the research on microstructural evolution during radiation damage accompanied with collision cascades, particularly by neutrons, is summarized from the researches reported in 1995 and 1996. The major part of irradiation was performed with JMTR (Japan Material Testing Reactor), and comparisons were made with the result of D-T fusion neutron irradiation with RTNS-II (Rotating Target Neutron Source, LLNL). The subjects concerned are: (1) Microstructure evolution during irradiation, (2) Processes controlling the accumulation of defects during cascade damage, (3) Subcascades from the viewpoint of point defect cluster formation, (4) Mechanism of suppression of microstructure evolution during temperature cycle neutron irradiation of nickel, (5) Identification of the nature of neutron-irradiation-induced point defect clusters in copper by means of electron irradiation, (6) Easy one-dimensional motion of small interstitial clusters, (7) Identification of the nature of small point defect clusters in neutron irradiated Fe-16Ni-15Cr, (8) Spatial distribution of nucleation of point defect clusters by high energy particle irradiation, (9) Crystallographic orientation dependence of dislocation structures in neutron irradiated metals, (10) Examination of defect accumulation mode with multi-section-removable-irradiation-rig in Japan Material Testing Reactor, (11) Estimation of freely migrating point defects during cascade damage by the growth of helical dislocations, (12) Identification of the nature of neutron-irradiation-induced small point defect clusters in Nickel by means of electron irradiation, (13) Thermal stability of point defect clusters in neutron irradiated Fe-16Ni-15Cr. (author)

Tensile behavior of EUROFER ODS steel after neutron irradiation up to 16.3 dpa between 250 and 450 °C

International Nuclear Information System (INIS)

Materna-Morris, Edeltraud; Lindau, Rainer; Schneider, Hans-Christian; Möslang, Anton

2015-01-01

Highlights: • The first 9%CrWVTa steel (0.5% Y_2O_3), EUROFER ODS HIP, have been neutron irradiated up to 16.3 dpa, between 250 and 450 °C, in the High Flux Reactor (HFR). • After post-irradiation tensile tests, there was not any increase of the upper yield strength or strain localization after irradiation which is typical of RAFM steels. • Initially higher yield strength, R_p_0_._2, and distinctive tensile strength, R_m, of EUROFER ODS HIP compared to EUROFER97 steel. • These values increased due to the neutron irradiation at lower irradiation temperatures. - Abstract: During the development of structural material for future fusion reactors, a 50 kg heat of reduced-activation ferritic-martensitic 9%CrWVTa steel with nanoscaled Y_2O_3-particles, EUROFER97 ODS HIP, was produced using powder metallurgy fabrication technology. This first batch of EUROFER97 ODS HIP and, for comparison, the steel EUROFER97 were prepared for a post-irradiation tensile test program. During neutron irradiation in the HFR (High Flux Reactor, The Netherlands), an accumulated dose of up to 16.3 dpa was reached for 771 days at full power, with the irradiation temperature ranging between 250 and 450 °C. During the post-examinations, all specimens showed the highest tensile strength at lower irradiation temperatures between 250 and 350 °C. However, ODS-alloy and steel were found to clearly differ in the mechanical behavior, which could be documented by fully instrumented tensile tests. In the un-irradiated state, tensile strength of the ODS-alloy already was increased considerably by about 60% compared to the steel. Strengthening was further increased by another 20% after neutron irradiation, but with a much better ductility than observed in the steel. The typical irradiation-induced strain localization of EUROFER97 or RAFM steels could not be observed in the EUROFER97 ODS HIP alloy.

Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

International Nuclear Information System (INIS)

1998-10-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

Ferritic/martensitic steels: Promises and problems

International Nuclear Information System (INIS)

Klueh, R.L.; Ehrlich, K.; Abe, F.

1992-01-01

Ferritic/martensitic steels are candidate structural materials for fusion reactors because of their higher swelling resistance, higher thermal conductivity, lower thermal expansion, and better liquid-metal compatibility than austenitic steels. Irradiation effects will ultimately determine the applicability of these steels, and the effects of irradiation on microstructure and swelling, and on the tensile, fatigue, and impact properties of the ferritic/martensitic steels are discussed. Most irradiation studies have been carried out in fast reactors, where little transmutation helium forms. Helium has been shown to enhance swelling and affect tensile and fracture behavior, making helium a critical issue, since high helium concentrations will be generated in conjunction with displacement damage in a fusion reactor. These issues are reviewed to evaluate the status of ferritic/martensitic steels and to assess the research required to insure that such steels are viable candidates for fusion applications

Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

International Nuclear Information System (INIS)

2005-06-01

This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

Intense neutron irradiation facility for fusion reactor materials

Energy Technology Data Exchange (ETDEWEB)

Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

Current status of reduced-activation ferritic/martensitic steels R and D for fusion energy

International Nuclear Information System (INIS)

Kimura, Akihiko

2005-01-01

Reduced-activation ferritic/martensitic (RAF/M) steels have been considered to be the prime candidate for the fusion blanket structural material. The irradiation data obtained up to now indicates rather high feasibility of the steels for application to fusion reactors because of their high resistance to degradation of material performance caused by both the irradiation-induced displacement damage and transmutation helium atoms. The martensitic structure of RAF/M steels consists of a large number of lattice defects before the irradiation, which strongly retards the formation of displacement damage through absorption and annihilation of the point defects generated by irradiation. Transmutation helium can be also trapped at those defects in the martensitic structure so that the growth of helium bubbles at grain boundaries is suppressed. The major properties of the steels are well within our knowledge, and processing technologies are mostly developed for fusion application. RAF/M steels are now certainly ready to proceed to the next stage, that is, the construction of International Thermo-nuclear Experimental Reactor Test Blanket Modules (ITER-TBM). Oxide dispersion strengthening (ODS) steels have been developed for higher thermal efficiency of fusion power plants. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening accompanied by no loss of ductility. High-Cr ODS steels whose chromium concentration was in the range from 14 to 19 mass% showed high resistance to corrosion in supercritical pressurized water. It is shown that the 14Cr-ODS steel is susceptible to neither hydrogen nor helium embrittlement. A combined utilization of ODS steels with RAF/M steels will be effective to realize fusion power early at a reasonable thermal efficiency. (author)

Assessment of martensitic steels for advanced fusion reactors

International Nuclear Information System (INIS)

Wareing, J.; Tavassoli, A.A.

1995-01-01

Martensitic steels are currently considered in Europe to be prime structural candidate materials for the first wall and breeding blanket of the DEMO fusion reactor. In this design, reactor power and wall loading will be significantly higher than those of an experimental reactor. ITER and will give rise to component operating temperatures in the range 250 to 550 0 C with neutron doses higher than 70 dpa. These conditions render austenitic stainless steel, which will be used in ITER, less favourable. Factors contributing to the promotion of martensitic steels are their excellent resistance to irradiation induced swelling, low thermal expansion and high thermal conductivity allied to advanced industrial maturity, compared to other candidate materials vanadium alloys. This paper described the development and optimisation of the steel and weld metal. Using data design rules generated on modified 9 Cr 1 Mo steel during its qualification as a steam generator material for the European Fast Reactor (EFR), interim design guidelines are formulated. Whilst the merits of the steel are validated, it is shown that irradiation embrittlement at low temperature, allied to the need for prolonged post-weld hat treatment and the long term creep response of welds remain areas of some concern. (author). 18 refs., 6 figs., 2 tabs

Neutron irradiation control in the neutron transmutation doping process in HANARO using SPND

Energy Technology Data Exchange (ETDEWEB)

Kang, Gi-Doo; Kim, Myong-Seop [Korea Atomic Energy Research Institute, Yuseong, Daejeon, 305-353, (Korea, Republic of)

2015-07-01

The neutron irradiation control method by using self-powered neutron detector (SPND) is developed for the neutron transmutation doping (NTD) application in HANARO. An SPND is installed at a fixed position of the upper part of the sleeve in HANARO NTD hole for real-time monitoring of the neutron irradiation. It is confirmed that the SPND is significantly affected by the in-core condition and surroundings of the facility. Furthermore, the SPND signal changes about 15% throughout a whole cycle according to the change of the control rod position. But, it is also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation can is not so big while moving of the control rod. Accordingly, the relationship between the ratio of the neutron flux to the SPND signal output and the control rod position is established. In this procedure, the neutron flux measurement by using zirconium foil is utilized. The real NTD irradiation experiments are performed using the established relationship. The irradiated neutron fluence can be controlled within ±1.3% of the target one. The mean value of the irradiation/target ratio of the fluence is 0.9992, and the standard deviation is 0.0071. Thus, it is confirmed that the extremely accurate irradiation would be accomplished. This procedure can be useful for the SPND application installed at the fixed position to the field requiring the extremely high accuracy. (authors)

Conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at SINQ

International Nuclear Information System (INIS)

Zanini, L.; Baluc, N.; Simone, A. De; Eichler, R.; Joray, S.; Manfrin, E.; Pouchon, M.; Rabaioli, S.; Schumann, D.; Welte, J.; Zhernosekov, K.

2011-12-01

This comprehensive, illustrated report by the Paul Scherrer Institute PSI in Switzerland documents the proposals concerning the conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at the PSI's Swiss Spallation Neutron Source SINQ facility. The need for fast neutron irradiation is discussed and the possibility of using SINQ as a fast neutron irradiation facility is considered. The production of isotopes, tracers and medical isotopes is discussed, as are fission and fusion reactor technologies. The characteristics of the neutron spectrum in SINQ are discussed. The neutronic and radioprotection calculations for an irradiation station at SINQ are looked at in detail and extensive examples of work done and results obtained are presented and discussed. Radioprotection issues are also looked at. Further contributions in the report cover the hot/cold irradiation station in the SINQ target. An appendix provides detailed drawings of the facility's pneumatic delivery system

APT characterization of high nickel RPV steels

International Nuclear Information System (INIS)

Miller, M.K.; Russell, K.F

2004-01-01

Full text: The microstructures of several high nickel content pressure vessel steels have been characterized by atom probe tomography. The purposes of this study were to investigate the influence of high nickel levels on the response to neutron irradiation of high and low copper pressure vessel steels and to establish whether any additional phases were present after neutron irradiation. The nickel levels in these steels were at least twice that typically found in Western pressure vessel steels. Two different types of pressure vessel steels with low and high copper contents were selected for this study. The first set of alloys was low copper (∼0.05% Cu) base (15Ch2NMFAA) and weld (12Ch2N2MAA) materials used in a VVER-1000 reactor. The composition of the lower nickel VVER-1000 base material was Fe- 0.17 wt% C, 0.30% Si, 0.46% Mn, 2.2% Cr, 1.26% Ni, 0.05% Cu, 0.01% S, 0.008% P, 0.10% V and 0.50% Mo. The composition of the higher nickel VVER-1000 weld material was Fe- 0.06 wt % C, 0.33% Si, 0.80% Mn, 1.8% Cr, 1.78% Ni, 0.07% Cu, 0.009% S, 0.005% P, and 0.63% Mo. The VVER-1000 steels were irradiated in the HSSI Program's irradiation facilities at the University of Michigan, Ford Nuclear Reactor at a temperature of 288 o C for 2,137 h at an average flux of 7.08 x 10 11 cm 2 s -1 for a fluence of 5.45 x 10 18 n cm -2 (E >1 MeV) and for 5,340 h at an average flux of 4.33 x 10 11 cm -2 s -1 for a fluence of 8.32 x 10 1 28 n cm -2 (E >1 MeV). Therefore, the total fluence was 1.38 x 10 19 n cm -2 (E >1 MeV). The second type of pressure vessel steel was a high copper (0.20% Cu) weld from the Palisades reactor. The average composition of the Palisades weld was Fe- 0.11 wt% C, 0.18% Si, 1.27% Mn, 0.04% Cr, 1.20% Ni, 0.20% Cu, 0.017% S, 0.014% P, 0.003% V and 0.55% Mn. The Palisades weld, designated weldment 'B' from weld heat 34B009, was irradiated at a temperature of 288 o C and a flux of ∼7 x 10 11 cm -2 s -1 to a fast fluence of 1.4 x 10 19 n cm -2 (E >1 MeV). These three

Stress-induced martensitic transformations in NiTi and NiTi-TiC composites investigated by neutron diffraction

International Nuclear Information System (INIS)

Vaidyanathan, R.; Dunand, D.C.

1999-01-01

Superelastic NiTi (51.0 at.% Ni) specimens reinforced with 0, 10 and 20 vol.% TiC particles were deformed under uniaxial compression while neutron diffraction spectra were collected. The experiments yielded in-situ measurements of the thermoelastic stress-induced transformation. The evolution of austenite/martensite phase fractions and of elastic strains in the reinforcing TiC particles and the austenite matrix were obtained by Rietveld refinement during the loading cycle as the austenite transforms to martensite (and its subsequent back transformation during unloading). Phase fractions and strains are discussed in terms of load transfer in composites where the matrix undergoes a stress-induced phase transformation. (orig.)

Effect of microstructure on helium bubble growth in irradiated nickel

International Nuclear Information System (INIS)

Sattler, M.L.

1986-01-01

Thin nickel films were irradiated with 80 keV helium ions at varying doses and varying temperatures in order to obtain a variety of final microstructures. The growth of bubbles was examined during in-situ irradiations at 950 0 C where migration and coalescence events were observed for bubbles as large as 60 nm. Further direct observations of bubble growth were made during annealing of the irradiated specimens. For sample with no visible bubbles before annealing, the heating to 0.51 T/sub M/ produced bubbles that increased in diameter with annealing time to the power n. For bubbles in the grain interior, n ∼ 1, and on the grain boundaries, n ∼ 0.6. Since no migration and coalescence or ripening theories predict this behavior, a theory described by transient diffusion to spherical sinks was developed to discuss the behavior. This theory predicts that n = 1 for bubbles growing in the grain interior and n = 0.5 for bubbles on the grain boundary. In other annealing of irradiated samples containing large bubble populations, the growth of large bubbles and shrinking of small bubbles was observed at a temperature equal to 0.54 T/sub M/. The theory of Ostwald ripening properly described this type of bubble growth. Mass spectrometer measurements of He content in the irradiated specimens showed a greater He retention in the Ni films that contained a significant bubble population than those with no visible bubbles

Nanocluster irradiation evolution in Fe-9%Cr ODS and ferritic-martensitic alloys

Science.gov (United States)

Swenson, M. J.; Wharry, J. P.

2017-12-01

The objective of this study is to evaluate the influence of dose rate and cascade morphology on nanocluster evolution in a model Fe-9%Cr oxide dispersion strengthened steel and the commercial ferritic/martensitic (F/M) alloys HCM12A and HT9. We present a large, systematic data set spanning the three alloys, three irradiating particle types, four orders of magnitude in dose rate, and doses ranging 1-100 displacements per atom over 400-500 °C. Nanoclusters are characterized using atom probe tomography. ODS oxide nanoclusters experience partial dissolution after irradiation due to inverse Ostwald ripening, while F/M nanoclusters undergo Ostwald ripening. Damage cascade morphology is indicative of nanocluster number density evolution. Finally, the effects of dose rate on nanocluster morphology provide evidence for a temperature dilation theory, which purports that a negative temperature shift is necessary for higher dose rate irradiations to emulate nanocluster evolution in lower dose rate irradiations.

Thermal Decompositon Studies Of Pre-Irradiated Nickel (II) Azides ...

African Journals Online (AJOL)

The effect of pre-irradiation on the thermal decomposition of three samples of nickel (II) azide was studied. It was found that the rates of thermal decomposition of Ni(OH)N3 increased substantially with increase in pre-irradiation dosage. The initial reaction rates change from time-dependant nucleation law for the unirradiated ...

Neutron irradiation of bacteriophage λ

International Nuclear Information System (INIS)

Bozin, D.; Milosevic, M. . E-mail address of corresponding author: bozinde@vin.bg.ac.yu

2005-01-01

Double strand breaks (DSB) are the most dangerous lesions in DNA caused by irradiation, but many other lesions, usually called mutations, have not been clearly identified. These lesions, like DSB, can be the source of serious chromosomal damages and finally - cell death. Growing interest in heavy particles for radiotherapy and radioprotection encourages the search of the molecular basis of their action. In this respect, we chose bacteriophage λ1390 as the model system for the study of consequences of neutron irradiation. This derivative of λ phage possesses an unique ability to reversibly reorganize their genome in response to various selective pressures. The phages were irradiated with 13 Gy of mixed neutrons (7.5 Gy from fast and 5.6 Gy from thermal neutrons) and phages genomes were tested to DSB and mutations. Additionally, the stability of λ capsid proteins were tested. After all tests, we can conclude that, under our conditions, low flux of neutrons does not induce neither DNA strand break or DNA mutation nor the stability of λ capsid proteins. (author)

Fast-neutron dosimetry in the seed-irradiation facility, ASTRA reactor, Seibersdorf

International Nuclear Information System (INIS)

Ahnstroem, G.; Burtscher, A.; Casta, J.

1967-01-01

An important part of the co-ordinated programme on the neutron irradiation of seeds has been the construction of a fast-neutron irradiation facility for swimming-pool reactors. This facility was installed around 70 cm from the core in the ASTRA reactor swimming-pool at the end of December, 1966. Also, for this programme a pair of constant potential ionization chambers have been constructed at the Institute of Biochemistry, Stockholm University. These chambes are of the type described in the technical annex and are the same size as the seed-irradiation vials to be used in the seed-irradiation container (diam. =15 mm, length = 60 mm). Some preliminary dosimetry experiments were undertaken to test the irradiation facility and the ionization chambers, and to investigate the usefulness of the dosimetry instructions in the Technical Annex. The results of these experiments are discussed in this paper. 3 refs, 6 figs, 7 tabs

Development of new ferritic / martensitic steels for fuel cladding in fast neutron reactors

International Nuclear Information System (INIS)

Ratti, M.

2009-11-01

Many studies are directed toward the development of ferritic / martensitic ODS materials for applications in Gen IV programs. In this study, the mechanisms of formation of nano-phases (Y, Ti, O) and the influence of titanium on the precipitation refinement have been analyzed by small angle neutron scattering, X-ray diffraction and neutron diffraction. The obtained results allow developing new materials reinforced by nitrides (NDS which stands for Nitride Dispersion Strengthened). A first CEA patent is now being registered on these NDS materials processed by mechanical alloying. However, microstructural and mechanical characterizations are necessary to improve these new alloys. At last, a tensile and creep database has been acquired on an ODS Fe-18Cr material between room temperature and 650 C. These tests allow a qualitative description of the ODS mechanical behaviour. (author)

Neutron diffraction study of stress-induced martensitic transformation and variant change in Fe-Pd shape memory alloy

International Nuclear Information System (INIS)

Oliver, E.C.; Mori, T.; Daymond, M.R.; Withers, P.J.

2003-01-01

Neutron diffraction spectra were recorded during tensile testing of Fe-30.5 at.% Pd shape memory alloy at temperatures above M s and below M f . Peak intensity changes indicate that the application of tensile stress to initially fully austenitic material results in the preferential martensitic transformation of grains oriented with austenite parallel to the tensile axis. Tensile stress applied to initially fully martensitic material causes the greatest extent of reorientation in those variants oriented with martensite lying parallel to the tensile axis. These results are interpreted using a simple elasticity-based theory. Additionally, diffraction peak shifts provide information on the development of lattice strain in differently oriented grain families during loading. This indicates that above M s the alloy exhibits high single crystal elastic anisotropy. Below M f the apparent stiffnesses of different grain families suggest that axially compressive internal stresses develop in those grain families in which most variant reorientation occurs. These stresses act to reverse the variant changes upon subsequent unloading

Fission neutron irradiation of copper containing implanted and transmutation produced helium

DEFF Research Database (Denmark)

Singh, B.N.; Horsewell, A.; Eldrup, Morten Mostgaard

1992-01-01

High purity copper containing approximately 100 appm helium was produced in two ways. In the first, helium was implanted by cyclotron at Harwell at 323 K. In the second method, helium was produced as a transmutation product in 800 MeV proton irradiation at Los Alamos, also at 323 K. The distribut......High purity copper containing approximately 100 appm helium was produced in two ways. In the first, helium was implanted by cyclotron at Harwell at 323 K. In the second method, helium was produced as a transmutation product in 800 MeV proton irradiation at Los Alamos, also at 323 K...... as well as the effect of the presence of other transmutation produced impurity atoms in the 800 MeV proton irradiated copper will be discussed....

Resistivity recovery of neutron-irradiated and cold-worked thorium

International Nuclear Information System (INIS)

Tang, J.T.

1976-01-01

Recovery of neutron-irradiated and cold-worked thorium was studied using electrical resistivity measurements. Thorium wires containing 30 and 300 wt ppM carbon were irradiated to fast neutron fluence of 1.3 x 10 18 n/cm 2 (E greater than 0.1 MeV). Another group of thorium wires containing 45, 300 and 600 wt ppM carbon were laterally compressed 5 to 40 percent. Both irradiation and cold-working were performed at liquid nitrogen temperature. The induced resistivity was found to increase with carbon content for both treatments. Isochronal recovery studies were performed in the 120--420 0 K temperature range. Two recovery stages (II and III) were found for both cold-worked and irradiated samples. In all cases the activation energies were determined by use of the ratio-of-slope method. Consistent results were observed for both irradiated and cold-worked specimens within the experimental error in the two stages. Other methods were also used in determining the activation energy of stage III for irradiated samples. All analysis methods indicated that the activation energies decreased with increasing carbon content for differently treated specimens. Possible reasons for such behavior are discussed. The annealing data obtained do not fit a simple chemical rate equation but follow the empirical exponential equation proposed by Avrami. A model of detrapping of interstitials from impurities is suggested for stage II recovery. On the basis of the observed low activation energy and high retention of defects above stage III, a divacancy migration model is proposed for stage III recovery

Triple ion-beam studies of radiation damage in 9Cr2WVTa ferritic/martensitic steel

International Nuclear Information System (INIS)

Lee, E.H.; Hunn, J.D.; Rao, G.R.; Klueh, R.L.; Mansur, L.K.

1997-01-01

To simulate radiation damage under a future Spallation Neutron Source (SNS) environment, irradiation experiments were conducted on a candidate 9Cr-2WVTa ferritic/martensitic steel using the Triple Ion Facility (TIF) at ORNL. Irradiation was conducted in single, dual, and triple ion beam modes using 3.5 MeV Fe ++ , 360 keV He + , and 180 keV H + at 80, 200, and 350 degrees C. These irradiations produced various defects comprising black dots, dislocation loops, line dislocations, and gas bubbles, which led to hardening. The largest increase in hardness, over 63 %, was observed after 50 dpa for triple beam irradiation conditions, revealing that both He and H are augmenting the hardening. Hardness increased less than 30 % after 30 dpa at 200 degrees C by triple beams, compatible with neutron irradiation data from previous work which showed about a 30 % increase in yield strength after 27.2 dpa at 365 degrees C. However, the very large concentrations of gas bubbles in the matrix and on lath and grain boundaries after these simulated SNS irradiations make predictions of fracture behavior from fission reactor irradiations to spallation target conditions inadvisable

Neutron Dosimetry and Irradiation of Solids; Dosimetrie des neutrons et irradiation des solides

Energy Technology Data Exchange (ETDEWEB)

Perriot, G; Schmitt, A P [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1962-07-01

Results of work at C.E.A. from 1958 to 1960 are reviewed. The possibilities offered by classical dosimetry methods are discussed. The tests which led to the utilization, for fast neutron dosimetry, of resistivity variations induced in solid W by such neutrons are described. Experimental W irradiation results led to a definition of neutron efficiency which describes the relations between neutron energy and their effects on materials. Possibilities offered by detectors which make use of radiation damage and are sensitive to neutrons at keV energies were explored. In other work, the principal French reactors were classified according to their ability to produce damage in materials such as W. (authors) [French] Dans ce rapport on a presente les resultats essentiels de travaux qui ont ete effectues de 1958 a 1980 par des chercheurs du CEA issus de differents services. En meme temps qu'une revue des possibilites offertes a l'epoque par les methodes classiques de dosimetrie (utilisation des detecteurs par activation), on a decrit les essais qui devaient permettre d'utiliser, a la dosimetrie les neutrons rapides, les variations de resistivite qu'ils creent dans un corps solide (tungstene). L'irradiation du tungstene a montre l'importance qu'il y avait a definir 'l'efficacite' des neutrons, c'est-a-dire leur aptitude plus ou moins grande, selon leur energie, a creer des defauts dans les materiaux. L'efficacite d'un emplacement d'irradiation se trouvant liee au spectre neutronique, on a vu les difficultes qu'il y avait a utiliser les detecteurs par activation des qu'on n'avait plus affaire a un spectre en 1/E ou de fission et on a pu entrevoir les possibilites offertes par les detecteurs utilisant la creation des defauts qui repondent a tous les neutrons d'energies, superieures a quelques keV. Enfin, on a classe les principaux types de Piles Francaises selon leur aptitude a creer plus ou moins rapidement des dommages dans des materiaux comme le tungstene. (auteur)

Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

Science.gov (United States)

Kim, Myong-Seop; Park, Sang-Jun

2009-08-01

Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

Chemical reactions induced by fast neutron irradiation

International Nuclear Information System (INIS)

Katsumura, Y.

1989-01-01

Here, several studies on fast neutron irradiation effects carried out at the reactor 'YAYOI' are presented. Some indicate a significant difference in the effect from those by γ-ray irradiation but others do not, and the difference changes from subject to subject which we observed. In general, chemical reactions induced by fast neutron irradiation expand in space and time, and there are many aspects. In the time region just after the deposition of neutron energy in the system, intermediates are formed densely and locally reflecting high LET of fast neutrons and, with time, successive reactions proceed parallel to dissipation of localized energy and to diffusion of the intermediates. Finally the reactions are completed in longer time region. If we pick up the effects which reserve the locality of the initial processes, a significant different effect between in fast neutron radiolysis and in γ-ray radiolysis would be derived. If we observe the products generated after dissipation and diffusion in longer time region, a clear difference would not be observed. Therefore, in order to understand the fast neutron irradiation effects, it is necessary to know the fundamental processes of the reactions induced by radiations. (author)

Irradiation test of HAFM and tag gas samples at the standard neutron field of 'YAYOI'

International Nuclear Information System (INIS)

Iguchi, Tetsuo

1997-03-01

To check the accuracy of helium accumulation neutron fluence monitors (HAFM) as new technique for fast reactor neutron dosimetry and the applicability of tag gas activation analysis to fast reactor failed fuel detection, their samples were irradiated at the standard neutron field of the fast neutron source reactor 'YAYOI' (Nuclear Engineering Research Laboratory, University of Tokyo). Since October in 1996, the HAFM samples such as 93% enriched boron (B) powders of 1 mg and natural B powders of 10 mg contained in vanadium (V) capsule were intermittently irradiated at the reactor core center (Glory hole: Gy) and/or under the leakage neutron field from the reactor core (Fast column: FC). In addition, new V capsules filled with enriched B of 40 mg and Be of 100 mg, respectively, were put into an experimental hole through the blanket surrounding the core. These neutron fields were monitored by the activation foils consisting of Fe, Co, Ni, Au, 235 U, 237 Np etc., mainly to confirm the results obtained from 1995's preliminary works. In particular, neutron flux distributions in the vicinity of irradiated samples were measured in more detail. At the end of March in 1997, the irradiated neutron fluence have reached the goal necessary to produce the detectable number of He atoms more than ∼10 13 in each HAFM sample. Six kinds of tag gas samples, which are the mixed gases of isotopically adjusted Xe and Kr contained in SUS capsules, were separately irradiated three times at Gy under the neutron fluence of ∼10 16 n/cm 2 in average. After irradiation, γ-ray spectra were measured for each sample. Depending on the composition of tag gas mixtures, the different patterns of γ-ray peak spectra from 79 Kr, 125 Xe, etc. produced through tag gas activation were able to be clearly identified. These experimental data will be very useful for the benchmark test of tag gas activation calculation applied to the fast reactor failed fuel detection. (author)

Hair dosimetry following neutron irradiation.

Science.gov (United States)

Lebaron-Jacobs, L; Gaillard-Lecanu, E; Briot, F; Distinguin, S; Boisson, P; Exmelin, L; Racine, Y; Berard, P; Flüry-Herard, A; Miele, A; Fottorino, R

2007-05-01

Use of hair as a biological dosimeter of neutron exposure was proposed a few years ago. To date, the (32)S(n,p)(32)P reaction in hair with a threshold of 2.5 MeV is the best choice to determine the fast neutron dose using body activation. This information is essential with regards to the heterogeneity of the neutron transfer to the organism. This is a very important parameter for individual dose reconstruction from the surface to the deeper tissues. This evaluation is essential to the adapted management of irradiated victims by specialized medical staff. Comparison exercises between clinical biochemistry laboratories from French sites (the CEA and COGEMA) and from the IRSN were carried out to validate the measurement of (32)P activity in hair and to improve the techniques used to perform this examination. Hair was placed on a phantom and was irradiated at different doses in the SILENE reactor (Valduc, France). Different parameters were tested: variation of hair type, minimum weight of hair sample, hair wash before measurement, delivery period of results, and different irradiation configurations. The results obtained in these comparison exercises by the different laboratories showed an excellent correlation. This allowed the assessment of a dose-activity relationship and confirmed the feasibility and the interest of (32)P measurement in hair following fast neutron irradiation.

The intrinsic gettering in neutron irradiation Czochralski-silicon

CERN Document Server

Li Yang Xian; Niu Ping Juan; Liu Cai Chi; Xu Yue Sheng; Yang Deren; Que Duan Lin

2002-01-01

The intrinsic gettering in neutron irradiated Czochralski-silicon is studied. The result shows that a denuded zone at the surface of the neutron irradiated Czochralski-silicon wafer may be formed through one-step short-time annealing. The width of the denuded zone is dependent on the annealing temperature and the dose of neutron irradiation, while it is irrelated to the annealing time in case the denuded zone is formed. The authors conclude that the interaction between the defects induced by neutron irradiation and the oxygen in the silicon accelerates the oxygen precipitation in the bulk, and becomes the dominating factor of the quick formation of intrinsic gettering. It makes the effect of thermal history as the secondary factor

Work Hardening, Dislocation Structure, and Load Partitioning in Lath Martensite Determined by In Situ Neutron Diffraction Line Profile Analysis

Science.gov (United States)

Harjo, Stefanus; Kawasaki, Takuro; Tomota, Yo; Gong, Wu; Aizawa, Kazuya; Tichy, Geza; Shi, Zengmin; Ungár, Tamas

2017-09-01

A lath martensite steel containing 0.22 mass pct carbon was analyzed in situ during tensile deformation by high-resolution time-of-flight neutron diffraction to clarify the large work-hardening behavior at the beginning of plastic deformation. The diffraction peaks in plastically deformed states exhibit asymmetries as the reflection of redistributions of the stress and dislocation densities/arrangements in two lath packets: soft packet, where the dislocation glides are favorable, and hard packet, where they are unfavorable. The dislocation density was as high as 1015 m-2 in the as-heat-treated state. During tensile straining, the load and dislocation density became different between the two lath packets. The dislocation character and arrangement varied in the hard packet but hardly changed in the soft packet. In the hard packet, dislocations that were mainly screw-type in the as-heat-treated state became primarily edge-type and rearranged towards a dipole character related to constructing cell walls. The hard packet played an important role in the work hardening in martensite, which could be understood by considering the increase in dislocation density along with the change in dislocation arrangement.

Detection and measurement of neutron-irradiated gemstones

International Nuclear Information System (INIS)

Bunnak, S.; Jerachanchai, S.; Chinudomsub, K.; Saiyut, K.

1990-01-01

Color enhance gemstone, neutron-irradiated topaz, was analyzed by gamma spectrometry for examining characteristic and activity. Topaz was irradiated in the wet-tube facility of the Research Reactor TRR/1 which neutron fluence is 2.52x10 17 neutron per square centimeter. After 100 days of decay, topaz was sampling to the qualitative and quantitative analysis using multichannel analyzer of Nuclear Data Model ND65 and hyper pure germanium detector. Calculation and evaluation were done by microcomputer IBM/PC 640 KB RAM. The qualitative analysis showed that the neutron-irradiated topaz has 2 major isotopes, i.e., Ta-182 and Sc-46. Quantitative activity was compared with reference standard source Eu-152 (NBS) and the results were shown in the table 1. The Health Physics Division, OAEP, inspected on 6240.9 gm of the neutron-irradiated topaz using standard release limit 2 nCi/gm (74 Bq/gm). It was found that only 423.9 gm out of the total amount were over the standard release limit

Polarizing neutron by light-irradiated graphene

International Nuclear Information System (INIS)

Peng, Feng

2015-01-01

We study the spin orientation of the neutron scattered by light-irradiated graphene and calculate the average value of spin z-component of the neutron in terms of a generating functional technique. Our calculation results indicate that there is a remarkable neutron polarization effect when a neutron penetrates graphene irradiated by a circularly polarized light. We analyse the dynamical source of generating this effect from the aspect of photon-mediated interaction between the neutron spin and valley pseudospin. By comparing with the polarization induced by a magnetic field, we find that this polarization may be equivalent to the one led by a magnetic field of several hundred Teslas if the photon frequency is in the X-ray frequency range. This provides an approach of polarizing neutrons. (copyright 2015 by WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Non-destructive Determination of Martensitic Content by Means of Magnetic Methods

Energy Technology Data Exchange (ETDEWEB)

Niffenegger, M.; Bauer, R.; Kalkhof, D

2003-07-01

The detection of material degradation in a pre-cracked stage would be very advantageous. Therefore the main objective of the EC 5th Framework Programme Project CRETE (Contract No. FIS5-1999-00280) was to assess the capability and the reliability of innovative NDT-inspection techniques for the detection of material degradation, induced by low cycle fatigue (LCF) and neutron irradiation of metastable austenitic and ferritic low-alloy steel. Within work package WP6 and WP7 several project partners tested aged or irradiated samples, using various advanced measuring techniques, such as acoustic, magnetic and thermoelectric ones. These indirect methods require a careful interpretation of the measured signal in terms of micro-structural evolutions due to ageing of the material. Therefore the material had to be characterized in its undamaged, as well as in its damaged state. Based on results from former investigations, main attention was paid to the content of martensitic phase as an indicator for fatigue. Since most NDT-methods are considered as indirect methods for the detection of martensite, neutron diffraction was applied as a reference method for a quantitative determination of martensite. The material characterization performed at PSI and INSA de Lyon is published in the PSI Bericht Nr. 03-17, July 2003, (ISSN 1019-0643). The present report only describes the magnetic methods applied at PSI for the detection of material degradation and summarises the results obtained in WP3 of the CRETE project. The report is issued simultaneously as a PSI report and the CRETE work package WP3 report. At PSI the following magnetic methods were applied to LCF specimens: (1) Ferromaster for measuring the magnetic permeability, (2) Eddy current impedance measuring by means of a Giant Magneto Resistance sensor (GMR), (3) Remanence field measurements using high sensitive Fluxgate and SQUID sensors. With these methods three sets of fatigue specimens, made from different metastable

Stress-induced martensite variant reorientation in magnetic shape memory Ni–Mn–Ga single crystal studied by neutron diffraction

Czech Academy of Sciences Publication Activity Database

Molnár, Peter; Šittner, Petr; Lukáš, Petr; Hannula, S.-P.; Heczko, Oleg

2008-01-01

Roč. 17, č. 3 (2008), 035014/1-035014/4 ISSN 0964-1726 Institutional research plan: CEZ:AV0Z10100520; CEZ:AV0Z10480505 Keywords : NiMnGa single crystal * neutron diffraction * stress induced martensite reorientation Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.743, year: 2008

Needs of in-situ materials testing under neutron irradiation

International Nuclear Information System (INIS)

Noda, K.; Hishinuma, A.; Kiuchi, K.

1989-01-01

Under neutron irradiation, the component atoms of materials are displaced as primary knock-on atoms, and the energy of the primary knock-on atoms is consumed by electron excitation and nuclear collision. Elementary irradiation defects accumulate to form damage structure including voids and bubbles. In situ test under neutron irradiation is necessary for investigating into the effect of irradiation on creep behavior, the electric properties of ceramics, transport phenomena and so on. The in situ test is also important to investigate into the phenomena related to the chemical reaction with environment during irradiation. Accelerator type high energy neutron sources are preferable to fission reactors. In this paper, the needs and the research items of in situ test under neutron irradiation using a D-Li stripping type high energy neutron source on metallic and ceramic materials are described. Creep behavior is one of the most important mechanical properties, and depends strongly on irradiation environment, also it is closely related to microstructure. Irradiation affects the electric conductibity of ceramics and also their creep behavior. In this way, in situ test is necessary. (K.I.)

Tensile and low cycle fatigue properties of EUROFER97-steel after 16.3 dpa neutron irradiation at 523, 623 and 723 K

Energy Technology Data Exchange (ETDEWEB)

Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu; Möslang, A., E-mail: anton.moeslang@kit.edu; Schneider, H.-C., E-mail: hans-christian.schneider@kit.edu

2013-11-15

Neutron-irradiated specimens of the reduced-activation tempered martensitic steel EUROFER97 were tested by tensile and low cycle conditions to detect the impact of irradiation on strength and lifetime. The irradiation temperature ranged from 523 to 723 K with an accumulated dose of up to 16.3 dpa. Tensile tests revealed a significant irradiation-induced hardening below 673 K with a peak of ∼430 MPa at 573 K but none was seen at 723 K, as expected. Despite the significant irradiation-induced reduction of uniform elongation, the total elongation is only reduced by about 50% below 673 K. Post-irradiation strain-controlled fatigue tests have been carried out at T{sub irrad} = T{sub test} = 523, 623 and 723 K. Pronounced cyclic softening was observed in all specimens. At 623 and 723 K, neutron irradiation had no effect on fatigue life within the data scatter. A significant lifetime increase has been observed at T{sub irrad} = T{sub test} = 523 K that advances with decreasing stress amplitude Δε (1% → 0.5%) up to a factor of ten. Scanning electron microscopy (SEM) analysis revealed ductile fracture and fatigue striations on the fracture surfaces. After push–pull fatigue testing, transmission electron microscopy (TEM) investigations showed the typical sub-cell formation, even at T{sub irrad} = T{sub test} = 523 K.

Defect structure in proton-irradiated copper and nickel

International Nuclear Information System (INIS)

Tsukuda, Noboru; Ehrhart, P.; Jaeger, W.; Schilling, W.; Dworschak, F.; Gadalla, A.A.

1987-01-01

This single crystals of copper or nickel with a thickness of about 10 μm are irradiated with 3 MeV protons at room temperature and the structures of resultant defects are investigated based on measurements of the effects of irradiation on the electrical resistivity, length, lattice constants, x-ray diffraction line profile and electron microscopic observations. The measurements show that the electrical resistivity increases with irradiation dose, while leveling off at high dose due to overlapping of irradiation cascades. The lattice constants decreases, indicating that many vacancies still remain while most of the interstitial stoms are eliminated, absorbed or consumed for dislocation loop formation. The x-ray line profile undergoes broadening, which is the result of dislocation loops, dislocation networks and SFT's introduced by the proton irradiation. Various defects have different effects though they cannot be identified separately from the profile alone. A satellite peak appears at a low angle, which seems to arise from periodic defect structures that are found in electron microscopic observations. In both copper and nickel, such periodic defect structures are seen over a wide range from high to low dose. Defect-free and defect-rich domains (defect walls), 0.5 to several μm in size, are alingned parallel to the {001} plane at intervals of 60 nm. The defect walls, which consist of dislocations, dislocation loops and SFT's, is 20 - 40 nm thick. (Nogami, K.)

DNA-repair after irradiation of cells with gamma-rays and neutrons

International Nuclear Information System (INIS)

Altmann, H.

1975-11-01

The structural alterations of calf thymus DNA produced by neutron or gamma irradiation were observed by absorption spectra, sedimentation rate and viscosity measurements. Mixed neutron-gamma irradiation produced fewer single and double strand breaks compared with pure gamma irradiation. RBE-values for mixed neutron-gamma radiation were less than 1, and DNA damage decreased with increasing neutron dose rate. Repair processes of DNA occuring after irradiation were measured in mouse spleen suspensions and human lymphocytes using autoradiographic methods and gradient centrifugations. The number of labelled cells was smaller after mixed neutron-gamma irradiation than after gamma irradiation. The rejoining of strand breaks in alkaline and neutral sucrose was more efficient after gamma irradiation than after mixed neutron-gamma irradiation. Finally, the effect of detergents Tween 80 and Nonident P40 on unscheduled DNA synthesis was studied by autoradiography after mixed neutron-gamma irradiation (Dn=5 krad). The results showed that the DNA synthesis was inhibited by detergent solutions of 0.002%

Study of boron carbide evolution under neutron irradiation

International Nuclear Information System (INIS)

Simeone, D.

1999-01-01

Owing to its high neutron efficiency, boron carbide (B 4 C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the 10 B(n,α) 7 Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B 4 C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the 7 Li(p,p'γ) 7 Li reaction, allows to measure lithium profile in B 4 C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B 4 C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B 4 C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B 4 C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B 4 C. (author)

Neutron-irradiation facilities at the Intense Pulsed Neutron Source-I for fusion magnet materials studies

International Nuclear Information System (INIS)

Brown, B.S.; Blewitt, T.H.

1982-01-01

The decommissioning of reactor-based neutron sources in the USA has led to the development of a new generation of neutron sources that employ high-energy accelerators. Among the accelerator-based neutron sources presently in operation, the highest-flux source is the Intense Pulsed Neutron Source (IPNS), a user facility at Argonne National Laboratory. Neutrons in this source are produced by the interaction of 400 to 500 MeV protons with either of two 238 U target systems. In the Radiation Effects Facility (REF), the 238 U target is surrounded by Pb for neutron generatjion and reflection. The REF has three separate irradiation thimbles. Two thimbles provide irradiation temperatures between that of liquid He and several hundred degrees centigrade. The third thimble operates at ambient temperature. The large irradiation volume, the neutron spectrum and flux, the ability to transfer samples without warm up, and the dedication of the facilities during the irradiation make this ideally suited for radiation damage studies on components for superconducting fusion magnets. Possible experiments for fusion magnet materials are discussed on cyclic irradiation and annealing of stabilizers in a high magnetic field, mechanical tests on organic insulation irradiated at 4 K, and superconductors measured in high fields after irradiation

Segregation of a copper-nickel alloy after electron irradiation

International Nuclear Information System (INIS)

Wagner, W.

1979-09-01

In the present work measurement of diffuse neutron scattering are used to determine short range segregation effects of the alloy Cu 0 sub(.) 414 Ni 0 sub(.) 586 after thermal annealing and 3 MeV electron irradiation in the temperature range between 370 K and 600 K. In addition neutron small angle scattering measurement are performed after irradiation to study possible long range segregation effects. Residual resistivity measurements are performed in parallel in order tp orientate the relatively expensive neutron scattering measurements with respect to the residual changes (orig./KBI) [de

Impact of neutron irradiation on thermal helium desorption from iron

Energy Technology Data Exchange (ETDEWEB)

Hu, Xunxiang, E-mail: hux1@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Field, Kevin G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Taller, Stephen [University of Michigan, Ann Arbor, MI 48109 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wirth, Brian D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States)

2017-06-15

The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 10{sup 18} He/m{sup 2}. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

Influence of helium embrittlement on post-irradiation creep rupture behaviour of austenitic and martensitic stainless steels

International Nuclear Information System (INIS)

Wassilew, C.

1982-01-01

The author has investigated the influence of helium embrittlement on the creep rupture properties of the austenitic stainless steels 1.4970 and 1.4962 and the martensitic stainless steel 1.4914 after irradiation in the BR-2 reactor in Mol, Belgium. The results show that austenitic steels react much more strongly to the embrittlement effect of the helium than do martensitic steels. The causes of the lower embrittlement tendency of the martensitic than of both austenitic stainless steels were analysed carefully. A new embrittlement model was developed on the basis of data derived from the creep rupture experiments, and reinforced by a simple metallographic investigation of the fracture zone and its immediate environment. This model pays specific attention to the role of the twin planes as the most efficient area of increased vacancy production, and takes into account the ability of the twin boundaries to transport these vacancies with reduced energy and low loss into the high-angle grain boundaries. (author)

Radiation damage of pixelated photon detector by neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Isamu [KEK, 1-1 Oho Tsukuba 305-0801 (Japan)], E-mail: isamu.nakamura@kek.jp

2009-10-21

Radiation Damage of Pixelated Photon Detector by neutron irradiation is reported. MPPC, one of PPD or Geiger-mode APD, developed by Hamamatsu Photonics, is planned to be used in many high energy physics experiments. In such experiments radiation damage is a serious issue. A series of neutron irradiation tests is performed at the Reactor YAYOI of the University of Tokyo. MPPCs were irradiated at the reactor up to 10{sup 12}neutron/cm{sup 2}. In this paper, the effect of neutron irradiation on the basic characteristics of PPD including gain, noise rate, photon detection efficiency is presented.

Swelling and fracturing of borides under neutron irradiation

International Nuclear Information System (INIS)

Krainy, A.G.; Ogorodnikov, V.V.; Grinik, E.U.; Chirko, L.I.; Shinakov, A.A.

1994-01-01

The neutron irradiation of high temperature borides, which are included in boron-containing reactor materials, results in high internal stresses, leading to considerable swelling and micro- and macro-fracturing. Experimental results over a large range of temperature and fluences, show a change of damage mechanism for borides within 400-530 C: the macro-cracking with formation of annular and radial cracks is observed below this temperature zone. The accumulation of micro-fractures and the process of gas swelling take place at irradiation temperatures above 530 C. The effect of the high internal stresses is compared to external pressure. 12 refs., 4 figs

Induced defects in neutron irradiated GaN single crystals

International Nuclear Information System (INIS)

Park, I. W.; Koh, E. K.; Kim, Y. M.; Choh, S. H.; Park, S. S.; Kim, B. G.; Sohn, J. M.

2005-01-01

The local structure of defects in undoped, Si-doped, and neutron irradiated free standing GaN bulk crystals, grown by hydride vapor phase epitaxy, has been investigated by employing Raman scattering and cathodoluminescence. The GaN samples were irradiated to a dose of 2 x 10 17 neutrons in an atomic reactor at Korea Atomic Energy Research Institute. There was no appreciable change in the Raman spectra for undoped GaN samples before and after neutron irradiation. However, a forbidden transition, A 1 (TO) mode, appeared for a neutron irradiated Si-doped GaN crystal. Cathodoluminescence spectrum for the neutron irradiated Si-doped GaN crystal became much more broadened than that for the unirradiated one. The experimental results reveal the generation of defects with locally deformed structure in the wurtzite Si-doped GaN single crystal

Design, Construction, and Modeling of a 252Cf Neutron Irradiator

Directory of Open Access Journals (Sweden)

Blake C. Anderson

2016-01-01

Full Text Available Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

A search for long-lived radionuclides produced by fast-neutron irradiations of copper, silver, europium, terbium, and hafnium

International Nuclear Information System (INIS)

Meadows, J.W.; Smith, D.L.; Ikeda, Y.; Konno, C.

1990-01-01

Identical sample packets, each containing samples of elemental copper, silver, europium, terbium, and hafnium, as well as titanium, iron and nickel as dosimeters, have been irradiated in three distinct accelerator neutron fields (at Argonne National Laboratory and Los Alamos National Laboratory in the U.S.A., and Japan Atomic Energy Research Institute, Tokai, Japan) as part of an interlaboratory research collaboration to search for the production of long-lived radionuclides for fusion waste disposal applications. This paper is a progress report on this project. To date, we have detected the following activities, and have obtained preliminary experimental cross section values for several of these: Ag-106m,108m,110m; Eu-150m,152g,154; Tb-158,160; and Hf-175,178m2,179m2,181. (author). 11 refs, 1 fig., 4 tabs

Irradiation hardening and localized deformation of neutron-irradiated α-iron single crystals

International Nuclear Information System (INIS)

Mughrabi, H.; Stroehle, D.; Wilkens, M.

1981-01-01

The early yielding behaviour of neutron-irradiated α iron single crystals orientated for single slip was investigated as a function of neutron dose. In the range of neutron doses between approx. equal to 10 18 and approx. equal to 10 19 n/cm 2 , the irradiation hardening increment was found to be almost constant. Qualitative modifications of this behaviour were observed in the case of predeformed specimens. The localized deformation of the neutron-irradiated specimens by dislocation channelling was investigated by slip-line observations, transmission electron microscopy and X-ray topography. A model of localized deformation is proposed in order to explain the development of the observed asymmetric dislocation double layers which bound the channels and transmit characteristic misorientations. (orig.)

Emulation of neutron irradiation effects with protons: validation of principle

International Nuclear Information System (INIS)

Was, G.S.; Busby, J.T.; Allen, T.; Kenik, E.A.; Jensson, A.; Bruemmer, S.M.; Gan, J.; Edwards, A.D.; Scott, P.M.; Andreson, P.L.

2002-01-01

This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 deg. C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 deg. C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 deg. C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a 'W' shaped profile at 1.0 dpa and then into a 'V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary

Thermal neutron converter for irradiations with fission neutrons

International Nuclear Information System (INIS)

Wagner, F.M.; Kampfer, S.; Kastenmuller, A.; Waschkowski, W.; Bucherl, Th.; Kampfer, S.

2007-01-01

The new research reactor FRM II at Garching started operation in March 2004. The compact core is cooled by light water, and moderated by heavy water. Two fuel plates mounted in the heavy water tank convert thermal to fast neutrons. The fast neutron flux in the connected beam tube is up to 7 centre dot 10 8 s -1 cm -2 (depending on filters and collimation); the mean neutron energy is about 1.6 MeV. There are two irradiation rooms along the beam. The first is mainly used for medical therapy (MEDAPP facility), the second for materials characterization (NECTAR facility). At the former therapy facility RENT at the old research reactor FRM, the same beam quality was available until July 2000. Therefore, only a small program is run for the determination of the biological effectiveness of the new beam. The neutron and gamma dose rates in the medical beam are 0.54 and 0.20 Gy/min, respectively. The therapy facility MEDAPP is still under examination according to European regulations for medical devices. Full medical operation will start in 2007. The radiography and tomography facility NECTAR is in operation and aims at non-destructive inspection of objects up to 400 kg mass and 80 centre dot 80 centre dot 80 cm 3 in size. As for fission neutrons the macroscopic cross section of hydrogen is much higher than for other materials (e. g. Fe and Pb), one special application is the detection of hydrogen-containing materials (e. g. oil) in dense materials

Neutron irradiation therapy machine

International Nuclear Information System (INIS)

1980-01-01

Conventional neutron irradiation therapy machines, based on the use of cyclotrons for producing neutron beams, use a superconducting magnet for the cyclotron's magnetic field. This necessitates complex liquid He equipment and presents problems in general hospital use. If conventional magnets are used, the weight of the magnet poles considerably complicates the design of the rotating gantry. Such a therapy machine, gantry and target facilities are described in detail. The use of protons and deuterons to produce the neutron beams is compared and contrasted. (U.K.)

Study of electroless Ni-W-P alloy coating on martensitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Nikitasari, Arini, E-mail: arini-nikitasari@yahoo.com; Mabruri, Efendi, E-mail: efendi-lipi@yahoo.com [Research Center for Metallurgy and Materials, Indonesian Institute of Sciences (470 Building, Puspiptek, Serpong, Indonesia 15313) (Indonesia)

2016-04-19

Electroless nickel phospor (Ni-P) is widely used in many industries due to their corrosion and wear resistance, coating uniformity, and ability to coat non-conductive surfaces. The unique properties of tungsten such as high hardness, higher melting point, lower coefficient of linear thermal expansion, and high tensile strength have created a lot of interest in developing ternary Ni-W-P alloys. This article presents the study of electroless Ni-W-P alloys coating using acid or alkaline bath on martensitic stainless steel. Nickel sulfate and sodium tungstate were used as nickel and tungsten sources, respectively, and sodium hypophosphite was used as a reducing agent. Acid or alkaline bath refer to bath pH condition was adjusted by adding sulfuric acid. Martensitic stainless steel was immersed in Ni-W-P bath for 15, 30, and 60 minutes. The substrate of martensitic stainless steel was subjected to pre-treatment (polishing and cleaning) and activation prior to electroless plating. The plating characteristics were investigated for concentration ratio of nickel and hypophosphite (1:3), sodium tungstate concentration 0,1 M, immersion time (15 min, 30 min, 60 min), and bath condition (acid, alkaline). The electroless Ni-W-P plating was heat treated at 400°C for 1 hour. Deposits were characterized using scanning electron microscope (SEM) and corrosion measurement system (CMS).

Final report on neutron irradiation at low temperature to investigate plastic instability and a high temperature to study cavitation

International Nuclear Information System (INIS)

Singh, B.N.; Eldrup, M.; Golubov, S.I.; Edwards, D.J.; Jung, P.

2005-01-01

Effects of neutron irradiation on defect accumulation and physical and mechanical properties of pure iron and F82H and EUROFER 97 ferritic-martensitic steels have been investigated. Tensile specimens were neutron irradiated to a dose level of 0,23 dpa at 333 and 573 K. Electrical resistivity and tensile properties were measured both in the unirradiated and irradiated condition. Some additional specimens of pure iron were irradiated at 333 K to doses of 10-3, 10-2 and 10-1 dpa and tensile tested at 333 K. To investigate the effect of helium on cavity nucleation and growth, specimens of pure iron and EUROFER 97 were implanted with different amounts of helium at 323 K and subsequently neutron irradiated to doses of 10-3, 10-2 and 10-1 dpa at 323 K. Defect microstructures were investigated using positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM). Numerical calculations, based on the production bias model (PBM) were carried out to study the details of evolution of cavities with and without helium generation. The phenomena of dislocation decoration and raft formation, which are important for understanding radiation hardening and plastic flow localization, have been studied using the Kinetic Monte Carlo (KMC) code during a realistic dynamic irradiation of bcc iron at 300 K. Molecular dynamics (MD) simulations have been carried out to study the stress dependencies of dislocation velocity and drag coefficient for an edge dislocation decorated with small SIA loops.The present report describes both experimental procedure and calculational methodology employed in the present work. The main results of all these investigations, both experimental and theoretical, are highlighted with appropriate examples. Finally, a brief summary is given of the main results conclusions. (au)

High-resolution electron microscopy studies of the precipitation of copper under neutron irradiation in an Fe-1.3WT% Cu alloy

International Nuclear Information System (INIS)

Nicol, A. C.

1998-01-01

We have studied by electron microscopy the copper-rich precipitates in an Fe-1.3wt%Cu model alloy irradiated with neutrons to doses of 8.61 x 10 -3 dpa and 6.3 x 10 -2 dpa at a temperature of ∼270 C. In the lower dose material a majority (ca. 60%)of the precipitates visible in high-resolution electron microscopy were timed 9R precipitates of size ∼2-4 nm, while ca. 40% were untwinned. In the higher dose material, a majority (ca. 75%) of visible precipitates were untwinned although many still seemed to have a 9R structure. The average angle α between the herring-bone fringes in the twin variants was measured as 125 degree, not the 129 degree characteristic of precipitates in thermally-aged and electron-irradiated material immediately after the bcc->9R martensitic transformation. We argue that these results imply that the bcc->9R transformation of small (<4 nm) precipitates under neutron irradiation takes place at the irradiation temperature of 270 C rather than after subsequent cooling. Preliminary measurements showed that precipitate sizes did not depend strongly on dose, with a mean diameter of 3.4 ± 0.7 nm for the lower dose material, and 3.0 ± 0.5 nm for the higher dose material. This result agrees with the previous assumption that the lack of coarsening in precipitates formed under neutron irradiation is a consequence of the partial dissolution of larger precipitates by high-energy cascades

Nickel and Copper-Free Sintered Structural Steels Containing Mn, Cr, Si, and Mo Developed for High Performance Applications

Directory of Open Access Journals (Sweden)

Cias A.

2017-03-01

Full Text Available In an attempt to study the sinterability of potential high-strength nickel-free sintered structural steels containing Mn, Cr, Si and Mo compacts were prepared based on sponge and water atomised iron powders and on Astaloy prealloyed powders. To these were admixed ferromanganese, ferroslicon, and graphite. The samples were sintered at temperatures 1120 and 1250°C in laboratory tube furnaces in hydrogen, hydrogen-nitrogen atmospheres with dew points better than -60°C or in nitrogen in a semiclosed container in a local microatmosphere. After sintering the samples were slowly cooled or sinterhardened. Generally resultant microstructures were inhomogeneous, consisted of pearlite/ bainite/martensite, but were characterised by an absence of oxide networks. Sintering studies performed over a range of compositions have shown that superior strength, ranging beyond 900 MPa, along with reasonable tensile elongation, can be achieved with these new steels.

Effects of iron spallation products Ti, P and S on the physical metallurgy of 9Cr martensitic steels

International Nuclear Information System (INIS)

Danylova, O.; Carlan, Y. de; Hamon, D.; Brachet, J.C.; Alamo, A.

2002-01-01

The design of an Accelerator Driven System (ADS) requires that the 'window', which separates the proton accelerator from the spallation target, be able to withstand very severe irradiation conditions. Fe-9/12Cr martensitic steels are good candidates for the window material due to their intrinsic stability under neutron irradiation, but the influence of iron spallation elements on their behaviour is not known. To elucidate the effects of the spallation elements titanium, phosphorus and sulphur on the behaviour of martensitic steels, it was decides to obtain different castings of 9Cr 1Mo steels doped with these elements. The aim of this paper is to present the data obtained on the physical metallurgy of these steels and to show the possible methods of obtaining titanium, phosphorus and sulphur in solid solution for subsequent study of the evolution of the microstructure and mechanical properties. (authors)

Effect of neutron irradiation on vitreous carbon

International Nuclear Information System (INIS)

Kurolenkin, E.I.; Virgil'ev, Yu.S.; Chugunova, T.K.

1989-01-01

The change in mass (m), volume (V), specific electric resistance (ρ), coefficient of linear thermal expansion (α), dynamic elasticity modulus (E), and limit of bending strength (σ) of vitreous carbon are studied upon neutron irradiation. Samples for study were two forms of vitreous carbon obtained by hardening thermally reactive polymers at 900-1,000 degree K. Phenol-formaldehyde (bakelite lacquer A, Bakelite A) and furfural-phenol-formaldehyde (FM-2) resin were used. They were irradiated in the experimental water - water VVR-M reactor between 360-1,030 degree K. The maximal neutron flux was 1.65·10 21 neut/cm 2 . Neutron irradiation of vitreous carbon led to its shrinkage and accompanied weakening. Shrinkage and weakening of vitreous carbon was decreased with an increase of treatment and irradiation temperatures

Atomic structure of radiation damages in FCC-metals after neutron irradiation

International Nuclear Information System (INIS)

Popova, E.V.; Ivchenko, V.A.; Kozlov, A.V.

2005-01-01

Full text: Radiation clusters, formed at a neutron irradiation, are a product of evolution of cascade areas. The quantitative information about clusters can be used for verification of calculations of cascade damage ability, in particular, cascade efficiency. Data about concentration clusters and an average of the vacancies containing in them, allow to receive total of the vacancies reserved in them and to use them for comparison to results of calculations of cascade damage ability. A correctness of such comparison by that above, than below temperature of a neutron irradiation. The purpose of work was experimental studying radiation clusters formed in FCC-metals at a low temperature neutron irradiation methods of dilatometry, field ion (FIM) and transmission electronic microscopy (TEM). Radiation clusters were studied: in industrial austenite steel C0.05Crl6Nil5Mo2Mnl, irradiated in reactor Rw-2a at temperature 310 K up to fluence intermediate and fast neutrons (with E > 0,1 MeV) 6.7·l0 21 m -2 ; in a modelling material - Pt (cleanliness of 99.99 %) with the same - FCC-structure in an initial condition and after an irradiation in reactor RWW-2M at temperature 310 K up to fluence intermediate and fast neutrons (with E > 0.1 MeV) 3.5·10 22 m -2 . As a result of an irradiation of steel and pure Pt, in these materials by methods FIM and TEM many radiation clusters, the accelerated neutrons initiated by interaction with substance was revealed. It is established that these damage areas represent the depleted zones containing separate vacancies, and also small vacancy complexes, with the 'belt' interstitial atoms. The quantitative estimation of the sizes of such radiating defects is lead and their density in volume is experimentally established. So the neutron irradiation of steel at temperature 310 K up to fluence 6.7·10 21 m -2 causes formation radiation clusters which average diameter according to TEM makes 3 nanometers. Observable by methods FIM clusters have the

Microstructural investigations of fast reactor irradiated austenitic and ferritic-martensitic stainless steel fuel cladding

International Nuclear Information System (INIS)

Agueev, V.S.; Medvedeva, E.A.; Mitrofanova, N.M.; Romanueev, V.V.; Tselishev, A.V.

1992-01-01

Electron microscopy has been used to characterize the microstructural changes induced in advanced fast reactor fuel claddings fabricated from Cr16Ni15Mo3NbB and Cr16Ni15Mo2Mn2TiVB austenitic stainless steels in the cold worked condition and Cr13Mo2NbVB ferritic -martensitic steel following irradiation in the BOR-60, BN-350 and BN-600 fast reactors. The data are compared with the results obtained from a typical austenitic commercial cladding material, Cr16Ni15Mo3Nb, in the cold worked condition. The results reveal a beneficial effect of boron and other alloying elements in reducing void swelling in 16Cr-15Ni type austenitic steels. The high resistance of ferritic-martensitic steels to void swelling has been confirmed in the Cr13Mo2NbVB steel. (author)

Recent status and improvement of reduced-activation ferritic-martensitic steels for high-temperature service

Energy Technology Data Exchange (ETDEWEB)

Tan, L., E-mail: tanl@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Tavassoli, A.-A.F.; Henry, J. [DMN/Dir, DEN, CEA Saclay, 91191, Gif-sur-Yvette Cedex (France); Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe, 76021 (Germany); Sakasegawa, H. [National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori, 039-3212 (Japan); Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Tanigawa, H. [National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori, 039-3212 (Japan); Huang, Q. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2016-10-15

Reduced-activation ferritic-martensitic (RAFM) steels, candidate structural materials for fusion reactors, have achieved technological maturity after about three decades of research and development. The recent status of a few developmental aspects of current RAFM steels, such as aging resistance, plate thickness effects, fracture toughness, and fatigue, is updated in this paper, together with ongoing efforts to develop next-generation RAFM steels for superior high-temperature performance. In addition to thermomechanical treatments, including nonstandard heat treatment, alloy chemistry refinements and modifications have demonstrated some improvements in high-temperature performance. Castable nanostructured alloys (CNAs) were developed by significantly increasing the amount of nanoscale MX (M = V/Ta/Ti, X = C/N) precipitates and reducing coarse M{sub 23}C{sub 6} (M = Cr). Preliminary results showed promising improvement in creep resistance and Charpy impact toughness. Limited low-dose neutron irradiation results for one of the CNAs and China low activation martensitic are presented and compared with data for F82H and Eurofer97 irradiated up to ∼70 displacements per atom at ∼300–325 °C.

Biological Effects of Neutron and Proton Irradiations. Vol. II. Proceedings of the Symposium on Biological Effects of Neutron Irradiations

International Nuclear Information System (INIS)

1964-01-01

During recent years the interest in biological effects caused by neutrons has been increasing steadily as a result of the rapid development of neutron technology and the great number of neutron sources being used. Neutrons, because of their specific physical characteristics and biological effects, form a special type of radiation hazard but, at the same time, are a prospective tool for applied radiobiology. This Symposium, held in Brookhaven at the invitation of the United States Government from 7-11 October 1963, provided an opportunity for scientists to discuss the experimental information at present available on the biological action of neutrons and to evaluate future possibilities. It was a sequel to the Symposium on Neutron Detection, Dosimetry and Standardization, which was organized by the International Atomic Energy Agency in December 1962 at Harwell. The Symposium was attended by 128 participants from 17 countries and 6 international organizations. Fifty-four papers were presented. The following subjects were discussed in various sessions: (1) Dosimetry. Estimation of absorbed dose of neutrons in biological material. (2) Biological effects of high-energy protons. (3) Cellular and genetic effects. (4) Pathology of neutron irradiation, including acute and chronic radiation syndromes (mortality, anatomical and histological changes, biochemical and metabolic disturbances) and delayed consequences. (5) Relative biological effectiveness of neutrons evaluated by different biological tests. A Panel on Biophysical Considerations in Neutron Experimentation, with special emphasis on informal discussions, was organized during the Symposium. The views of the Panel are recorded in Volume II of the Proceedings. Many reports were presented on the important subject of the relative effectiveness of the biological action of neutrons, as well as on the general pathology of neutron irradiation and the cellular and genetic effects related to it. Three survey papers considered

Activation of 45-MeV proton irradiation and proton-induced neutron irradiation in polymers

International Nuclear Information System (INIS)

Ra, Se-Jin; Kim, Kye-Ryung; Jung, Myung-Hwan; Yang, Tae-Keon

2010-01-01

During beam irradiation experiments with more than a few MeV energetic protons, the sample activation problem can be very severe because it causes many kinds of additional problems for the post-processing of the samples, such as time loss, inconvenience of sample handling, personal radiation safety, etc. The most serious problem is that immediate treatment of the sample is impossible in some experiments, such as nano-particle synthesizing. To solve these problems, we studied why the samples are activated and how the level of the activation can be reduced. It is known that the main reasons of activation are nuclear reactions with elements of the target material by primary protons and secondary produced neutrons. Even though the irradiation conditions are same, the level of the activation can be different depending on the target materials. For the nanoparticle synthesizing experiments, the target materials can be defined as the container and the sample itself. The reduction of the activation from the container is easier than the reduction from the sample. Therefore, we tried to reduce the activation level by changing the container materials. In this paper, the results are displayed for some candidate container materials, such as polymethyl methacrylate, polystyrene, Glass, etc., with 45-MeV and 10-nA proton beams. As a result, PS is the most suitable material for the container because of its relatively low level of the activation by protons. Also the contribution of secondary produced neutrons to the activation is negligible.

Study of damages by neutron irradiation in lithium aluminates

International Nuclear Information System (INIS)

Palacios G, O.

1999-01-01

Lithium aluminates proposed to the production of tritium in fusion nuclear reactors, due to the thermal stability that they present as well as the behavior of the aluminium to the irradiation. As a neutron flux with profile (≅ 14 Mev) of a fusion reactor is not available. A irradiation experiment was designed in order to know the micro and nano structure damages produced by fast and thermal neutrons in two irradiation positions of the fusion nuclear reactor Triga Mark III: CT (Thermal Column) and SIFCA (System of Irradiation Fixed of Capsules). In this work samples of lithium aluminate were characterized by XRD (X-Ray Diffraction), TEM (Transmission Electron Microscopy) and SEM (Scanning Electron Microscopy). Two samples were prepared by two methods: a) coalition method and b) peroxide method. This characterization comprised original and irradiated samples. The irradiated sample amounted to 4 in total: one for each preparation method and one for each irradiation position. The object of this analysis was to correlate with the received neutron dose the damages suffered by the samples with the neutron irradiation during long periods (440 H), in their micro and nano structure aspects; in order to understand the changes as a function of the irradiation zone (with thermal and fast neutron flux) and the preparation methods of the samples and having as an antecedent the irradiation in SIFCA position by short times (2h). The obtained results are referred to the stability of γ -aluminate phase, under given conditions of irradiation and defined nano structure arrangement. They also refer to the proposals of growth mechanism and nucleation of new phases. The error associated with the measurement of neutron dose is also discussed. (Author)

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

2008-01-01

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

2009-01-01

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

Effects of nickel and cobalt addition on creep strength and microstructure of the precipitation-strengthened 15Cr ferritic steels

Energy Technology Data Exchange (ETDEWEB)

Shibuya, Masachika; Toda, Yoshiaki; Sawada, Kota; Kushima, Hideaki; Kimura, Kazuhiro [National Inst. for Materials Science, Tsukuba, Ibaraki (Japan)

2010-07-01

Creep strength of 15Cr ferritic steel with ferrite matrix was increased by precipitation strengthening of intermetallic compounds. It was higher than those of 9-12Cr ferritic steels with a tempered martensitic microstructure strengthened by carbide and carbonitride. Addition of nickel was confirmed to improve Charpy impact toughness of the 15Cr steels, however, creep strength was slightly reduced by the addition of nickel. Microstructure of the 15Cr steel changes from ferrite single phase to dual phases of ferrite and martensite with the addition of nickel which is an austenite stabilizing element. The 15Cr steels investigated in the previous study, contain 3mass% of cobalt which is also an austenite stabilizing element, therefore, the influence of nickel and cobalt combination on mechanical properties and microstructure of the 15Cr-1Mo-6W-V-Nb steel is investigated in this study. Creep strength, Charpy impact toughness and microstructure of the steel were strongly influenced by the composition of nickel and cobalt. Design guideline of the 15Cr steel is discussed with respect to a role of microstructure and combination of nickel and cobalt addition. (orig.)

Neutron irradiation effects on the mechanical properties of organic composite materials

International Nuclear Information System (INIS)

Egusa, S.; Kirk, M.A.; Birtcher, R.C.

1984-01-01

Neutron irradiations with low γ-ray flux in the Intense Pulsed Neutron Source were carried out on four kinds of cloth-filled organic composites (filler: E-glass or carbon fiber; matrix: epoxy or polyimide resin) and a unidirectional alumina fiber/epoxy composite. These composites were examined with regard to the mechanical properties at room temperature. Following irradiation at room temperature, the Young's (tensile) modulus of these composites remains practically unchanged up to a total neutron fluence of 5.0x10 18 n/cm 2 (1.4x10 18 n/cm 2 for E>0.1 MeV). The shear modulus and the ultimate strength, on the other hand, decrease significantly at this neutron fluence for the glass/epoxy and glass/polyimide composites, whereas for the other composites both properties do not degrade. This result is most likely ascribed to the radiation damage at fiber/matrix interface due to recoil particles produced by a 10 B(n,α) 7 Li reaction in the boron-containing E-glass fibers. Only for the E-glass fiber composites, in fact, the fracture propagation energy is appreciably increased by irradiation, while for the other composites the propagation energy is scarcely changed, thus confirming the significant contribution due to the 10 B reaction. As to the 5 K irradiation, degradation of the present composites was not observed up to a total neutron fluence of 1.0x10 18 n/cm 2 (7.0x10 17 n/cm 2 for E>0.1 MeV) when tested at room temperature. (orig.)

Impurities effect on the swelling of neutron irradiated beryllium

International Nuclear Information System (INIS)

Donne, M.D.; Scaffidi-Argentina, F.

1995-01-01

An important factor controlling the swelling behaviour of fast neutron irradiated beryllium is the impurity content which can strongly affect both the surface tension and the creep strength of this material. Being the volume swelling of the old beryllium (early sixties) systematically higher than that of the more modem one (end of the seventies), a sensitivity analysis with the aid of the computer code ANFIBE (ANalysis of Fusion Irradiated BEryllium) to investigate the effect of these material properties on the swelling behaviour of neutron irradiated beryllium has been performed. Two sets of experimental data have been selected: the first one named Western refers to quite recently produced Western beryllium, whilst the second one, named Russian refers to relatively old (early sixties) Russian beryllium containing a higher impurity rate than the Western one. The results obtained with the ANFIBE Code were assessed by comparison with experimental data and the used material properties were compared with the data available in the literature. Good agreement between calculated and measured values has been found

Neutron irradiated uranium silicides studied by neutron diffraction and Rietveld analysis

International Nuclear Information System (INIS)

Birtcher, R.C.; Mueller, M.H.; Richardson, J.W. Jr.; Faber, J. Jr.

1989-11-01

Uranium silicides have been considered for use as reactor fuels in both high power and low enrichment applications. However, U 3 Si was found to become amorphous under irradiation and to become mechanically unstable to rapid growth by plastic flow. U 2 Si 2 appears to be stable against amorphization at low displacement rates, but the extent of this stability is uncertain. Although the mechanisms responsible for plastic flow in U 3 Si and other amorphous systems are unknown, as is the importance of crystal structure for amorphization, it may not be surprising that these materials amorphize, in light of the fact that many radioactive nuclide - containing minerals are known to metaminctize (lose crystallinity) under irradiation. The present experiment follows the detailed changes in the crystal structures of U 3 Si and U 3 Si 2 introduced by neutron bombardment and subsequent uranium fission at room temperature. U-Si seems the ideal system for a neutron diffraction investigation since the crystallographic and amorphous forms can be studied simultaneously by combining conventional Rietveld refinement of the crystallographic phases with Fourier-filtering of the non-crystalline scattering component

Deuterium Depth Profile in Neutron-Irradiated Tungsten Exposed to Plasma

International Nuclear Information System (INIS)

Shimada, Masashi; Cao, G.; Hatano, Y.; Oda, T.; Oya, Y.; Hara, M.; Calderoni, P.

2011-01-01

The effect of radiation damage has been mainly simulated using high-energy ion bombardment. The ions, however, are limited in range to only a few microns into the surface. Hence, some uncertainty remains about the increase of trapping at radiation damage produced by 14 MeV fusion neutrons, which penetrate much farther into the bulk material. With the Japan-US joint research project: Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), the tungsten samples (99.99 % pure from A.L.M.T., 6mm in diameter, 0.2mm in thickness) were irradiated to high flux neutrons at 50 C and to 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). Subsequently, the neutron-irradiated tungsten samples were exposed to a high-flux deuterium plasma (ion flux: 1021-1022 m-2s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment (TPE) at the Idaho National Laboratory (INL). First results of deuterium retention in neutron-irradiated tungsten exposed in TPE have been reported previously. This paper presents the latest results in our on-going work of deuterium depth profiling in neutron-irradiated tungsten via nuclear reaction analysis. The experimental data is compared with the result from non neutron-irradiated tungsten, and is analyzed with the Tritium Migration Analysis Program (TMAP) to elucidate the hydrogen isotope behavior such as retention and depth distribution in neutron-irradiated and non neutron-irradiated tungsten.

A neutron irradiator to perform nuclear activation

International Nuclear Information System (INIS)

Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

2001-01-01

The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

Influence of neutron irradiation on etching of SiC in KOH

Science.gov (United States)

Mokhov, E. N.; Kazarova, O. P.; Soltamov, V. A.; Nagalyuk, S. S.

2017-07-01

The effect of reactor neutron irradiation on the etch rate of SiC in potassium hydroxide has been studied. In the case of high irradiation doses (1019-1021 cm-2), the etch rate of silicon carbide has been shown to drastically rise, especially in the [0001]Si direction. This considerably mitigates the orientation anisotropy of polar face etching. After high-temperature annealing (up to 1200-1400°C), a higher etch rate of irradiated crystals persists. The results have been explained by the high concentration of radiation-induced (partially clustered) defects they contain.

Neutron irradiation effects in pressure vessel steels and weldments

Energy Technology Data Exchange (ETDEWEB)

Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

1994-12-31

This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

Damages to gladiolu corm caused by fast neutron irradiation

International Nuclear Information System (INIS)

Zhang Zhiwei; Wang Dan; Zhang Dongxue; Zheng Chun

2007-01-01

Gladiolus corms were irradiated to 100-500kGy by fast neutrons in the CFBR-II pulsed reactor, Scanning electron microscope images of the irradiated samples revealed significant radiation damages to the gladiolus corms, and the mutagenic effects were studied by SDS-polyacrylamide gel electrophoresis (SDS-PAGE). Within the dose range, radiation damage to the corm increased with the dose, with corm epidermis of the samples irradiated in vertical incidence being more serious than those irradiated in side-incidence to the same dose. Biological characters were investigated via field experiments, and the bands of protein subunit were analyzed by SDS-PAGE. The results showed that the fast neutrons irradiation inhibited growth of M1 generation seedling significantly. Protein expression was obviously inhibited by the irradiation. The study indicates that fast neutron induction is an effective way for gladiolus breeding. And the results may lay a foundation for studies on fast neutron mutation breeding. (authors)

A study by internal friction of defects produced in iron and nickel irradiated at very low temperatures; Etude au moyen du frottement interne des defauts crees par irradiation a tres basse temperature dans le fer et le nickel

Energy Technology Data Exchange (ETDEWEB)

Keating-Hart, G de [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1969-07-01

This work represents a contribution to the study of point defects in metals. After a brief review of the theory of internal friction we will discuss some technical innovations aimed at increasing the flexibility of the apparatus at our disposal. These innovations have allowed us to extend our range of measurements down to 20 deg. K. We will then discuss our experimental results. Firstly, we describe preliminary experiments on electron irradiated iron and the evidence for a laminar structure. Secondly, we present and account of the first measurements on nickel after neutron irradiation at 27 deg. K. We will compare the results with those obtained by other methods in this laboratory. Essentially we have observed transitory peaks at low temperature due to close Frenkel pairs and we have noticed the absence of a peak which would correspond to the magnetic after effect band of stage I{sub E}. An attempt is made to explain the disappearance of the observed peaks upon the application of an internal saturating magnetic field. (author) [French] Ce memoire constitue une contribution a l'etude des defauts ponctuels dans les metaux. Apres un bref apercu theorique sur le frottement interne, nous presenterons quelques realisations techniques destinees a accroitre les possibilites des instruments qui nous ont ete confies. Ces dernieres nous ont permis d'etendre la gamme des mesures jusqu'a 20 deg. K. Nous parlerons ensuite de nos resultats experimentaux. En premier lieu, ceux obtenus au cours de premieres experiences, sur le fer irradie aux electrons mettent en evidence des structures de laminage. En second lieu, nous exposerons les premieres mesures realisees sur du nickel irradie aux neutrons; nous comparerons ces resultats avec ceux obtenus par d'autres moyens experimentaux dans le laboratoire. Nous avons observe essentiellement des pics fugitifs a basse temperature dus aux paires proches de Frenckel et nous avons constate l'absence d'un pic correspondant a la bande de tramage

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

2015-03-15

Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

Science.gov (United States)

Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

2013-01-01

Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

EPR of alanine irradiated by neutrons

International Nuclear Information System (INIS)

Pivovarov, S.P.; Seredavina, T.A.; Zhdanov, S.V.; Mul'gin, S.I.; Zhakparov, R.K.

2001-01-01

In the work the first results of EPR studies of alanine, irradiated with diverse doses at neutron cyclotron generator different conditions and on the critical reactor stand are presented. A dose linearity dependence of EPR signal is observing, the methods of γ-background contribution separation are discussed. Obtain results is giving the basis to recommendation of alanine as an effective detector irradiation. However it is demanded the farther study on clarification of radiation sensitivity value dependence on the neutron energy spectrum form

Certification of a nickel metal reference material for neutron dosimetry (EC Nuclear Reference Material 521)

International Nuclear Information System (INIS)

Pauwels, J.

1988-01-01

Nickel metal, of 99.99 % nominal purity and natural isotopic composition, in the form of 0.1 mm thick foil and 0.5 mm diameter wire has been certified for its cobalt mass fraction. The certified value of cobalt (<0.1μg.g-1) is based on 38 results obtained by neutron activation analysis, emission spectrometry with inductively coupled plasma excitation and atomic absorption spectrometry, whereas the isotopic composition of the nickel was verified by thermal ionization mass spectrometry. The material is intended to be used as a reference material in neutron metrology

Swelling of structural materials in fast neutron reactors

International Nuclear Information System (INIS)

Seran, J.L.

1983-06-01

The physical origin of swelling in irradiated materials and the main parameters acting on swelling of SS 316 are examined: temperature, neutron dose, dose rate, chemical composition, strain hardening. Results obtained, in Rapsodie and Phenix reactors, with fuel cans and with the hexagonal tube containing the fuel pins are analyzed and compared with results found in litterature. In conclusion hot swelling of SS 316 is too important at high doses and is will be replaced by austenitic steels stabilized by Ti and ferritic steels or high nickel steels with structural hardening [fr

Convoluted dislocation loops induced by helium irradiation in reduced-activation martensitic steel and their impact on mechanical properties

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada K7L 3N6 (Canada); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Wen, Yongming [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China)

2014-06-01

Helium irradiation induced dislocation loops in reduced-activation martensitic steels were investigated using transmission electron microscopy. The specimens were irradiated with 100 keV helium ions to 0.8 dpa at 350 °C. Unexpectedly, very large dislocation loops were found, significantly larger than that induced by other types of irradiations under the same dose. Moreover, the large loops were convoluted and formed interesting flower-like shape. The large loops were determined as interstitial type. Loops with the Burgers vectors of b=〈100〉 were only observed. Furthermore, irradiation induced hardening caused by these large loops was observed using the nano-indentation technique.

Use of L-cysteine for minimization of inorganic Hg loss during thermal neutron irradiation

International Nuclear Information System (INIS)

Anderson, D.L.

2009-01-01

Thermal neutron irradiation experiments performed with cellulose-based L-cysteine-treated and untreated Hg standards showed Hg losses of 59-81% for untreated standards but only about a 0.2% loss for treated standards. These results and others for multielement standards showed that Hg loss is highly dependent on total mass and placement of materials in the irradiation vessel and that distribution of volatilized Hg was fairly uniform throughout the sample-containing region of the vessel. Polyethylene trapped volatile Hg much more efficiently than cellulose and a multielement standard containing inorganic Se selectively trapped Hg lost from a co-irradiated multielement standard containing Hg. (author)

Data analysis of a polycrystalline nickel sample obtained with neutron diffraction

International Nuclear Information System (INIS)

Parente, C.B.R.; Mazzochi, V.L.; Araujo Kaschny, J.R. de; Costa, M.S. da; Rizzo, P.; Campos, W.M.S.

1990-01-01

A simple analysis of the nickel structure was made. Neutron diffraction data were used, obtained with polycrystalline nickel placed in a cylindrical sample-holder with dimensions of 1,5cm of diameter and 5cm of height. The theoretical intensities were calculated of 3 forms: 1. without considering the temperature and obsorption effects, 2. considering only the temperature effect and 3. considering both the temperature and absorption effects. The disagreement factors found in this 3 cases were, respectively, R 1 = 13.4%, R 2 = 7,7% e R 3 = 7,5%. (L.C.)

F-type centers in neutron-irradiated AIN

International Nuclear Information System (INIS)

Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru; Okada, Moritami; Nakagawa, Masuo.

1990-01-01

The production of point defects by neutron irradiation and thermal decay in sintered AIN polycrystal are investigated. The absorption band at 370 nm is observed after reactor neutron irradiation to a dose of 10 16 n/cm 2 (E > 0.1 MeV). The defect corresponding to the band is tentatively assigned as an F-type center from the optical absorption and electron spin resonance. (author)

Antiradiation Vaccine: Technology Development Of Prophylaxis, Prevention And Treatment Of Biological Consequences And Complications After Neutron Irradiation.

Science.gov (United States)

Popov, Dmitri; Maliev, Slava; Jones, Jeffrey

irradiation generated in a canal of Research Reactor BBP-M and BBP-M. Mixed neutron beam contained 95% of fast neutron irradiation and 5% of gamma-irradiation. Neutron energy - 1.98 - 2.30 Me V energy. Dose - 10.7 Gy., 0.22 Gy-min. Scheme of experiments: Rabbits from all groups were irradiated in a canal of Research Reactor together. Group A: control-5 rabbits; Group B:placebo-5 rabbits; Group C: radioprotectant Cystamine (50 mg-kg)-5 rabbits, 15 minutes before irradiation Group D:Radio-protectant Mexamine (10 mg-kg)-5 rabbits { 15 minutes before irradiation; Group E: Antiradiation Vaccine: subcutaneus administration or I-M - 2 ml of active substance , 20 days before irradiation. Results: Control Group A - 100% mortality within the next two hours after neutron irradiation with clinical symptoms of acute cerebrovascular syndrome. Group B - 100% mortality less than two hours following irradiation. Group C - 100% mortality within 8-10 hours after irradiation. Group D - 100% mortality within 8-11 hours after irradiation. In Groups A - D the development of extremely severe form of Acute Radiation Cerebrovascular Syndrome produced rapid death. Group E - 100% mortality within 240 hours ( 9|10 days) following neutron irradiation with animals exhibiting cardiovascular, cerebrovascular and gastrointestinal clinical symptoms. Discussion: A pre-irradiation vaccination with Antiradiation Vaccine is effective against mild and even high doses of neutron radiation. Vaccination with antiradiation Vaccine prolonged survival time of rabbits, exposed to a high dose LD100, of neutron radiation: from two hours (control) up to 11 days. We also postulate that radiation toxins,isolated from lymph of gamma-irradiated animals are likely similar to structure of radiation toxins circulated in blood and lymph of neutron irradiated animals. Toxico-kinetics and toxico-dynamics of radiation toxins of after neutron-irradiation were quite unique and distinguished from different types of radiation

Neutron and gamma irradiation damage to organic materials.

Energy Technology Data Exchange (ETDEWEB)

White, Gregory Von, II; Bernstein, Robert

2012-04-01

This document discusses open literature reports which investigate the damage effects of neutron and gamma irradiation on polymers and/or epoxies - damage refers to reduced physical chemical, and electrical properties. Based on the literature, correlations are made for an SNL developed epoxy (Epon 828-1031/DDS) with an expected total fast-neutron fluence of {approx}10{sup 12} n/cm{sup 2} and a {gamma} dosage of {approx}500 Gy received over {approx}30 years at < 200 C. In short, there are no gamma and neutron irradiation concerns for Epon 828-1031/DDS. To enhance the fidelity of our hypotheses, in regards to radiation damage, we propose future work consisting of simultaneous thermal/irradiation (neutron and gamma) experiments that will help elucidate any damage concerns at these specified environmental conditions.

Mechanical and wear properties of pre-alloyed molybdenum P/M steels with nickel addition

Directory of Open Access Journals (Sweden)

Yamanoglu R.

2012-01-01

Full Text Available The aim of this study is to understand the effect of nickel addition on mechanical and wear properties of molybdenum and copper alloyed P/M steel. Specimens with three different nickel contents were pressed under 400 MPa and sintered at 1120ºC for 30 minutes then rapidly cooled. Microstructures and mechanical properties (bending strength, hardness and wear properties of the sintered specimens were investigated in detail. Metallographical investigations showed that the microstructures of consolidated specimens consist of tempered martensite, bainite, retained austenite and pores. It is also reported that the amount of pores varies depending on the nickel concentration of the alloys. Hardness of the alloys increases with increasing nickel content. Specimens containing 2% nickel showed minimum pore quantity and maximum wear resistance. The wear mechanism changed from abrasive wear at low nickel content to adhesive wear at higher nickel content.

Solidification phenomena in nickel base brazes containing boron and silicon

International Nuclear Information System (INIS)

Tung, S.K.; Lim, L.C.; Lai, M.O.

1996-01-01

Nickel base brazes containing boron and/or silicon as melting point depressants are used extensively in the repair and joining of aero-engine hot-section components. These melting point depressants form hard and brittle intermetallic compounds with nickel which are detrimental to the mechanical properties of brazed joints. The present investigation studied the microstructural evolution in nickel base brazes containing boron and/or silicon as melting point depressant(s) in simple systems using nickel as the base metal. The basic metallurgical reactions and formation of intermetallic compounds uncovered in these systems will be useful as a guide in predicting the evolution of microstructures in similar brazes in more complex systems involving base metals of nickel base superalloys. The four filler metal systems investigated in this study are: Ni-Cr-Si; Ni-Cr-B; Ni-Si-B and Ni-Cr-Fe-Si-B

Influence of neutron irradiation on ferromagnetic metallic glasses

International Nuclear Information System (INIS)

Miglierini, M.; Nasu, Saburo; Sitek, J.

1992-01-01

Transmission 57 Fe Moessbauer spectroscopy is used to study effects of neutron irradiation on magnetic properties of Fe-based ferromagnetic metallic glasses. Elastic stress centers are produced during the process of neutron irradiation as a result of atom mixing. Rearrangement of the atoms causes changes in the average value of the hyperfine field distribution and orientation of the net magnetic moment. They are shown to depend on the composition of the investigated samples. Cr-doped metallic glasses depict transformation from ferromagnetic to paramagnetic state at room temperature after neutron irradiation implying changes in the Curie temperature. Presence of Ni in the samples reduces the effects of radiation damage. (orig.)

Experimental investigation of high He/dpa microstructural effects in neutron irradiated B-alloyed Eurofer97 steel by means of small angle neutron scattering (SANS and electron microscopy

Directory of Open Access Journals (Sweden)

R. Coppola

2016-12-01

Full Text Available High He/dpa microstructural effects have been investigated, by means of small-angle neutron scattering (SANS and transmission electron microscopy (TEM, in B-alloyed ferritic/martensitic steel Eurofer97-1 (0.12 C, 9 Cr, 0.2V, 1.08W wt%, B contents variable between 10 and 1000ppm, neutron irradiated at the High Flux Reactor of the JRC-Petten at temperatures between 250 °C and 450 °C, up do a dose level of 16 dpa. Under these irradiation parameters, B activation is expected to produce corresponding helium contents variable between 80 and 5600appm, with helium bubble distributions relevant for the technological applications. The SANS measurements were carried out under magnetic field to separate nuclear and magnetic SANS components; a reference, un-irradiated sample was also measured to evaluate as accurately as possible the genuine effect of the irradiation on the microstructure. Increasing the estimated helium content from 400 to 5600appm, the analysis of the SANS cross-sections yields an increase in the volume fraction, attributed to helium bubbles, of almost one order of magnitude (from 0.007 to 0.038; furthermore, the difference between nuclear and magnetic SANS components is strongly reduced. These results are discussed in correlation with TEM observations of the same samples and are tentatively attributed to the effect of drastic microstructural changes in Eurofer97-1 for high He/dpa ratio values, possibly relating to the dissolution of large B-carbides due to transmutation reactions.

Neutron Focusing Mirrors for Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

Science.gov (United States)

Rai, Durgesh K.; Wu, Huarui; Abir, Muhammad; Giglio, Jeffrey; Khaykovich, Boris

Post irradiation examination (PIE) of samples irradiated in nuclear reactors is a challenging but necessary task for the development on novel nuclear power reactors. Idaho National Laboratory (INL) has neutron radiography capabilities, which are especially useful for the PIE of irradiated nuclear fuel. These capabilities are limited due to the extremely high gamma-ray radiation from the irradiated fuel, which precludes the use of standard digital detectors, in turn limiting the ability to do tomography and driving the cost of the measurements. In addition, the small 250 kW Neutron Radiography Reactor (NRAD) provides a relatively weak neutron flux, which leads to low signal-to-noise ratio. In this work, we develop neutron focusing optics suitable for the installation at NRAD. The optics would separate the sample and the detector, potentially allowing for the use of digital radiography detectors, and would provide significant intensity enhancement as well. The optics consist of several coaxial nested Wolter mirrors and is suited for polychromatic thermal neutron radiation. Laboratory Directed Research and Development program of Idaho National Laboratory.

Effects of irradiation and thermal aging upon fatigue-crack growth behavior of reactor pressure boundary materials. [Neutrons

Energy Technology Data Exchange (ETDEWEB)

James, L. A.

1978-10-01

Two processes that have the potential to produce degradation in the properties of pressure boundary materials are neutron irradiation and long-time thermal aging. This paper uses linear-elastic fracture mechanics techniques to assess the effect of these two processes upon the fatigue-crack growth behavior of a number of alloys commonly employed in reactor pressure boundaries. The materials evaluated include ferritic steels, austenitic stainless steels, and nickel-base alloys typical of those employed in a number of reactor types including water-cooled, gas-cooled, and liquid-metal-cooled designs.

Determination of platinum group elements and gold in reference materials by instrumental neutron activation analysis and inductively coupled plasma-mass spectrometry with nickel sulphide fire-assay collection

International Nuclear Information System (INIS)

Morcelli, C.P.R.; Figueiredo, A.M.G.; Sarkis, J.E.S.; Kakazu, M.; Enzweiler, J.

2002-01-01

Instrumental Neutron Activation Analysis (INAA) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) after NiS fire assay were used to determine platinum group elements (PGE) and Au in the geological reference materials peridotite GPt-3 and pyroxene peridotite GPt-4 (IGGE, China). INAA has been one of the most useful analytical techniques for PGEs and Au determination, due to its high sensitivity and accuracy. After the fire assay, the NiS button is dissolved in concentrated hydrochloric acid, leaving a residue of insoluble PGE sulphides; the solution is filtered and the filter is directly irradiated with neutrons. In more recent years, ICP-MS with nickel fire assay collection and tellurium coprecipitation has been used as a mean of analysing PGE, with the main advantage of avoiding problems of losses of PGE during the HCl digestion of the NiS button. Buttons were prepared by mixing the sample (10-15 g) with fluxes, nickel and sulphur in a fire clay crucible and fused at temperatures around 1000 deg C. For NAA, the filters containing the PGEs and Au were irradiated at the IEA-R1 research nuclear reactor at IPEN. The measurements of the induced gamma-ray activity were carried out in an hyperpure Ge detector. In general, the results obtained by the two techniques were in good agreement with recommended values. INAA results exhibited higher values than the recommended values for Pd and Pt in GPt-3, while the opposite effect was observed for GPt-4 sample. Ru was not detected by INAA. On the other hand, Rh and Ir were determined more accurately by INAA (relative errors better than 10%). The ICP-MS analytical technique showed better detection limits, and all the PGE were determined. Results obtained for Pt and Pd presented accuracy better than 5% while losses were observed for Os and Ir. (author)

Neutron Absorbing Ability Variation in Neutron Absorbing Material Caused by the Neutron Irradiation in Spent Fuel Storage Facility

Energy Technology Data Exchange (ETDEWEB)

Sohn, Hee Dong; Han, Seul Gi; Lee, Sang Dong; Kim, Ki Hong; Ryu, Eag Hyang; Park, Hwa Gyu [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

2014-10-15

In spent fuel storage facility like high density spent fuel storage racks and dry storage casks, spent fuels are stored with neutron absorbing materials installed as a part of those facilities, and they are used for absorbing neutrons emitted from spent fuels. Usually structural material with neutron absorbing material of racks and casks are located around spent fuels, so it is irradiated by neutrons for long time. Neutron absorbing ability could be changed by the variation of nuclide composition in neutron absorbing material caused by the irradiation of neutrons. So, neutron absorbing materials are continuously faced with spent fuels with boric acid solution or inert gas environment. Major nuclides in neutron absorbing material are Al{sup 27}, C{sup 12}, B{sup 11}, B{sup 10} and they are changed to numerous other ones as radioactive decay or neutron absorption reaction. The B{sup 10} content in neutron absorbing material dominates the neutron absorbing ability, so, the variation of nuclide composition including the decrease of B{sup 10} content is the critical factor on neutron absorbing ability. In this study, neutron flux in spent fuel, the activation of neutron absorbing material and the variation of nuclide composition are calculated. And, the minimum neutron flux causing the decrease of B{sup 10} content is calculated in spent fuel storage facility. Finally, the variation of neutron multiplication factor is identified according to the one of B{sup 10} content in neutron absorbing material. The minimum neutron flux to impact the neutron absorbing ability is 10{sup 10} order, however, usual neutron flux from spent fuel is 10{sup 8} order. Therefore, even though neutron absorbing material is irradiated for over 40 years, B{sup 10} content is little decreased, so, initial neutron absorbing ability could be kept continuously.

Irradiation Effects in Fortiweld Steel Containing Different Boron Isotopes

International Nuclear Information System (INIS)

Grounes, M.

1967-07-01

Tensile specimens and miniature impact specimens of the low alloyed pressure vessel steel Fortiweld have been irradiated at 265 deg C in R2 to two neutron doses, 6.5 x 10 18 n/cm 2 (> 1 MeV) and 4 x 10 19 n/cm 2 (thermal) and also 9.0 x 10 18 n/cm 2 (> 1 MeV) and 6 x 10 19 n/cm 2 (thermal). Material from three laboratory melts, in which the boron consisted of 10 B, 11 B and natural boron respectively, were investigated. The results both of tensile tests and impact tests with miniature impact specimens show that the 10 B-alloyed material was changed more and the 11 B-alloyed material was changed less than the material containing natural boron. At the higher neutron dose the increase in yield strength (0.2 % offset yield strength) was 11 kg/mm in the 10 B containing material compared to 5 kg/mm in the 11 B-containing material. The decrease in total elongation was 5 and 0 percentage units respectively. The transition temperature was increased 190 deg C at the higher neutron dose in the 10 B-alloyed material, 40 deg C in the 11 B-alloyed material and 80 deg C in the material containing natural boron

Compaction and sintering of nickel powder used encapsulation of irradiation targets

Energy Technology Data Exchange (ETDEWEB)

Miyano, Rosana S.L.; Guimaraes, Raquel R.F.L.; Rossi, Jesualdo L., E-mail: rosatac@gmail.com, E-mail: raquel.lucchesi@icloud.com, E-mail: jelrossi@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (CCTM/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Ciencia e Tecnologia de Materiais; Wendhausen, Paulo A.P.; Evangelista, Leandro L., E-mail: paulo.wendhausen@ufsc.br, E-mail: leandro.materiais@gmail.com [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Laboratorio de Materiais

2015-07-01

The objective of this study was to develop an alternative way to produce targets for irradiation containing uranium, for the pair of {sup 99}Mo production {sup 99m}Tc radionuclide. These targets were obtained by powder metallurgy, the compact serving as means for encapsulation a uranium cylinder to be irradiated. The targets were compacted in an axial hydraulic press applying different pressures up to 800 MPa. The sintering temperature was 600 °C in hydrogen atmosphere and it was used two sintering cycles, one for 4 h and the for 4 h plus 8 h time. The nickel powder was of high purity, that in order to provide the sealing of the fissile content within the compacted. The bulk density of compacted was evaluated by the method geometric. The porosity was measured by mercury porosimetry technique. The microstructure was investigated by optical microscopy. The results obtained with sintering powders involving confirm the feasibility of achieving a casing for uranium targets. (author)

Compaction and sintering of nickel powder used encapsulation of irradiation targets

International Nuclear Information System (INIS)

Miyano, Rosana S.L.; Guimaraes, Raquel R.F.L.; Rossi, Jesualdo L.; Wendhausen, Paulo A.P.; Evangelista, Leandro L.

2015-01-01

The objective of this study was to develop an alternative way to produce targets for irradiation containing uranium, for the pair of 99 Mo production 99m Tc radionuclide. These targets were obtained by powder metallurgy, the compact serving as means for encapsulation a uranium cylinder to be irradiated. The targets were compacted in an axial hydraulic press applying different pressures up to 800 MPa. The sintering temperature was 600 °C in hydrogen atmosphere and it was used two sintering cycles, one for 4 h and the for 4 h plus 8 h time. The nickel powder was of high purity, that in order to provide the sealing of the fissile content within the compacted. The bulk density of compacted was evaluated by the method geometric. The porosity was measured by mercury porosimetry technique. The microstructure was investigated by optical microscopy. The results obtained with sintering powders involving confirm the feasibility of achieving a casing for uranium targets. (author)

A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2000-01-01

A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film

International Nuclear Information System (INIS)

Kalsi, P.C.

2000-01-01

The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)

Neutron irradiation test of copper alloy/stainless steel joint materials

International Nuclear Information System (INIS)

Yamada, Hirokazu; Kawamura, Hiroshi

2006-01-01

As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

Energy Technology Data Exchange (ETDEWEB)

Fernandez, P.; Lapena, J.; Blazquez, F. [Ciemat, Madrid (Spain)

2000-07-01

Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigree. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs.

Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

International Nuclear Information System (INIS)

Fernandez, P.; Lapena, J.; Blazquez, F.

2000-01-01

Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigrade. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs

Neutron irradiation effect of thermally-sensitized stainless steels

Energy Technology Data Exchange (ETDEWEB)

Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

1998-03-01

Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

Energy Technology Data Exchange (ETDEWEB)

Fukuda, Makoto, E-mail: makoto.fukuda@qse.tohoku.ac.jp [Tohoku University, Sendai, 980-8579 (Japan); Kiran Kumar, N.A.P.; Koyanagi, Takaaki; Garrison, Lauren M. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA, 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Hasegawa, Akira [Tohoku University, Sendai, 980-8579 (Japan)

2016-10-15

Neutron irradiation to single crystal pure tungsten was performed in the mixed spectrum High Flux Isotope Reactor (HFIR). To investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ∼90–∼800 °C and fast neutron fluences were 0.02–9.00 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. The hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten. - Highlights: • The microstructure and irradiation hardening of single crystal pure W irradiated in HFIR was investigated. • The neutron energy spectrum influence was evaluated by comparing the HFIR results with previous work in Joyo and JMTR. • In the dose range up to ∼1 dpa, the neutron energy spectrum influence of irradiation hardening was not clear. • In the dose range above 1 dpa, the neutron energy influence on irradiation hardening and microstructural development was clearly observed. • The irradiation induced precipitates caused significant irradiation hardening of pure W irradiated in HFIR.

Micro-Raman and photoluminescence studies of neutron-irradiated gallium nitride epilayers

International Nuclear Information System (INIS)

Wang, R.X.; Xu, S.J.; Fung, S.; Beling, C.D.; Wang, K.; Li, S.; Wei, Z.F.; Zhou, T.J.; Zhang, J.D.; Huang Ying; Gong, M.

2005-01-01

GaN epilayers grown on sapphire substrate were irradiated with various dosages of neutrons and were characterized using Micro-Raman and photoluminescence. It was found that the A 1 (LO) peak in the Raman spectra clearly shifted with neutron irradiation dosage. Careful curve fitting of the Raman data was carried out to obtain the carrier concentration which was found to vary with the neutron irradiation dosage. The variation of the full width at half maximum height of the photoluminescence was consistent with the Raman results. The neutron irradiation-induced structural defects (likely to be Ge Ga ) give rise to carrier trap centers which are responsible for the observed reduction in carrier concentration of the irradiated GaN

Evaluation of neutron irradiation fields for BNCT by using absorbed dose in a phantom

International Nuclear Information System (INIS)

Aizawa, O.

1993-01-01

In a previous paper, the author defined the open-quotes irradiation timeclose quotes as the time of irradiation in which the maximum open-quotes total background doseclose quotes becomes 2,500 RBE-cGy. In this paper, he has modified the definition a little as the time of irradiation in which the maximum open-quotes lμg/g B-10 doseclose quotes becomes 3,000 RBE-cGy, because he assumed that normal tissue contained 1μg/g B-10. Moreover, he has modified the dose criteria for BNCT as follows: The open-quotes eye doseclose quotes, open-quotes total body doseclose quotes and open-quotes except-head doseclose quotes should be less that 200, 100 and 50 RBE-cGy, respectively. He has added one more criterion for BNCT that the thermal neutron fluence at the tumor position should be over 2.5x10 12 n/cm 2 at the open-quotes irradiation timeclose quotes. The distance from the core side to the irradiation port in the open-quotes old configurationclose quotes of the Musashi reactor (TRIGA-II, 100kW) was 160 cm. He is now planning to design an eccentric core and to move the reactor core nearer to the irradiation port, distance between the core side and the irradiation port to be 140, 130 and 120cm. The other assumptions used in this paper are as follows: (1) The B-10 concentrations in tumor are 30 and/or 10μg/g. (2) The depth of the tumor is 5.0 cm to 5.5 cm from the surface. (3) The RBE values used are 1.0 for all gamma rays and 2.3 for B 10 (n,α) reaction products. (4) The RBE values for neutrons are the following three cases: the first case is using 1.6 for all neutrons; the second one is using 3.2 for non-thermal neutrons and 1.6 for thermal neutrons; the third case is using 4.8 for fast neutrons, 3.2 for faster epithermal and epithermal neutrons, and 1.6 for thermal neutrons

Irradiation temperature dependence of defect formation of nitrides (A1N and c-BN) during neutron irradiations

International Nuclear Information System (INIS)

Atobe, Kozo.; Okada, Moritami; Nakagawa, Masuo

2000-01-01

The nitrogen vacancy concentration in the more refractory nitrides (A1N and c-BN) is determined as a function of reactor fluence up to 5.2x10 17 thermal neutrons/cm 2 and a function of the irradiation temperature at 25, 50, 100, 150, 200, 250 K. It is found that there is no remarkable dependence of the defect formation in nitrides on the irradiation temperature. The production of damage in the nitrides is considerably different from that in oxides. From the irradiation experiments using thermal neutron irradiation field, it is suggested in reactor irradiation that the atomic displacements in the nitrides occur predominately from energetic particles of the nuclear reactions with thermal neutrons in addition to the elastic collisions by fast neutron

Positron lifetime study of copper irradiated by energetic protons or energetic neutrons

International Nuclear Information System (INIS)

Howell, R.H.

1979-03-01

Positron lifetime measurements of pure copper damaged by irradiation with energetic protons and neutrons are presented. Lifetime determinations of the bulk material and various traps were made, and the dependence of the trapping rate on dose and irradiation energy were investigated. The results from the neutron- and proton-irradiated samples point to the existence of traps with similar but distinct lifetime parameters, not varying greatly from values reported in deformation studies. Also, a trap with long lifetime is seen for some proton irradiations, but is never seen for the neutron irradiations. The trapping rate of the short-lifetime trap is a linear function of dose for proton-irradiated samples and nearly so for the neutron irradiation. 1 figure

40 CFR 721.5330 - Nickel salt of an organo compound containing nitrogen.

Science.gov (United States)

2010-07-01

... containing nitrogen. 721.5330 Section 721.5330 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Specific Chemical Substances § 721.5330 Nickel salt of an organo compound containing nitrogen. (a) Chemical... as nickel salt of an organo compound containing nitrogen (PMN P-92-686) is subject to reporting under...

Specific Heat Capacity of Alloy 690 for Simulating Neutron Irradiation

International Nuclear Information System (INIS)

Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Joo, Young Sun; Ahn, Sang Bok; Park, Jin Seok; Lee, Won Jae; Ryu, Woo Seok

2011-01-01

The KAERI(Korea Atomic Energy Research Institute) is developing new type of nuclear reactor, so called 'SMART'(System Integrated Modular Advanced Reactor) which has many features of small power and system integrated modular type. Alloy 690 was selected as the candidate material for the heat exchanger tube of the steam generator of SMART. The SMART R and D is now facing the stage of engineering verification and approval of standard design to apply to DEMO reactors. Therefore, the material performance under the relevant environment is required to be evaluated. The important material performance issues are mechanical properties i.e. (fracture toughness, tensile and hardness) and thermal properties i.e. (thermal diffusivity, specific heat capacity and thermal conductivity) for which the engineering database is necessary to design a steam generator. However, the neutron post irradiation characteristics of the alloy 690 are barely known. As a result, PIE(Post Irradiation Examination) of thermal properties are planed and performed successfully. But specific heat capacity measurement is not performed because of not having proper test system for irradiated materials. Therefore in order to verify the effect of neutron irradiation for alloy 690, simulation method is adopted. In general, high energy neutron bombardment in material bring about lattice defects i.e. void, pore and dislocation. Dominant factor to impact to heat capacity is mainly dislocation in material. Therefore, simulation of neutron irradiation is devised by material rolling method in order to make artificial dislocation in alloy 690 as same effect of neutron irradiation. After preparing test specimens, heat capacity measurements are performed and results are compared with rolled materials and un-rolled materials to verify the effect of neutron irradiation simulation. Main interest of simulation is that heat capacity value is changed by neutron irradiation

Influence of irradiation with energy-rich particles on the hardness of the Fe-Cr alloy; Einfluss der Bestrahlung mit energiereichen Teilchen auf die Haerte von Fe-Cr-Legierungen

Energy Technology Data Exchange (ETDEWEB)

Heintze, Cornelia

2013-01-14

Ferritic/martensitic and oxide dispersion strengthened ferritic/martensitic steels are candidate structural materials for components exposed to high neutron fluxes in future nuclear applications like fusion and generation IV fission reactors. The ductilebrittle transition and its shift to higher temperatures which is predominantly caused by irradiation hardening are main concerns for these materials. In the present work, the irradiation behaviour of binary Fe-Cr model alloys, which represent a simplified model for ferritic/martensitic steels, is studied. To this end irradiation with iron ions is used in order to simulate the neutron-induced damage. Due to the limited penetration depth characterization methods suitable for thin layers have to be applied. In the present case, nanohardness testing and transmission electron microscopy (TEM) are employed. The results, including the irradiation-induced hardness change of the layer as a function of chromium content, fluence and irradiation temperature and, for selected cases, quantitative TEM analyses, were exploited to identify irradiation-induced dislocation loops as one source of irradiation hardening. Additional results of small-angle neutron scattering experiments on neutron-irradiated specimens of the same alloys show that nm-scaled α'-phase precipitates also significantly contribute to the irradiation-induced hardness increase. An Orowan model is used to estimate the obstacle strengths posed to dislocation glide by these lattice defects. The topic is stepwise extended to more complex situations with respect to the irradiation conditions and the materials. Considering simultaneous and sequential irradiations with iron- and helium-ions it is shown that the effect of helium on irradiation hardening depends on the chronological order of the irradiations and that the simultaneous introduction of helium in fusion-relevant concentrations amplifies irradiation hardening based on a synergistic effect. There is no

Extreme embrittlement of austenitic stainless steel irradiated to 75--81 dpa at 335--360 C

International Nuclear Information System (INIS)

Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V.; Garner, F.A.

1998-01-01

This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at ∼270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the swelling regime to displacement rate and the large, unprecedented levels of swelling reached at 335--360 C at these high neutron fluences. The failure mechanism appears to be identical to that observed at similar swelling levels in other austenitic steels irradiated in US fast reactors at 400--425 C, whereby stress-concentration between voids and nickel segregation at void surfaces predisposes the steel to an epsilon martensite transformation followed by formation of alpha martensite at crack tips. The very slow strain rate inherent in such creep tests and the relatively high helium levels may also contribute to the failure

A new and unusual deformation behavior observed in 12Cr18Ni10Ti stainless steel irradiated at 307 deg. C to 55 dpa in BN-350

International Nuclear Information System (INIS)

Gusev, M.; Maksimkin, O.; Osipov, I.S.; Garner, F.

2007-01-01

Full text of publication follows: It is currently accepted that neutron irradiation of stainless steels in general leads to increased strength, reduction of ductility and inevitably to embrittlement. The microstructural origins of such changes in mechanical behavior are well understood. Occasionally, however, a new phenomenon is observed at higher fluences. Void-induced embrittlement is an example whereby the ductility loss is strongly accelerated when new microstructural conditions develop from voids that cause stress concentration, removal of nickel from the matrix and thereby induce a martensitic transformation. This process occurs at moderately high temperatures where high void swelling can occur. It now appears that there is another, previously unobserved phenomenon that develops in austenitic steel irradiated to relatively high dose and relatively low temperature. In this case, however, the loss of plasticity commonly developed at lower dose is reversed and is replaced by an unusually high deformation. The plastic deformation was studied of miniature flat tensile specimens of 12Cr18Ni10Ti austenitic steel cut from a fuel assembly wrapper irradiated in the BN-350 reactor to 55 dpa at 580 K (307 deg. C). A new optical extensometry technique was employed that uses a video camera and multiple tiny markers painted on the specimen, allowing visualization and recording of the strain distribution as it develops along the specimen. The total deformation derived from the engineering diagrams for these specimens was 35-40%, while 3-7% was expected from previous studies conducted at lower dpa levels. The video record showed that the material resists necking and involves a moving deformation wave that initiates near one of the tensile grippers and spreads along ∼3/4 of the gauge length before failure occurs. Such behavior, often called a 'moving neck' has been observed previously in pure iron and Al-Mg alloys but has not been observed in irradiated stainless steels

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

Irradiation creep in reactor graphites for HTR applications. [Neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Veringa, H J; Blackstone, R [Stichting Reactor Centrum Nederland, Petten

1976-01-01

A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400/sup 0/C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400/sup 0/C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density.

Effects of Ti element on the microstructural stability of 9Cr–WVTiN reduced activation martensitic steel under ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Jin, Shuoxue; Li, Tiecheng; Chen, Jihong [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Suo, Jinping; Yang, Feng [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada)

2014-12-15

Microstructure of 9Cr–WVTiN reduced-activation martensitic steels with two different Ti concentrations irradiated with Fe{sup +}, He{sup +} and H{sup +} at 300 °C was studied with transmission electron microscopy. Small dislocation loops were observed in the irradiated steels. The mean size and number density of dislocation loops decreased with the increase of Ti concentration. The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in the low Ti-concentration 9Cr–WVTiN steel.

Neutronics analysis of International Fusion Material Irradiation Facility (IFMIF). Japanese contributions

International Nuclear Information System (INIS)

Oyama, Yukio; Noda, Kenji; Kosako, Kazuaki.

1997-10-01

In fusion reactor development for demonstration reactor, i.e., DEMO, materials tolerable for D-T neutron irradiation are absolutely required for both mechanical and safety point of views. For this requirement, several kinds of low activation materials were proposed. However, experimental data by actual D-T fusion neutron irradiation have not existed so far because of lack of fusion neutron irradiation facility, except fundamental radiation damage studies at very low neutron fluence. Therefore such a facility has been strongly requested. According to agreement of need for such a facility among the international parties, a conceptual design activity (CDA) of International Fusion Material Irradiation Facility (IFMIF) has been carried out under the frame work of the IEA-Implementing Agreement. In the activity, a neutronics analysis on irradiation field optimization in the IFMIF test cell was performed in three parties, Japan, US and EU. As the Japanese contribution, the present paper describes a neutron source term as well as incident deuteron beam angle optimization of two beam geometry, beam shape (foot print) optimization, and dpa, gas production and heating estimation inside various material loading Module, including a sensitivity analysis of source term uncertainty to the estimated irradiation parameters. (author)

Irradiation creep of the martensitic steel no. 1.4914 between 400 deg C and 600 deg C (Mol 5B)

International Nuclear Information System (INIS)

Herschbach, K.; Doser, W.

1983-01-01

The irradiation induced creep of the martensitic steel DIN No. 1.4914 was investigated in the temperature range from 400 to 600 deg C for stresses up to 200 Mpa using the Mol 5B irradiation rig. The results point to a behavior quite different from that observed in the austenitic steels as will be discussed in detail. The creep is thermally activated and non-linearly dependent upon the applied stress. (author)

Method of measuring neutron spectra in JMTR exclusively used for irradiation and their evaluation

International Nuclear Information System (INIS)

Sakurai, Kiyoshi

1983-01-01

In the core of the Japan Materials Testing Reactor, about 60 capsules are irradiated. These are the material capsules for irradiating reactor materials, the fuel capsules for irradiating reactor fuel, the RI capsules for producing radioisotopes and so on. In the irradiation experiment using a reactor, the information on the neutron fluence is indispensable, and the neutron fluence in the irradiated specimen part is evaluated with a dosimeter or the nuclear calculation for the core of the JMTR. At the time of irradiating reactor materials, the dosimeter Fe-54 (n,p) Mn-54 is generally used for evaluating the neutron fluence more than 1 MeV. In the case of fuel irradiation, the thermal neutron fluence is evaluated with the dosimeter Co-59 (n,γ) Co-60. It is important to examine in detail neutron spectra by both calculation and experiment in the reactors exclusively used for irradiation such as the JMTR. The neutron irradiation field in the JMTR, neutron spectrum measuring experiment, the neutron flux monitors for standardizing data, the measurement of X-ray and gamma ray, neutron guess spectrum, the compilation of neutron cross section for SAND 2, and the unfolding of neutron spectra are reported. The degree of agreement of the neutron fluence more than 1 MeV by measurement and calculation was +- 10 to 20 %. (Kako, I.)

Irradiation Effects in Fortiweld Steel Containing Different Boron Isotopes

Energy Technology Data Exchange (ETDEWEB)

Grounes, M

1967-07-15

Tensile specimens and miniature impact specimens of the low alloyed pressure vessel steel Fortiweld have been irradiated at 265 deg C in R2 to two neutron doses, 6.5 x 10{sup 18} n/cm{sup 2} (> 1 MeV) and 4 x 10{sup 19} n/cm{sup 2} (thermal) and also 9.0 x 10{sup 18} n/cm{sup 2} (> 1 MeV) and 6 x 10{sup 19} n/cm{sup 2} (thermal). Material from three laboratory melts, in which the boron consisted of {sup 10}B, {sup 11}B and natural boron respectively, were investigated. The results both of tensile tests and impact tests with miniature impact specimens show that the {sup 10}B-alloyed material was changed more and the {sup 11}B-alloyed material was changed less than the material containing natural boron. At the higher neutron dose the increase in yield strength (0.2 % offset yield strength) was 11 kg/mm in the {sup 10}B containing material compared to 5 kg/mm in the {sup 11}B-containing material. The decrease in total elongation was 5 and 0 percentage units respectively. The transition temperature was increased 190 deg C at the higher neutron dose in the {sup 10}B-alloyed material, 40 deg C in the {sup 11}B-alloyed material and 80 deg C in the material containing natural boron.

Standard Practice for Conducting Irradiations at Accelerator-Based Neutron Sources

CERN Document Server

American Society for Testing and Materials. Philadelphia

1996-01-01

1.1 This practice covers procedures for irradiations at accelerator-based neutron sources. The discussion focuses on two types of sources, namely nearly monoenergetic 14-MeV neutrons from the deuterium-tritium T(d,n) interaction, and broad spectrum neutrons from stopping deuterium beams in thick beryllium or lithium targets. However, most of the recommendations also apply to other types of accelerator-based sources, including spallation neutron sources (1). Interest in spallation sources has increased recently due to their proposed use for transmutation of fission reactor waste (2). 1.2 Many of the experiments conducted using such neutron sources are intended to simulate irradiation in another neutron spectrum, for example, that from a DT fusion reaction. The word simulation is used here in a broad sense to imply an approximation of the relevant neutron irradiation environment. The degree of conformity can range from poor to nearly exact. In general, the intent of these simulations is to establish the fundam...

A theoretical study of a nickel SPND using Warren's model

International Nuclear Information System (INIS)

Mahant, A.K.; Rao, P.S.; Misra, S.C.

1998-01-01

Various parameters for a nickel SPND viz. neutron and gamma sensitivity, their dependence on detector dimensions, effect of activity build up during irradiation in the reactor on neutron sensitivity and burn up have been calculated using an analytical model by Warren and Shah. The results show that the main component of neutron sensitivity is due to neutron-induced capture gamma rays (99.93%). Delayed β signal is negligible and signal due to external gamma rays constitute about 0.06%. Low gamma sensitivity to external gamma rays, negligible activity buildup and low burn up rate make it a very promising prompt SPND for power reactors. (orig.)

Neutron irradiation facilities for fission and fusion reactor materials studies

International Nuclear Information System (INIS)

Rowcliffe, A.F.

1985-01-01

The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

Neutron irradiation effects in advanced superconductors

International Nuclear Information System (INIS)

Yoshida, H.; Kodaka, H.; Miyata, K.; Hayashi, Y.; Atobe, K.

1988-01-01

This paper reports the effects of neutron irradiation on superconducting transitions studied by susceptibility and resistivity measurements for A15 type compounds, Laves-phase compounds and oxide superconductors. For A15 superconductors, the transition temperature (T c ) decreased with increasing neutron fluence and showed large drop started at about 5 x 10 18 n/cm 2 (E > 0.1 MeV). Post-irradiation annealing gave recovery of T c , but the behaviors were different for the materials with different composition and microstructure. The Laves-phase compounds showed less degradation than the A15 superconductors. For oxide superconductors very sensitive transition change was observed, including the radiation-induced superconductivity

Change in properties of superconducting magnet materials by fusion neutron irradiation

International Nuclear Information System (INIS)

Nishimura, Arata; Nishijima, Shigehiro; Takeuchi, Takao; Nishitani, Takeo

2007-01-01

A fusion reactor will generate a lot of high energy neutron and much energy will be taken out of the neutrons by a blanket system. Since some neutrons will stream out of a plasma vacuum vessel through neutral beam injection ports and penetrate a blanket system, a superconducting magnet system, which provides high magnetic field to confirm high energy particles, will be irradiated by a certain amount of neutrons. By developing the new NBI system or by reducing the penetration, the neutron fluence to the superconducting magnet will be able to be reduced. However, it is not easy to achieve the lower streaming and penetration at the present. Therefore, investigations on irradiation behavior of superconducting magnet materials are desired and some novel researches have been performed from 1970s. In general, the critical current of the superconducting wire increases under fast neutron environment comparing with that of the non-irradiated wire, and then decreased to almost zero as an increase of neutron fluence. On the other hand, the critical temperature of the wire starts to get down around 10 22 n/m 2 of neutron fluence and the temperature margin will be decreased during the operation by the neutron irradiation. In this paper, some aspects of irradiated materials will be overviewed and general tendency will be discussed focussing on knock-on effect of fast neutron and long range ordering of A15 compounds

Neutron capture therapy with thermal neutrons at IRT MIFI

International Nuclear Information System (INIS)

Zajtsev, K.N.; Portnov, A.A.; Savkin, V.A.; Kulakov, V.N.; Khokhlov, V.F.; Shejno, I.N.; Vajnson, A.A.; Kozlovskaya, N.G.; Meshcherikova, V.V.; Mitin, V.N.; Yarmonenko, S.P.

2001-01-01

Combined preclinical investigations into neutron capture therapy are conducted. Malignant melanoma was adopted as the line of investigation; boron-containing and gadolinium-containing preparations were used during the neutron capture therapy working off. Preparations produce secondary varying radiations when used in tumor. Dogs with spontaneous melanoma were used for the experiments. Procedures for the irradiation of dogs by neutron beam as the stage before use for the treatment of oncology patients were finished off; efficiency of neutron beam influence on normal tissues during the irradiation of dogs with melanoma (and without it) in antitumor and side effect sense was estimated [ru

Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

International Nuclear Information System (INIS)

Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

2007-01-01

The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

Directory of Open Access Journals (Sweden)

Hu J.-P.

2016-01-01

Full Text Available Radiation dosimetry for Neutron Capture Therapy (NCT has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR. In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1 in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2 out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3 beam shutter upgrade to reduce strayed neutrons and gamma dose, (4 beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5 beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates to reduce prompt gamma and fast neutron doses, (6 sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7 holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4–7

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

Energy Technology Data Exchange (ETDEWEB)

Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N.

2014-05-23

Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D₂O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7

Mechanodynamical analysis of nickel-titanium alloys for orthodontics application

International Nuclear Information System (INIS)

Arruda, Carlos do Canto

2002-01-01

Nickel-titanium alloys may coexist in more than one crystalline structure. There is a high temperature phase, austenite, and a low temperature phase, martensite. The metallurgical basis for the superelasticity and the shape memory effect relies in the ability of these alloys to transform easily from one phase to another. There are three essential factors for the orthodontist to understand nickel-titanium alloys behaviour: stress; deflection; and temperature. These three factors are related to each other by the stress-deflection, stress-temperature and deflection-temperature diagrams. This work was undertaken with the objective to analyse commercial nickel-titanium alloys for orthodontics application, using the dynamical mechanical analyser - DMA. Four NiTi 0,017 X 0,025'' archwires were studied. The archwires were Copper NiTi 35 deg C (Ormco), Neo Sentalloy F200 (GAC), Nitinol Superelastic (Unitek) and NiTi (GAC). The different mechanodynamical properties such as elasticity and damping moduli were evaluated. Each commercial material was evaluated with and without a 1 N static force, aiming to evaluate phase transition temperature variation with stress. The austenitic to martensitic phase ratio, for the experiments without static force, was in the range of 1.59 to 1.85. For the 1 N static force tests the austenitic to martensitic phase ratio, ranged from 1.28 to 1.57 due to the higher martensite elasticity modulus. With elastic modulus variation with temperature behaviour, the orthodontist has the knowledge of the force variation applied in the tooth in relation to the oral cavity temperature change, for nickel-titanium alloys that undergo phase transformation. The damping capacity of the studied alloys depends on the materials state: martensitic phase; austenitic phase or during phase transformation. The martensitic phase shows higher dumping capacity. During phase transformation, an internal friction peak may be observed for the CuNiTi 35 deg C and Neo Sentalloy F

Neutron metrology in the HFR. Steel irradiation. R139-801 (SINAS)

International Nuclear Information System (INIS)

Ketema, D.J.

1999-02-01

The R139-80 series irradiation experiments is part of the NRG materials test programme to evaluate the irradiation behaviour of several types of austenitic stainless steel. Within this programme five R139-80 specimen holders were irradiated in the HFR Petten to different dose levels. This report presents the final metrology results obtained from activation monitors in a specimen holder, coded as R139-801, containing 12 Compact Tension (CT-10 mm) specimens made from the austenitic stainless steel types 308LSXB/TIG and 304-SXB. The R139-801 assembly was irradiated in channel 1 of a TRIO type facility placed in HFR core-position F8. The aim of this irradiation of specimen holder R139-801 was to reach a minimum target damage level of 7.5 dpa for the specimens at a temperature of 335C. The monitor sets are used to calculate the thermal and fast neutron fluences, displacements per atom and the generated helium content. Additionally detailed information concerning an estimation of the fluence and damage doses received by each specimen and its temperature during irradiation are presented. The main results of the thermal and fast neutron fluence measurements are presented. The results indicate that the obtained damage levels in the steel specimens loaded in this specimen holder vary from 5.8 to 7.9 dpa. The temperatures of the specimens during irradiation varied between 304 and 337C. 14 refs

Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications

Science.gov (United States)

Zinkle, S. J.; Boutard, J. L.; Hoelzer, D. T.; Kimura, A.; Lindau, R.; Odette, G. R.; Rieth, M.; Tan, L.; Tanigawa, H.

2017-09-01

Reduced activation ferritic/martensitic steels are currently the most technologically mature option for the structural material of proposed fusion energy reactors. Advanced next-generation higher performance steels offer the opportunity for improvements in fusion reactor operational lifetime and reliability, superior neutron radiation damage resistance, higher thermodynamic efficiency, and reduced construction costs. The two main strategies for developing improved steels for fusion energy applications are based on (1) an evolutionary pathway using computational thermodynamics modelling and modified thermomechanical treatments (TMT) to produce higher performance reduced activation ferritic/martensitic (RAFM) steels and (2) a higher risk, potentially higher payoff approach based on powder metallurgy techniques to produce very high strength oxide dispersion strengthened (ODS) steels capable of operation to very high temperatures and with potentially very high resistance to fusion neutron-induced property degradation. The current development status of these next-generation high performance steels is summarized, and research and development challenges for the successful development of these materials are outlined. Material properties including temperature-dependent uniaxial yield strengths, tensile elongations, high-temperature thermal creep, Charpy impact ductile to brittle transient temperature (DBTT) and fracture toughness behaviour, and neutron irradiation-induced low-temperature hardening and embrittlement and intermediate-temperature volumetric void swelling (including effects associated with fusion-relevant helium and hydrogen generation) are described for research heats of the new steels.

Coolant compatibility studies. The effect of irradiation on tensile properties and stress corrosion cracking sensitivity of martensitic steels. MANET 4 - complementary studies

International Nuclear Information System (INIS)

Nystrand, A.C.

1994-02-01

Tensile and stress corrosion cracking tests have been carried out on MANET-type (1.4914 and FV448) and reduced activation (LA12TaLC) high-chromium martensitic steels. The materials had previously been exposed up to 5000 h at ∼275 degrees C in the core, above the core and remote from the core of a high pressure water loop in the Studsvik R2 reactor. After the mechanical testing the materials were examined visually and metallographically. The steel samples exposed in the core section showed large increases in tensile yield strengths when tested at 250 degrees C. However, the magnitude of the radiation hardening was considerably smaller in the reduced activation steel compared to the commercial steels; this observation is consistent with published data on other high-chromium martensitic steels and is associated with the lower chromium content of the LA12TaLC steel (8.9%) compared with those of the commercial steels (10.6 and 11.3%). Irradiation assisted stress corrosion cracking (IASCC) was not detected in any of the stressed steel samples after autoclave testing for times up to 1500 h at 250 degrees C in air-saturated high purity water. This apparent resistance to IASCC may be due to the high chromium martensitic steels not being sensitized by the irradiation in a comparable manner to that shown by the austenitic steels. However, additional studies are required to clarify some of the existing uncertainties with respect to IASCC of these martensitic steels

Proceedings of the second milestone meeting of European laboratories on the development of ferritic/martensitic steels for fusion technology

International Nuclear Information System (INIS)

Daum, E.; Ehrlich, K.; Schirra, M.

1997-05-01

In the frame of the European Fusion Technology Program a series of ferritic/martensitic developmental alloys, the composition of which had been optimized towards low long-term activation, was investigated and compared with conventional 9-12%CrMoVNb steels. It could be shown that by these chemical modifications neither the physical metallurgy nor the transformation behavior was changed markedly. Tensile-, creep-rupture- and fatigue properties are somewhat reduced, whereas the fracture toughness and impact data are far superior to conventional materials. This is an important advantage, especially if the expected detrimental effect of neutron irradiation on the latter properties is taken into account. First results of low-fluence irradiations indicate that the new alloys are less prone to irradiation-induced DBTT shifts. (orig./HM) [de

Synergistic effects of neutron and gamma ray irradiation of a commercial CHMOS microcontroller

International Nuclear Information System (INIS)

Xiao-Ming, Jin; Ru-Yu, Fan; Wei, Chen; Dong-Sheng, Lin; Shan-Chao, Yang; Xiao-Yan, Bai; Yan, Liu; Xiao-Qiang, Guo; Gui-Zhen, Wang

2010-01-01

This paper presents the experimental results of a combined irradiation environment of neutron and gamma rays on 80C196KC20, which is a 16-bit high performance member of the MCS96 microcontroller family. The electrical and functional tests were made in three irradiation environments: neutron, gamma rays, combined irradiation of neutron and gamma rays. The experimental results show that the neutron irradiation can affect the total ionizing dose behaviour. Compared with the single radiation environment, the microcontroller exhibits considerably more severe degradation in neutron and gamma ray synergistic irradiation. This phenomenon may cause a significant hardness assurance problem. (condensed matter: structure, thermal and mechanical properties)

Effect of neutron irradiation on p-type silicon

International Nuclear Information System (INIS)

Sopko, B.

1973-01-01

The possibilities are discussed of silicon isotope reactions with neutrons of all energies. In the reactions, 30 Si is converted to a stable phosphorus isotope forming n-type impurities in silicon. The above reactions proceed as a result of thermal neutron irradiation. An experiment is reported involving irradiation of two p-type silicon single crystals having a specific resistance of 2000 ohm.cm and 5000 to 20 000 ohm.cm, respectively, which changed as a result of irradiation into n-type silicon with a given specific resistance. The specific resistance may be pre-calculated from the concentration of impurities and the time of irradiation. The effects of irradiation on other silicon parameters and thus on the suitability of silicon for the manufacture of semiconductor elements are discussed. (J.K.)

Structural properties and neutron irradiation effects of ceramics

International Nuclear Information System (INIS)

Yano, Toyohiko

1994-01-01

In high temperature gas-cooled reactors and nuclear fusion reactors being developed at present, various ceramics are to be used in the environment of neutron irradiation for undertaking important functions. The change of the characteristics of those materials by neutron irradiation must be exactly forecast, but it has been known that the response of the materials is different respectively. The production method of ceramics and the resulted structures of ceramics which control their characteristics are explained. The features of covalent bond and ionic bond, the synthesis of powder and the phase change by heating, sintering and sintering agent, and grain boundary phase are described. The smelling of ceramics by neutron irradiation is caused by the formation of the clusters of Frenkel defects and minute spot defects. Its restoration by annealing is explained. The defects remaining in materials after irradiation are the physical defects by flipping atoms cut due to the collision with high energy particles and the chemical defects by nuclear transformation. Some physical defects can be restored, but chemical defects are never restored. The mechanical properties of ceramics and the effect of irradiation on them, and the thermal properties of ceramics and the effect of irradiation on them are reported. (K.I.)

Microstructure and mechanical properties in the weld heat affected zone of 9Cr-2W-VTa reduced activation ferritic/martensitic steel for fusion

Energy Technology Data Exchange (ETDEWEB)

Moon, Joonoh; Lee, Changhoon; Lee, Taeho; Jang, Minho; Park, Mingu [Korea Institute of Materials Science, Changwon (Korea, Republic of); Kim, Hyoung Chan [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Reduced activation ferritic/martensitic (RAFM) steel demonstrated excellent resistance to the neutron irradiation and mechanical properties. The investigation of weldability in company with the development of RAFM steel is essential for construction of the fusion reactor. Generally, the superior mechanical properties of the RAFM steel can be upset during welding process due to microstructural change by rapid heating and cooling in the weld heat affected zone (HAZ). The phase transformation and mechanical properties in the weld HAZ of RAFM steel were investigated. The base steel consisted of tempered martensite and two carbides. During rapid welding thermal cycle, the microstructure of the base steel was transformed into martensite and δ-ferrite. In addition, the volume fraction of δ-ferrite and grain size increased with increase in the peak temperature and heat input. The strength of the HAZs was higher than that of the base steel due to the formation of martensite, whereas the impact properties of the HAZs deteriorated as compared with the base steel due to the formation of δ-ferrite. The PWHT improved the impact properties of the HAZs, resulting from the formation of tempered martensite.

Fusion neutron irradiation of Ni(Si) alloys at high temperature

International Nuclear Information System (INIS)

Huang, J.S.; Guinan, M.W.; Hahn, P.A.

1987-09-01

Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab

Fusion neutron irradiation of Ni(Si) alloys at high temperature

Energy Technology Data Exchange (ETDEWEB)

Huang, J.S.; Guinan, M.W.; Hahn, P.A.

1987-09-01

Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

Remarkable photo-catalytic degradation of malachite green by nickel doped bismuth selenide under visible light irradiation

Science.gov (United States)

Kulsi, Chiranjit; Ghosh, Amrita; Mondal, Anup; Kargupta, Kajari; Ganguly, Saibal; Banerjee, Dipali

2017-01-01

Bismuth selenide (Bi2Se3) and nickel (Ni) doped Bi2Se3 were prepared by a solvothermal approach to explore the photo-catalytic performance of the materials in degradation of malachite green (MG). The presence of nickel was confirmed by X-ray photoelectron spectroscopy (XPS) measurement in doped Bi2Se3. The results showed that the nickel doping played an important role in microstructure and photo-catalytic activity of the samples. Nickel doped Bi2Se3 sample exhibited higher photo-catalytic activity than that of the pure Bi2Se3 sample under visible-light irradiation. The photo-catalytic degradation followed first-order reaction kinetics. Fast degradation kinetics and complete (100% in 5 min of visible light irradiation) removal of MG was achieved by nickel doped Bi2Se3 in presence of hydrogen peroxide (H2O2) due to modification of band gap energies leading to suppression of photo-generated electron-hole recombination.

Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

International Nuclear Information System (INIS)

Chisela, F.

1986-01-01

The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

Annealing of dislocation loops in neutron-irradiated copper investigated by positron annihilation

International Nuclear Information System (INIS)

Gauster, W.B.; Mantl, S.; Schober, T.; Triftshauser, W.

1975-01-01

Positron annihilation angular correlation measurements were carried out on neutron-irradiated copper as a function of annealing temperature. Two types of specimens were used: single crystals irradiated with fast neutrons, and 10 B-doped polycrystalline samples irradiated with thermal neutrons. All irradiations were at approximately 320 0 K. A structure in the annealing curve, not previously observed by other techniques, indicates that between 460 and 600 0 K the dislocation loops present after irradiation dissociate and more effective positron trapping sites are formed. (auth)

Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

International Nuclear Information System (INIS)

Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

2007-01-01

Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

NEUTRON SPECTRUM MEASUREMENTS USING MULTIPLE THRESHOLD DETECTORS

Energy Technology Data Exchange (ETDEWEB)

Gerken, William W.; Duffey, Dick

1963-11-15

From American Nuclear Society Meeting, New York, Nov. 1963. The use of threshold detectors, which simultaneously undergo reactions with thermal neutrons and two or more fast neutron threshold reactions, was applied to measurements of the neutron spectrum in a reactor. A number of different materials were irradiated to determine the most practical ones for use as multiple threshold detectors. These results, as well as counting techniques and corrections, are presented. Some materials used include aluminum, alloys of Al -Ni, aluminum-- nickel oxides, and magesium orthophosphates. (auth)

A microstructural comparison of two nuclear-grade martensitic steels using small-angle neutron scattering

Science.gov (United States)

Coppola, R.; Fiori, F.; Little, E. A.; Magnani, M.

1997-06-01

Results are presented of a small-angle neutron scattering (SANS) study on two 10-13% Cr martensitic stainless steels of interest for nuclear applications, viz. DIN 1.4914 (MANET specification, for fusion reactors) and AISI 410. The investigation has focussed principally on microstructural effects associated with the differences in chromium content between the two alloys. The size distribution functions determined from nuclear and magnetic SANS components for the two steels given identical heat treatments are in accord with an interpretation based on the presence of ˜ 1 nm size CCr aggregates in the microstructure. Much larger (˜ 10 nm) scattering inhomogeneities with different magnetic contrast are also present and tentatively identified as carbides.

Strčess-induced martensitic transformation in Cu-Al-Zn-Mn polycrystal investigated by two in -situ neutron diffraction techniques

Czech Academy of Sciences Publication Activity Database

Šittner, Petr; Lukáš, Petr; Neov, Dimitar; Daymond, M. R.; Novák, Václav; Swallowe, G. M.

2002-01-01

Roč. 324, - (2002), s. 225-234 ISSN 0921-5093 R&D Projects: GA MŠk ME 186; GA ČR GV202/97/K038; GA AV ČR IAA1010909 Institutional research plan: CEZ:AV0Z1010914 Keywords : Stress-induced martensitic transformation * Cu-Al-Zn-Mn polycrystal * neutron diffraction technique Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.107, year: 2002

Neutron irradiation of rat embryos in utero

International Nuclear Information System (INIS)

Vogel, H.H. Jr.

1978-01-01

In the rat radiation is most effective in producing congenital anomalies during the organ-forming period (days 9 to 13), which is approximately equivalent to the 14th to 50th days of human pregnancy. We have exposed female Sprague--Dawley rats on the 18th day of pregnancy to single whole-body doses of fission neutrons (20 to 150 rads). After 20 rads there was a small decrease in body weight which lasted from birth to weaning. During this period 9% of the irradiated rats died compared with 4% of the controls. After 50 rads, 65/275 (23.6%) of the rats died between birth and weaning, and the body-weight loss of the survivors was increased. After 100 rads, 62/133 (47%) died at birth or day 1 and 103/133 (77.4%) died before weaning. A large and significant decrease in body weight persisted in the survivors. After 150 rads of fission neutrons, all 95 rats died within 48 hr of birth. From cross-fostering experiments, we believe this is a direct effect of radiation on the embryos and not an indirect action through the mother or her milk. The LD 50 for the period from birth to weaning is approximately 75 rads of fission neutrons. Studies of organ weight were conducted daily for the first week after birth in an attempt to find the cause of radiation mortality. Body weight of the irradiated animals averaged only about one-half that of the controls. The liver, kidney, brain, and testes of the neutron-irradiated rats weighed significantly less than those of the controls. The weights of the spleen, lungs, duodenum, and stomach were decreased but not significantly. The bone marrow appeared depleted in the irradiated long bones, but the spleen maintained active hematopoiesis 1 to 2 months after neutron exposure

Determination of urine metabolites containing radioactivatable elements by molecular neutron activation analysis

International Nuclear Information System (INIS)

Blotcky, A.J.; Rack, E.P.

1986-01-01

As urine is a final stage for the metabolic pathways of essential trace elements or chemical toxins, it is becoming increasingly important to not only report levels of trace elements but to determine the molecular or ionic identity of these trace elements. For a biological system such as urine, a molecular neutron activation analysis (MONAA) approach must involve a deproteinization step, where necessary, to ensure that metabolites such as amino acids, bases, r nucleosides are not protein bound prior to chemical separation. This can involve the simple application of ammonia or acid hydrolysis. All separations for the metabolites containing the radioactivatable element must be performed prior to neutron irradiation and subsequent radioassay for the metabolite. Separation procedures can include high-pressure liquid chromotography (HPLC), ion-exchange chromatography, size exclusion chromatography, solvent extraction, and/or gas chromatography. After separation, the separated metabolite is neutron irradiated and and radioassayed for the radioactivity in the metabolite. A review of previous work involving the determination of hormonal iodine, iodoamino acids, chlorinated pesticides, trimethyl-selenonium, and selenoamino acids in urine is discussed

Analysis of the non-isothermal austenite-martensite transformation in 13% Cr-type martensitic stainless steels

International Nuclear Information System (INIS)

Garcia-De-Andris, C.; Alvarez, L.F.

1996-01-01

In martensitic stainless steels, as in other alloyed containing carbide-forming elements, the carbide dissolution and precipitation processes that take place during heat treatment can cause modifications to the chemical composition of the austenite phase of these steels. The chemical composition of this phase is a fundamental factor for the evolution of the martensitic transformation. As a result of their influence on the dissolution and precipitation processes, the parameters of the quenching heat treatment exert a strong influence on the behavior of the martensitic transformation in these steels. In the present study, the effect of the heating temperature and the cooling rate on the martensitic transformation in two 13% Cr-type martensitic stainless steels with different carbon contents were properly evaluated. (author)

Container Inspection Utilizing 14 MeV Neutrons

Science.gov (United States)

Valkovic, Vladivoj; Sudac, Davorin; Nad, Karlo; Obhodas, Jasmina

2016-06-01

A proposal for an autonomous and flexible ship container inspection system is presented. This could be accomplished by the incorporation of an inspection system on various container transportation devices (straddle carriers, yard gentry cranes, automated guided vehicles, trailers). The configuration is terminal specific and it should be defined by the container terminal operator. This enables that no part of the port operational area is used for inspection. The inspection scenario includes container transfer from ship to transportation device with the inspection unit mounted on it. The inspection is performed during actual container movement to the container location. A neutron generator without associated alpha particle detection is used. This allows the use of higher neutron intensities (5 × 109 - 1010 n/s in 4π). The inspected container is stationary in the “inspection position” on the transportation device while the “inspection unit” moves along its side. The following analytical methods will be used simultaneously: neutron radiography, X-ray radiography, neutron activation analysis, (n, γ) and (n,n'γ) reactions, neutron absorption. and scattering, X-ray backscattering. The neutron techniques will utilize “smart collimators” for neutrons and gamma rays, both emitted and detected. The inspected voxel is defined by the intersection of the neutron generator and the detectors solid angles. The container inspection protocol is based on identification of discrepancies between the cargo manifest, elemental “fingerprint” and radiography profiles. In addition, the information on container weight is obtained during the container transport and screening by measuring of density of material in the container.

Effects of neutron irradiation on microstructure in experimental and commercial ferritic alloys

International Nuclear Information System (INIS)

Gelles, D.S.; Thomas, L.E.

1983-05-01

A series of microstructural studies have been undertaken on fast-reactor-irradiated specimens of experimental ferritic alloys and ferritic/martensitic commercial alloys covering a broad range of compositions and starting microstructures. It is found that voids do indeed form in ferritic alloys and that dislocation loops and tangles are created during irradiation at temperatures below 500 0 C. Swelling rates as high as 0.25% per 10 22 n/cm 2 have been measured. However, the major effect of irradiation is precipitation and precipitation can suppress void swelling completely and/or be responsible for degradation of mechanical properties

Martensitic cubic → tetragonal transition

International Nuclear Information System (INIS)

Schumann, H.

1983-01-01

Indium-thallium alloys containing 14 to 30% At. Tl have a cubic face-centred beta phase wich changes into a tetragonal face-centred alpha martensite during solidification. The martensite contains twin crystals that are large enough to be seen by means of a light microscope. The phenomenological crystallographic martensite theory was used to calculate Miller's index of the habit plane, the formation of the surface relief, the orientation relations and the critical thickness ratio of the twins. In a beta monocrystal frequently only one of the 24 crystallographic possible habit planes are formed at one end of the sample and migrate through the whole crystal when the temperature drops. Externally applied tension and compression influence in different ways the direction in which the habit plane moves and can even destroy the twinned structure, i.e. they can modify the substructure of the martensite crystal. This induces superelasticity, an effect that has also been described quantitatively. (author)

Neutron self-shielding with k0-NAA irradiations

International Nuclear Information System (INIS)

Chilian, C.; Chambon, R.; Kennedy, G.

2010-01-01

A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.

Level gauge using neutron radiation

International Nuclear Information System (INIS)

Mathew, P.J.

1985-01-01

Apparatus for determining the level of a solid or liquid material in a container comprises: a vertical guide within or alongside the container; a sensor positioned within the guide; means for moving the sensor along the guide; and means for monitoring the position of the sensor. The sensor comprises a source of fast neutrons, a detector for thermal neutrons, and a body of a neutron moderating material in close proximity to the detector. Thermal neutrons produced by fast neutron irradiation of the solid or liquid material, or thermal neutrons produced by irradiation of the neutron-moderating material by fast or epithermal neutrons reflected by the solid or liquid material, are detected when the sensor is positioned at or below the level of the material in the container

Diffusion Couple Alloying of Refractory Metals in Austenitic and Ferritic/Martensitic Steels

Science.gov (United States)

2012-03-01

stainless steel and ferritic/ martensitic steel can vary from structural and support components in the reactor core to reactor fuel...of ferritic/ martensitic steels compared to type 316 stainless steel after irradiation in Experimental Breeder Reactor-II at 420 ºC to ~80dpa (From...ferritic martensitic steel at Sandia National Laboratories. The 316 stainless steel had a certified composition of:

Fracture toughness and strength change of neutron-irradiated ceramic materials

International Nuclear Information System (INIS)

Dienst, W.; Zimmermann, H.

1994-01-01

In order to analyse the results of bending strength measurements on neutron-irradiated samples of Al 2 O 3 , AlN and SiC, fracture toughness measurements were additionally conducted. The neutron fluences concerned were mostly in the range of 0.6 to 3.2x10 26 n/m 2 at irradiation temperatures of 400 to 550 C. A fracture toughness decrease was generally observed for polycrystalline materials which, however, was considerably smaller than the reduction of the fracture strength. Exceptional increase of the fracture toughness seems typical for the effect of rather coarse irradiation defects. The irradiation-induced change of the fracture toughness of single crystal Al 2 O 3 appeared dependent on the crystallographic orientation; both reduced and increased fracture toughness after irradiation was observed. Recent results of neutron irradiation to about 2x10 25 n/m 2 at 100 C showed, that the strength decrease of various Al 2 O 3 grades sets in at (3-5)x10 24 n/m 2 and seems to be little dependent on the irradiation temperature. ((orig.))

Shielding of a neutron irradiator with {sup 241}Am-Be source

Energy Technology Data Exchange (ETDEWEB)

Oliveira, K.A.M. de; Crispim, V.R.; Silva, A.X., E-mail: koliveira@con.ufrj.b, E-mail: verginia@con.ufrj.b, E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear; Fonseca, E.S., E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2011-07-01

The equivalent dose rates at 1.0 cm from the outer surface of the shielding of a neutron irradiation system that uses {sup 241}Am-Be source with activity of 185 GBq (5 Ci) were determined. A theoretical-experimental approach including case studies, through computer simulations with MCNP code was employed to calculate the best shielding thickness. Following the construction of the neutron irradiator, dose measurements were conducted in order to validate data obtained from simulation. The neutron irradiator shielding was designed in such a way to allow transport of the neutron radiography system for in loco inspections ensuring workers' radiologic safety. (author)

Activation analysis of tritium breeder lithium lead irradiated by fusion neutrons in FDS-II

International Nuclear Information System (INIS)

Mingliang Chen

2006-01-01

R-and-D of fusion materials, especially their activation characteristics, is one of the key issues for fusion research in the world. Research on activation characteristics for low activation materials, such as reduced activation ferritic/martensitic steels, vanadium alloys and SiCf/SiC composites, is being done throughout the world to ensure the attractiveness of fusion power regarding safety and environmental aspects. However, there is less research on the activation characteristics of the other important fusion materials, such as tritium breeder etc.. Lithium lead (Li 17 Pb 83 ) is presently considered as a primary candidate tritium breeder for fusion power reactors because of its attractive characteristics. It can serve as a tritium breeder, neutron multiplier and coolant in the blanket at the same time. The radioactivity of Li 17 Pb 83 by D-T fusion neutrons in FDS-II has been calculated and analyzed. FDS-II is a concept design of fusion power reactor, which consists of fusion core with advanced plasma parameters extrapolated from the ITER (International Thermonuclear Experimental Reactor) and two candidates of liquid lithium breeder blankets (named SLL and DLL blankets). The neutron transport and activation calculation are carried out based on the one-dimensional model for FDS-II with the home-developed multi-functional code system VisualBUS and the multi-group data library HENDL1.0/MG and European Activation File EAF-99. The effects of irradiation time on the activation characteristics of Li 17 Pb 83 were analyzed and it concludes that the irradiation time has an important effect on the activation level of Li 17 Pb 83 . Furthermore, the results were compared with the activation levels of other tritium breeders, such as Li 4 SiO 4 , Li 2 TiO 3 , Li 2 O and Li etc., under the same irradiation conditions. The dominant nuclides to dose rate and activity of Li 17 Pb 83 were analyzed as well. Tritium generated by Li has a great contribution to the afterheat and

Investigation of neutron fluence using fluence monitors for irradiation test at WWR-K

International Nuclear Information System (INIS)

Romanova, N.K.; Takemoto, N.

2013-01-01

Irradiation test of a Si ingot is planned using WWR-K in Institute of Nuclear Physics Republic of Kazakhstan (INP RK) to develop an irradiation technology for Si semiconductor production by Neutron Transmutation Doping (NTD) method in the framework of an international cooperation between INP RK and Japan Atomic Energy Agency (JAEA), Japan. It is possible to irradiate the Si ingot of 6 inch in diameter at the K-23 irradiation channel in the WWR-K. The preliminary irradiation test using 4 Al ingots was performed to evaluate the actual neutronic irradiation field at the K-23 channel in the WWR-K. Each Al ingot has the same dimension as the Si ingot, and 15 fluence monitors are equipped in it. Iron wire and aluminum-cobalt wire are inserted into them, and it is possible to evaluate both fast and thermal neutron fluxes by measurement of these radiation activities after irradiation. This report described the results of the preliminary irradiation test and the neutronic calculations by Monte Carlo method in order to evaluate the neutronic irradiation field in the irradiation position for the silicon ingot at the channel in the WWR-K. (authors)

Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

International Nuclear Information System (INIS)

Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

2011-01-01

Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

Neutron and gamma irradiation effects on power semiconductor switches

Science.gov (United States)

Schwarze, G. E.; Frasca, A. J.

1990-01-01

The performance characteristics of high power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN Bipolar Junction Transistors (BJTs), and Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs). BJTs show a rapid decrease in gain, blocking voltage, and storage time for neutron irradiation, and MOSFETs show a rapid decrease in the gate threshold voltage for gamma irradiation.

The decontamination of soft-plated nickel surfaces compared to alternative surface materials used in radioactive transport and storage containers

International Nuclear Information System (INIS)

Zwicky, H.U.; Bedenig, D.O.; Bohringer, I.M.; Petrik, F.

1983-01-01

Surfaces of raw, nickel-plated, and epoxy-coated spheroidal graphite cast iron, together with stainless steel, were contaminated with a modified fission product solution then conditioned by heat treatment. This was followed by a variety of simple decontamination techniques. It was shown that the ease of removal of contaminations similar to those expected on a dry storage container surface is significantly affected by the roughness of the surface. The raw cast iron surface was virtually impossible to significantly decontaminate. Highest decontamination factors were obtained on nickel-plated and epoxy-painted surfaces using steam/detergent mixtures. Stainless steel only performed well in a polished condition. In a supplementary irradiation experiment, scanning electron microscopy indicated visible decomposition of an epoxy-painted surface at a gamma dose of 3.1 X 10 6 Gy (3.1 X 10 8 rad). A nickel-plated surface did not undergo any visible changes at the same dose

Nickel Foil as Transmutation Detector for Neutron Fluence Measurements

Directory of Open Access Journals (Sweden)

Klupák Vít

2016-01-01

Full Text Available Activation detectors are very often used for determination of the neutron fluence in reactor dosimetry. However, there are few disadvantages concerning these detectors; it is the demand of the knowledge of the irradiation history and a loss of information due to a radioactive decay in time. Transmutation detectors TMD could be a solution in this case. The transmutation detectors are materials in which stable or long-lived nuclides are produced by nuclear reactions with neutrons. From a measurement of concentration of these nuclides, neutron fluence can be evaluated regardless of the cooling time.

An investigation of neutron irradiation test on superplastic zirconia-ceramic materials

International Nuclear Information System (INIS)

Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hayashi, Kimio

2000-05-01

A neutron irradiation test on superplastic ceramic materials at high temperature has been proposed as an innovative basic research on high-temperature engineering using the High Temperature Engineering Test Reactor (HTTR). For the effective execution of the test, we reviewed the superplastic deformation mechanism of ceramic materials and discussed neutron irradiation effects on the superplastic deformation process of stabilized Tetragonal Zirconia Polycrystal (TZP), which is a representative superplastic ceramic material. As a result, we pointed out that the decrease in the activation energy for superplastic deformation is expected by the radiation-enhanced diffusion. We selected a fast neutron fluence of 5x10 20 n/cm 2 and an irradiation temperature of about 600degC as test conditions for the first irradiation test on TZP and decided to perform a preliminary irradiation test by the Japan Materials Testing Reactor (JMTR). Moreover, we estimated the radioactivity of irradiated TZP and indicated that it is in the order of 10 10 Bq/g (about 0.3 Ci/g) immediately after irradiation to a thermal neutron fluence of 3x10 20 n/cm 2 and that it decays to about 1/100 in a year. (author)

Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

Energy Technology Data Exchange (ETDEWEB)

Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi [Musashi Inst. of Tech., Tokyo (Japan); Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao

1996-04-01

SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young`s modulus lowers. In addition, its composition elements have the large (n,{alpha}) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

International Nuclear Information System (INIS)

Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi; Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao.

1996-01-01

SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young's modulus lowers. In addition, its composition elements have the large (n,α) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

Neutrons from Antiproton Irradiation

DEFF Research Database (Denmark)

Bassler, Niels; Holzscheiter, Michael; Petersen, Jørgen B.B.

the neutron spectrum. Additionally, we used a cylindrical polystyrene loaded with several pairs of thermoluminescent detectors containing Lithium-6 and Lithium-7, which effectively detects thermalized neutrons. The obtained results are compared with FLUKA imulations. Results: The results obtained...... spectrum is very low, and does not pose a problem for radiation therapy. However, the contribution from fast neutrons is much more significant. The dose equivalent contribution from neutrons originate from the patient alone and reaches levels which are found in passive moderated proton therapy. The exact...

Fusion neutron irradiation of Ni-Si alloys at high temperature*1

Science.gov (United States)

Huang, J. S.; Guinan, M. W.; Hahn, P. A.

1988-07-01

Two Ni-4% Si alloys, with different cold work levels, have been irradiated with 14-MeV fusion neutrons at 623 K, and their Curie temperatures have been monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2-MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14-MeV fusion neutrons is only 6-7% of that for an identical alloy irradiated by 2-MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6-7% for the fusion neutron irradiated sample.

Aspects of Low Temperature Irradiation in Neutron Activation Analysis

International Nuclear Information System (INIS)

Brune, D.

1968-08-01

Neutron irradiation of the sample while frozen in a cooling device inserted in a reactor channel has been carried out in the analysis of iodine in aqueous samples as well as of mercury in biological tissue and water. For the simultaneous irradiation of a large number of aqueous solutions the samples were arranged in a suitable geometry in order to avoid mutual flux perturbation effects. The influence of the neutron temperature on the activation process has been discussed. Potential applications of the low temperature irradiation technique are outlined

Aspects of Low Temperature Irradiation in Neutron Activation Analysis

Energy Technology Data Exchange (ETDEWEB)

Brune, D

1968-08-15

Neutron irradiation of the sample while frozen in a cooling device inserted in a reactor channel has been carried out in the analysis of iodine in aqueous samples as well as of mercury in biological tissue and water. For the simultaneous irradiation of a large number of aqueous solutions the samples were arranged in a suitable geometry in order to avoid mutual flux perturbation effects. The influence of the neutron temperature on the activation process has been discussed. Potential applications of the low temperature irradiation technique are outlined.

Verification of neutron irradiation on S/G tube materials

International Nuclear Information System (INIS)

Kang, Byoung Hwi; Lee, S. K.; Jang, D. Y.; Jo, K. H.

2010-12-01

The fluence monitors were fabricated with metal wires of the purity ≥ 99.9%, whose dimensions were 0.1mm diameter, about 3mm length, and around 150-200 μg mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position. The measured neutron fluences were compared to the calculated ones. (Errors ≤ 35%)

Nickel-based gadolinium alloy for neutron adsorption application in ram packages

International Nuclear Information System (INIS)

Robino, C.; McConnell, P.; Mizia, R.

2004-01-01

This paper will outline the results of a metallurgical development program that is investigating the alloying of gadolinium into a nickel-chromium-molybdenum alloy matrix. Gadolinium has been chosen as the neutron absorption alloying element due to its high thermal neutron absorption cross section and low solubility in the expected U.S. repository environment. The nickel-chromium-molybdenum alloy family was chosen for its known corrosion performance, mechanical properties, and weldability. The workflow of this program includes chemical composition definition, primary and secondary melting studies, ingot conversion processes, properties testing, and national consensus codes and standards work. The microstructural investigation of these alloys shows that the gadolinium addition is not soluble in the primary austenite metallurgical phase and is present in the alloy as gadolinium-rich second phase. This is similar to what is observed in a stainless steel alloyed with boron. The mechanical strength values are similar to those expected for commercial Ni-Cr-Mo alloys. The alloys have been corrosion tested in simulated Yucca Mountain aqueous chemistries with acceptable results. The initial results of weldability tests have also been acceptable. Neutronic testing in a moderated critical array has generated favorable results. An American Society for Testing and Materials material specification has been issued for the alloy and a Code Case has been submitted to the American Society of Mechanical Engineers for code qualification. The ultimate goal is acceptance of the alloy for use at the Yucca Mountain repository

Synergistic effects on dislocation loops in reduced-activation martensitic steel investigated by single and sequential hydrogen/helium ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Zhang, Weiping [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Luo, Fengfeng [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Institute of Applied Physics, Jiangxi Academy of Sciences, Nanchang 330029 (China); Yu, Yanxia; Zheng, Zhongcheng; Shen, Zhenyu [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ren, Yaoyao [Center for Electron Microscopy, Wuhan University, Wuhan 430072 (China); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2016-10-15

Single-beam and sequential-beam irradiations were performed to investigate the H/He synergistic effect on dislocation loops in reduced-activation ferritic/martensitic (RAFM) steels. The irradiations were carried out with 10 keV H{sup +}, 18 keV He{sup +} and 160 keV Ar{sup +}, alone and in combination at 723 K. He{sup +} single-beam irradiation induced much larger dislocation loops than that induced by both H{sup +} and Ar{sup +} single-beam irradiation. H{sup +} post-irradiation after He{sup +} irradiation further increased the size of dislocation loops, whilst He{sup +} post-irradiation or Ar{sup +} post-irradiation following H{sup +} irradiation only slightly increased the size of dislocation loops. The experiment results indicate that pre-implanted H{sup +} can drastically inhibit the growth while post-implanted H{sup +} can significantly enhance the growth of dislocation loops induced by He{sup +} irradiation. The mechanisms behind the complex synergistic phenomena between H and He and the different roles that H and He played in the growth of dislocation loops are discussed.

Neutron irradiation induced amorphization of silicon carbide

International Nuclear Information System (INIS)

Snead, L.L.; Hay, J.C.

1998-01-01

This paper provides the first known observation of silicon carbide fully amorphized under neutron irradiation. Both high purity single crystal hcp and high purity, highly faulted (cubic) chemically vapor deposited (CVD) SiC were irradiated at approximately 60 C to a total fast neutron fluence of 2.6 x 10 25 n/m 2 . Amorphization was seen in both materials, as evidenced by TEM, electron diffraction, and x-ray diffraction techniques. Physical properties for the amorphized single crystal material are reported including large changes in density (-10.8%), elastic modulus as measured using a nanoindentation technique (-45%), hardness as measured by nanoindentation (-45%), and standard Vickers hardness (-24%). Similar property changes are observed for the critical temperature for amorphization at this neutron dose and flux, above which amorphization is not possible, is estimated to be greater than 130 C

Characterization of defect accumulation in neutron-irradiated Mo by positron annihilation spectroscopy

DEFF Research Database (Denmark)

Eldrup, Morten Mostgaard; Li, Meimei; Snead, L.L.

2008-01-01

Positron annihilation lifetime spectroscopy measurements were performed on neutron-irradiated low carbon arc cast Mo. Irradiation took place in the high flux isotope reactor, Oak Ridge National Laboratory, at a temperature of 80 +/- 10 degrees C. Neutron fluences ranged from 2 x 10(21) to 8 x 10(......, as predicted by molecular dynamics simulations. (C) 2008 Elsevier B.V. All rights reserved....... at a very low-dose of similar to 10(-4) dpa. The average size of the cavities did not change significantly with dose, in contrast to neutron-irradiated bcc Fe where cavity sizes increased with increasing dose. It is suggested that the in-cascade vacancy clustering may be significant in neutron-irradiated Mo...

Colony form variation of Bacillus pumilus E601 after cultured and neutron irradiation

International Nuclear Information System (INIS)

Chen Xiaoming; Wei Baoli; Zhang Jianguo

2008-01-01

The distribution of two colony forms of Bacillus pumilus E601 and the effect of neutron irradiation on the colony form were reported. The translucent and opaque colonies were cultured several generations to observe the proportion of two form colonies. The spores of opaque colony were irradiated at 80, 800 and 2000 Gy of fast neutron from CFBR-II pulse pile, and the survivors of opaque colony were irradiated again at the same doses. The results showed that: (1) Bacillus pumilus E601 observed two types of colony form: translucent and opaque colony; (2) the translucent colony could produce both translucent and opaque colonies in equal, while the opaque colony couldn't produce translucent colony generally; (3) neutron irradiation could affect the colony form distribution. The ratio of survival translucent colony was increased with the increase of the first neutron irradiation doses, and the second neutron irradiation also increased the ratio of translucent colony. It was concluded that the instability of translucent colony was the main reason to produce two colony forms of Bacillus pumilus E601. The strain of translucent colony had a stronger ability to resist neutron irradiation than the opaque colony. (authors)

Further study of the glassy low-temperature properties of irradiated crystalline quartz: neutron and electron irradiation

International Nuclear Information System (INIS)

Laermans, C.; Daudin, B.

1979-01-01

Recently it has been shown that a quartz crystal after light fast neutron irradiation shows low temperature hypersonic properties which are similar to those found in glasses although the sample was still crystalline. Additional measurements have been carried out in the neutron-irradiated sample and a sample irradiated with high energy electrons has also been investigated. (Fast neutron dose 6 x 10 18 n/cm 2 , 2 MeV electron dose 3 x 10 19 e/cm 2 ). A magnetic field up to 1.5 T was found to have no influence in the hypersonic saturation behaviour of the neutron-irradiated sample (9 GHz, 1.65 K) and thermal conductivity measurements are consistent with a number of two level systems (2 LS) an order of magnitude lower than in vitreous silica as found before. Low temperature hypersonic measurements as a function of acoustic intensity and temperature as well as thermal conductivity measurements give no evidence for the presence of 2 LS in the electron irradiated sample. Considering the damage created in both samples this indicates that 2 LS are probably not related to point defects

Simulation analysis of radiation fields inside phantoms for neutron irradiation

International Nuclear Information System (INIS)

Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Y.; Miyahara, N.

2007-01-01

Radiation fields inside phantoms have been calculated for neutron irradiation. Particle and heavy-ion transport code system PHITS was employed for the calculation. Energy and size dependences of neutron dose were analyzed using tissue equivalent spheres of different size. A voxel phantom of mouse was developed based on CT images of an 8-week-old male C3H/HeNs mouse. Deposition energy inside the mouse was calculated for 2- and 10-MeV neutron irradiation. (author)

Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

1990-11-01

Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

Tensile and fracture toughness test results of neutron irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

1998-01-01

Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

Irradiation-induced microstructural changes in alloy X-750

International Nuclear Information System (INIS)

Kenik, E.A.

1997-01-01

Alloy X-750 is a nickel base alloy that is often used in nuclear power systems for it's excellent corrosion resistance and mechanical properties. The present study examines the microstructure and composition profiles in a heat of Alloy X-750 before and after neutron irradiation

Tritium release from neutron irradiated beryllium pebbles

Energy Technology Data Exchange (ETDEWEB)

Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

1998-01-01

One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

Effects of neutron irradiation on red blood cell labeling with technetium-99m

International Nuclear Information System (INIS)

Eng, R.R.; Conklin, J.J.; Grissom, M.P.

1982-01-01

The effects of in vivo and in vitro neutron irradiation on red blood cell radiolabeling with technetium-99m (Tc-99m) were studied. Blood from three dogs was irradiated with neutrons (725 rads, free in air dose) followed by radiolabeling with Tc-99m. The three dogs were subsequently whole body, neutron irradiated (250 rads, midline dose); and blood samples were drawn for radiolabeling at 24, 48, 72 and 96 hours post-irradiation. Blood from three control dogs was also drawn and radiolabeled on each day for comparison. The results show that there were no significant differences between the radiolabeling capacities of in vivo or in vitro neutron irradiated and control RBCs

Self-ion emulation of high dose neutron irradiated microstructure in stainless steels

Science.gov (United States)

Jiao, Z.; Michalicka, J.; Was, G. S.

2018-04-01

Solution-annealed 304L stainless steel (SS) was irradiated to 130 dpa at 380 °C, and to 15 dpa at 500 °C and 600 °C, and cold-worked 316 SS (CW 316 SS) was irradiated to 130 dpa at 380 °C using 5 MeV Fe++/Ni++ to produce microstructures and radiation-induced segregation (RIS) for comparison with that from neutron irradiation at 320 °C to 46 dpa in the BOR60 reactor. For the 304L SS alloy, self-ion irradiation at 380 °C produced a dislocation loop microstructure that was comparable to that by neutron irradiation. No voids were observed in either the 380 °C self-ion irradiation or the neutron irradiation conditions. Irradiation at 600 °C produced the best match to radiation-induced segregation of Cr and Ni with the neutron irradiation, consistent with the prediction of a large temperature shift by Mansur's invariant relations for RIS. For the CW 316 SS alloy irradiated to 130 dpa at 380 °C, both the irradiated microstructure (dislocation loops, precipitates and voids) and RIS reasonably matched the neutron-irradiated sample. The smaller temperature shift for RIS in CW 316 SS was likely due to the high sink (dislocation) density induced by the cold work. A single self-ion irradiation condition at a dose rate ∼1000× that in reactor does not match both dislocation loops and RIS in solution-annealed 304L SS. However, a single irradiation temperature produced a reasonable match with both the dislocation/precipitate microstructure and RIS in CW 316 SS, indicating that sink density is a critical factor in determining the temperature shift for self-ion irradiations.

Microchemical evolution of neutron-irradiated stainless steel

International Nuclear Information System (INIS)

Brager, H.R.; Garner, F.A.

1980-04-01

The precipitates that develop in AISI 316 stainless steel during irradiation play a dominant role in determining the dimensional and mechanical property changes of this alloy. This role is expressed primarily in a large change in matrix composition that alters the diffusional properties of the alloy matrix and also appears to alter the rate of acceptance of point defects at dislocations and voids. The major elemental participants in the evolution have been identified as nickel, silicon, and carbon. The exceptional sensitivity of this evolution to many variables accounts for much of the variability of response exhibited by this alloy in nominally similar irradiations

Materials irradiation subpanel report to BESAC neutron sources and research panel

International Nuclear Information System (INIS)

Birtcher, R.C.; Goland, A.N.; Lott, R.

1992-01-01

The future success of the nuclear power option in the US (fission and fusion) depends critically on the continued existence of a healthy national materials-irradiation program. Consideration of the requirements for acceptable materials-irradiation systems in a new neutron source has led the subcommittee to identify an advanced steady-state reactor (ANS) as a better choice than a spallation neutron source. However, the subcommittee also hastens to point out that the ANS cannot stand alone as the nation's sole high-flux mixed-spectrum neutron irradiation source in the next century. It must be incorporated in a broader program that includes other currently existing neutron irradiation facilities. Upgrading and continuing support for these facilities must be planned. In particular, serious consideration should be given to converting the HFIR into a dedicated materials test reactor, and long-term support for several university reactors should be established

Hardening and embrittlement mechanisms of reduced activation ferritic/martensitic steels irradiated at 573 K

Energy Technology Data Exchange (ETDEWEB)

Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Hashimoto, N. [Hokkaido Univ., Materials Science and Engineering Div., Graduate School of Engineering, Sapporo (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

2007-07-01

Full text of publication follows: It has been reported that reduced-activation ferritic/martensitic steels (RAFMs), such as F82H, ORNL9Cr-2WVTa, and JLF-1, showed a variety of changes in ductile-brittle transition temperature and yield stress after irradiation at 573 K up to 5 dpa, and those differences could not be interpreted solely by the difference of dislocation microstructure induced by irradiation. To investigate the impact of other microstructural feature, i.e. precipitates, the precipitation behavior of F82H, ORNL 9Cr-2WVTa, and JLF-1 was examined. It was revealed that irradiation-induced precipitation and amorphization of precipitates partly occurred and caused the different precipitation on block, packet and prior austenitic grain boundaries. In addition to these phenomena, irradiation-induced nano-size precipitates were also observed in the matrix. It was also revealed that the chemical compositions of precipitates approached the calculated thermal equilibrium state of M{sub 23}C{sub 6} at an irradiation temperature of 573 K. The calculation also suggests the presence of Laves phase at 573 K, which is usually not observed at this temperature, but the ion irradiation on aged F82H with Laves phase suggests that Laves phase becomes amorphous and could not be stable under irradiation at 573 K. This observation indicates the possibility that the irradiation-induced nano-size precipitation could be the consequence of the conflict between precipitation and amorphization of Laves phase. Over all, these observations suggests that the variety of embrittlement and hardening of RAFMs observed at 573 K irradiation up to 5 dpa might be the consequence of the transition phenomena that occur as the microstructure approaches thermal equilibrium during irradiation at 573 K. (authors)

ATF Neutron Irradiation Program Technical Plan

Energy Technology Data Exchange (ETDEWEB)

Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

2016-03-01

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

Fast and epithermal neutron radiography using neutron irradiator

International Nuclear Information System (INIS)

Oliveira, Karol A.M. de; Crispim, Verginia R.; Ferreira, Francisco J.O.

2013-01-01

The neutron radiography technique (NR) with neutrons in the energy range fast to epithermal is a powerful tool used in no-destructive inspection of bulky objects of diverse materials, including those rich in hydrogen, oxygen, nitrogen ad carbon. Thus, it can be used to identify, inclusions, voids and thickness differences in materials such as explosive artifacts and narcotics. Aiming at using NR with fast and epithermal neutrons, an Irradiator was constructed by: a 241 Am-Be source, with 5 Ci activity, a collimator with adjustable collimation rate, L/D; and a shield device composed by plates of borated paraffin and iron. The test specimens chosen were a Beam Purity Indicator (BPI) and an Indicator of Visual Resolution (IVR). The neutron radiography images obtained had a resolution of 444.4 μm and 363.6 μm respectively when registered in: 1) the sheet of the nuclear track solid detector, CR-39 type, through X (n,p) Y nuclear reaction; and 2) Kodak Industrex M radiographic film plate in close contact with a boron converter screen, both stored in a Kodak radiographic cassette. (author)

Effect of neutron irradiation on the breakdown voltage of power MOSFET's

International Nuclear Information System (INIS)

Hasan, S.M.Y.; Kosier, S.L.; Schrimpf, R.D.; Galloway, K.F.

1994-01-01

The effect of neutron irradiation on power metal-oxide-semiconductor field effect transistor (power MOSFET) breakdown voltage has been investigated. Transistors with various breakdown voltage ratings were irradiated in a TRIGA nuclear reactor with cumulative fluence levels up to 5 x 10 14 neutrons/cm 2 (1 MeV equivalent). Noticeable increases in the breakdown voltages are observed in n-type MOSFET's after 10 13 neutrons/cm 2 and in p-type MOSFETs after 10 12 neutrons/cm 2 . An increase in breakdown voltage of as much as 30% is observed after 5 x 10 14 neutrons/cm 2 . The increase in breakdown voltage is attributed to the neutron-irradiation-induced defects which decrease the mean free path and trap majority carriers in the space charge region. The effect of positive trapped oxide charge due to concomitant gamma radiation and the effect of the termination structure on the increase in breakdown voltage are considered. An empirical model is presented to predict the value of the breakdown voltage as a function of neutron fluence

Behavior of fluorine 18 in neutron irradiated zeolites

International Nuclear Information System (INIS)

Estevez Lopez, D.R.

1992-01-01

The transformation of Li-exchanged H-Y zeolite has been investigated at 300, 550, 850 and 1050 Centigrade degree, formation of quartz structure in addition to an amorphous phase, was nited. The Li-aluminosilicate obtained was neutron irradiated and the chemical behavior of 18 F produced by the reaction sequence 6 Li (n, α) 3 H, 16 O ( 3 H, n) 18 F, was studied. The neutron irradiated material was purged with argon-hydron gas streams. It was found that the amount of released 18 F depends on the temperature used (Author)

Induction of micronuclei by irradiation with neutrons produced from 600 MeV protons

CERN Document Server

Diehl-Marshall, I

1980-01-01

Chromosome damage can be detected by the observation of micronuclei . Because its sensitivity, the micronuclei induciton in meristems of Vicia faba bean roots has been selected to determine the radiation damage produced by doses ranging from 1 to 82 cGy of neutrons. The average number of miconuclei found between 24 and 48 hours after irradiation was plotted as a funciton of dose. The RBE values, using as reference radiation 60 Co y-rays, ranged from 3.5 for a neutron dose of 82 cGy to 18 for a neutron of 1 cGy. Even higher RBE values were deduced at low doses from the evaluation of cells containing multiple micronuclei.

International Conference on Martensitic Transformations (ICOMAT 92)

Science.gov (United States)

1993-03-05

shape memory alloys are considered to be useful materials for fission and fusion reactor devices that may utilize shape memory capabilities or...restoration. Ti-Ni specimens were neutron irradiated in the Japan Materials Testing Reactors (JMTR). The irradiation conditions employed were... Triphase Steel Shen Lian and Li Hua A low carbon-low alloy triphase (F-M-A) steel containing 7- 16% retained austenite is ob- tained by composition design

Measured and evaluated fast neutron cross sections of elemental nickel

International Nuclear Information System (INIS)

Guenther, P.; Smith, A.; Smith, D.; Whalen, J.; Howerton, R.

1975-07-01

Fast neutron total and scattering cross sections of elemental nickel are measured. Differential elastic scattering cross sections are determined from incident energies of 0.3 to 4.0 MeV. The cross sections for the inelastic neutron excitation of states at: 1.156 +- 0.015, 1.324 +- 0.015, 1.443 +- 0.015, 2.136 +- 0.013, 2.255 +- 0.030, 2.449 +- 0.030, 2.614 +- 0.020 and 2.791 +- 0.025 MeV are measured to incident neutron energies of 4.0 MeV. The total neutron cross sections are determined from 0.25 to 5.0 MeV. The experimental results are discussed in the context of optical and statistical models. It is shown that resonance width-fluctuation and correlation effects are significant. The present experimental and theoretical results, together with previously reported values, are used to construct a comprehensive evaluated elemental data file in the ENDF format. Some comparisons are made with previously reported evaluated files. In addition, some selected reactions which are widely used in dosimetry and other applications are presented as supplemental evaluated isotopic-data files. The numerical quantities are presented in tabular form. (3 tables, 29 figures)

The morphology of radiation damage in copper irradiated with neutrons at elevated temperatures

International Nuclear Information System (INIS)

Kemm, K.R.

1977-01-01

This thesis is an investigation of the radiation damage morphology of high purity copper crystals irradiated with fast neutrons at temperatures in the range of 250 to 400 degrees C. At these high temperatures neutron damage is found to accumulate into large 3-dimensional rafts up to 100 μm in size, and the well known homogeneous distribution of black dot damage which is characteristic of irradiations at low temperatures is not observed. The characteristics and composition of the rafts of damage at different temperatures in the range 250 to 400 degrees C have been compared and found to differ to a large extent. It has also been shown that the background areas between rafts contain a rather low density of damage at all temperatures studied. It is therefore concluded that many of the interstitial atoms formed during irradiation migrate over large distances through the crystal lattice to precipitate at the sites of the dislocations forming the large rafts, and so denuded inter-rafts areas are left behind. It is proposed that these large rafts originate from grown-in dislocations present in the crystal before irradiation

Studies of neutron irradiation effects at IPNS-REF

International Nuclear Information System (INIS)

Kirk, M.A.

1983-09-01

Neutron irradiation effects studies at the Radiation Effects Facility (REF) at the Intense Pulsed Neutron Source (IPNS) located at Argonne National Laboratory (ANL) are reviewed. A brief history of the development of this user facility is followed by an overview of the scientific program. Experiments unique to a spallation neutron source are covered in more detail. Future direction of research at this facility is suggested

Inhibition of growth of human breast cancer cells in culture by neutron capture using liposomes containing 10B.

Science.gov (United States)

Yanagië, H; Kobayashi, H; Takeda, Y; Yoshizaki, I; Nonaka, Y; Naka, S; Nojiri, A; Shinnkawa, H; Furuya, Y; Niwa, H; Ariki, K; Yasuhara, H; Eriguchi, M

2002-03-01

Cell destruction in boron neutron capture therapy is effected by nuclear reaction between 10B and thermal neutrons with the release of alpha-particles (4He) and lithium-7 ions (7Li). 4He kills cells within 10 microm of the site of 4He generation, therefore it is theoretically possible to destroy tumour cells without affecting adjacent healthy tissue, given selective delivery of compounds containing 10B. Liposomes wore prepared by vortex dispersion of solutions containing 10B compounds with dried lipid films and the effects of those compounds on human breast cancer cells in culture were examined after thermal neutral irradiation. [3H]-TdR incorporation by MRKnu/nu-1 cells treated with 10B-containing liposomes showed 40% suppression compared with liposomes without 10B, at 2 x 1012 n/cm2 thermal neutron fluence. Inhibition of tumour cell growth with liposomes prepared with 100 mm 10B-compound was as significant as with those made with 500 ppm 10B solution. The concentration of 10B in liposomes was 76.5 +/- 3.4 microg/mL. Boronated liposomes can thus deliver sufficient 10B atoms to this line of breast cancer cells in culture to effect cytotoxicity and suppression of growth after thermal neutron irradiation.

Characterization of hybrid self-powered neutron detector under neutron irradiation

CERN Document Server

Nakamichi, M; Yamamura, C; Nakazawa, M; Kawamura, H

2000-01-01

To evaluate the irradiation behaviour of a blanket mock-up on in-pile functional test, it is necessary to measure the neutron flux change in the in-pile mock-up by a neutron detector, such as the self-powered neutron detector (SPND). With its small-sized emitter, which has high sensitivity and fast response time, SPND is an indispensable tool in order to measure the local neutron flux change. In the case of an in-pile functional test, it is necessary that response time is less than 1s and ratio of SPND output current is more than 0.3 of output current of SPND with Rh emitter. Therefore, a hybrid SPND with high sensitivity and fast response time was developed. This hybrid SPND used a hybrid emitter, i.e. Co cladded Pt-13%Rh.

Estimate of absorbed dose received by individuals irradiated with neutrons

International Nuclear Information System (INIS)

Fonseca, E.S. da; Mauricio, C.L.P.

1995-01-01

An innovating methodology is proposed to estimate the absorbed dose received by individuals irradiated with neutrons in an accident, even in the case that the victim is not using any kind of neutron dosemeter. The method combines direct measurements of 24 Na and 32 P activated in the human body. The calculation method was developed using data taken from previously published papers and experimental measurements. Other irradiations results in different neutron spectra prove the validity of the methodology here proposed. Using a whole body counter to measure 24 Na activity, it is possible to evaluate neutron absorbed doses in the order of 140 μGy of very soft (thermal) spectra. For fast neutron fields, the lower limit for neutron dose detection increases, but the present method continues to be very useful in accidents, with higher neutron doses. (author). 5 refs., 1 fig., 4 tabs

Fast neutron irradiation deteriorates hippocampus-related memory ability in adult mice.

Science.gov (United States)

Yang, Miyoung; Kim, Hwanseong; Kim, Juhwan; Kim, Sung-Ho; Kim, Jong-Choon; Bae, Chun-Sik; Kim, Joong-Sun; Shin, Taekyun; Moon, Changjong

2012-03-01

Object recognition memory and contextual fear conditioning task performance in adult C57BL/6 mice exposed to cranial fast neutron irradiation (0.8 Gy) were examined to evaluate hippocampus-related behavioral dysfunction following acute exposure to relatively low doses of fast neutrons. In addition, hippocampal neurogenesis changes in adult murine brain after cranial irradiation were analyzed using the neurogenesis immunohistochemical markers Ki-67 and doublecortin (DCX). In the object recognition memory test and contextual fear conditioning, mice trained 1 and 7 days after irradiation displayed significant memory deficits compared to the sham-irradiated controls. The number of Ki-67- and DCX-positive cells decreased significantly 24 h post-irradiation. These results indicate that acute exposure of the adult mouse brain to a relatively low dose of fast neutrons interrupts hippocampal functions, including learning and memory, possibly by inhibiting neurogenesis.

Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

Energy Technology Data Exchange (ETDEWEB)

Korenev, Sergey E-mail: sergey_korenev@steris.com; Sikolenko, Vadim

2004-10-01

The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

Science.gov (United States)

Korenev, Sergey; Sikolenko, Vadim

2004-09-01

The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

Neutron dosimetry in containment of a pressurized water reactor utilizing the Panasonic UD-802 dosimetry system

International Nuclear Information System (INIS)

Kralick, S.C.

1984-01-01

The Panasonic UD-802 dosimeter was evaluated as a potential neutron dosimeter for use in containment of a PWR. The Panasonic UD-802 dosimeter, although designed as a beta and gamma dosimeter, is also sensitive to neutrons. UD-802 dosimeters were mounted on polyethylene phantoms and irradiated to known doses at selected locations in containment. The known neutron dose equivalents were determined based on remmeter dose rate measurements and stay times. The thermoluminescent response of the dosimeters and the known neutron dose equivalents were used to obtain a calibration factor at each location. The average calibration factor was 3.7 (unit of dosimeter response per mrem) and all calibration factors were within +-30% of this mean value. The dosimeter distance from the phantom was found to have minimal effect on the response but the system was directionally dependent, necessitating a correction in the calibration factor. The minimum significant dosimeter response was determined independent of any calibration factor. The minimum significant response of the UD-802 to neutrons is a function of the corresponding gamma exposure rate. It is concluded that the Panasonic UD-802 dosimeter can be used for neutron dosimetry in PWR containment

Irradiation system for neutron capture therapy using the small accelerator

International Nuclear Information System (INIS)

Kobayashi, Tooru; Hoshi, Masaharu

2002-01-01

Neutron capture therapy (NCT) is to kill tumor cells that previously incorporated the stable isotope which generates heavy charged particles with a short range and a high linear energy transfer (LET) on neutron irradiation. Boron-10 is ordinarily used as such an isotope. The tumor tissue is neutron-irradiated at craniotomy after preceding craniotomy for tumor extraction: therefore two surgeries are required for the present NCT in Japan. The reactions 10 B(n, αγ) 7 Li and 7 Li (p, n) 7 Be are thought preferential for patients and doctors if a convenient small accelerator, not the reactor used at present, is available in the hospital because only one craniotomy is sufficient. Authors' examinations of the system for NCT using the small accelerator involve irradiation conditions, desirable energy spectrum of neutron, characterization of thermal and epi-thermal neutrons, social, practical and technical comparison of the reactor and accelerator, and usefulness of the reaction 7 Li (p, n) 7 Be. The system devoted to the NCT is awaited in future. (K.H.)

Morphological profiles of neutron and X-irradiated small intestine

International Nuclear Information System (INIS)

Carr, K.E.; O'Shea, O.; Hazzard, R.A.; McCullough, J.S.; Hume, S.P.; Nelson, A.C.

1996-01-01

This paper describes the response of mouse small intestine, at several time points after treatment with neutron or X-irradiation, using doses expected to give similar effects in terms of crypt/microcolony survival. Using resin histology, the effects of radiation on the numbers of duodenal cell types and measurements of tissue areas were assessed. The results for individual parameters and for an estimate of overall damage are given in a data display, which summarises the morphological profile of the organ after both types of radiation. Damage and recovery were seen for many of the parameters studied but there was no standard response pattern applicable for all parameters. In particular, the response of individual crypt cell types could not be predicted from knowledge of the change in crypt numbers. With regard to the holistic response of the gut, neutron irradiation appeared to have caused more damage and produced more early effects than the X-irradiation. More specifically, neutron treatment led to more damage to the neuromuscular components of the wall, while X-irradiation produced early vascular changes. (author)

Apoptosis of nasopharyngeal carcinoma cell line (CNE-2) induced by neutron irradiation

International Nuclear Information System (INIS)

Liang Ke; He Shaoqin; Feng Yan; Tang Jinhua; Feng Qinfu; Shen Yu; Yin Weibo; Xu Guozhen; Liu Xinfan; Wang Luhua; Gao Li

1999-01-01

Objective: To study the apoptotic response of the nasopharyngeal carcinoma cell line (CNE-2) induced by neutron irradiation. Methods: CNE-2 cells were cultured as usual. Using the techniques of DNA agarose gel electrophoresis and DNA special fluorescent staining, the status of apoptosis in CNE-2 cells after neutron irradiation was detected. Results: It was shown that the apoptosis can be induced in CNE-2 cell after neutron radiation. Six hrs, after different doses of neutron (0/0.667/1.333/2.000/2.667/3.333 Gy) and X-ray 0/2/4/6/8/10 Gy) irradiation the apoptotic rates were 2.4%, 6.3%, 7.1%, 9.5%, 13.5%, 14.6% and 2.4%, 3.8%, 5.7%, 7.8%, 10.4%, 11.7%, respectively; at 48 hrs they were 18.3%, 21.5%, 22.8%, 29.3%, 34.2% and 13.7%, 17.6%, 21.3%, 25.6%, 28.9%, respectively. At 10 hrs after neutron irradiation the DNA ladder of apoptosis could be detected between 0.667-3.333 Gy doses in CNE-2 cells by DNA agarose gel electrophoresis. Conclusion: Neutron radiation can induce apoptosis in tumor cells. Compared with the X-ray, neutron induces apoptosis in larger extent than X-ray in the same condition; meanwhile, apoptosis after irradiation is dose and time dependent

Measurement and evaluation of fast neutron flux of CT and OR5 irradiation hole in HANARO

International Nuclear Information System (INIS)

Yang, Seong Woo; Choo, Kee Nam; Lee, Seung-Kyu; Kim, Yong Kyun

2012-01-01

The irradiation test has been conducted to evaluate the irradiation performance of many materials by a material capsule at HANARO. Since the fast neutron fluence above 1 MeV is important for the irradiation test of material, it must be measured and evaluated exactly at each irradiation hole. Therefore, a fast neutron flux was measured and evaluated by a 09M-02K capsule irradiated in an OR5 irradiation hole and a 10M-01K capsule irradiated in a CT irradiation hole. Fe, Ni, and Ti wires as the fluence monitor were used for the detection of fast neutron flux. Before the irradiation test, the neutron flux and spectrum was calculated for each irradiation hole using an MCNP code. After the irradiation test, the activity of the fluence monitor was measured by an HPGe detector and the reaction rate was calculated. For the OR5 irradiation hole, the radial difference of the fast neutron flux was observed from a calculated data due to the OR5 irradiation hole being located outside the core. Furthermore, a control absorber rod was withdrawn from the core as the increase of the irradiation time at the same irradiation cycle, so the distribution of neutron flux was changed from the beginning to the end of the cycle. These effects were considered to evaluate the fast neutron flux. Neutron spectrums of the CT and OR5 irradiation hole were adjusted by the measured data. The fluxes of a fast neutron above 1 MeV were compared with calculated and measured value. Although the maximum difference was shown at 18.48%, most of the results showed good agreement. (author)

Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

Energy Technology Data Exchange (ETDEWEB)

Franco, Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-08-01

The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential

A Study on the Thermal Neutron Filter for the Irradiation of Electronic Materials at HANARO

Energy Technology Data Exchange (ETDEWEB)

Yang, Seong Woo; Kim, Sung Ryul; Park, Seung Jae; Shin, Yoon Taeg; Cho, Man Soon; Cho, Kee Nam [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The representative example is a technique of making the semiconductor with the transmutation using the pure Si. This NTD (Neutron Transmutation Doping) Si is used as a high-quality semiconductor because it has a uniform resistance. Likewise, the electronic materials are being investigated to improve the performance of material using the neutron irradiation method. The mechanism for reaction between the electronic materials and the neutrons depends on the energy of the neutron. Capturing reaction by thermal neutrons causes the transmutation and a lot of defects are made by fast neutrons. The study for the effect by such neutron energy is necessary to understand the performance improvement of the irradiated electronic materials. The thermal neutron filter was investigated to be used for the irradiation of electronic materials at HANARO. IP irradiation hole was selected and the irradiation device was designed. The analysis was conducted considering four candidate materials.

Neutron irradiation effects on magnetic properties of some Heusler alloys

International Nuclear Information System (INIS)

Onodera, Hideya; Shinohara, Takeshi; Yamamoto, Hisao; Watanabe, Hiroshi

1975-01-01

The neutron irradiation effects were studied with measurements of temperature dependence of magnetization in ordered and disordered Heusler alloys. The irradiation was carried out in JMTR with a total flux of fast neutrons of 10 20 nvt. Fully ordered Cu 2 MnIn, partially ordered Cu 2 MnAl and completely disordered Cu 2 MnSn were prepared with various temperature treatments. The magnetization-temperature curves of each specimen were measured before and after irradiation. In the irradiated Cu 2 MnIn, the disordering by the irradiation gave rise to a decrease of magnetization, and the temperature dependence of magnetization showed that the disordered region contained various regions with different degrees of disorder. For the distribution of the disordered region, the calculation based on the theory of temperature spike by Seitz and Koekler gave a feasible result that a disordered region comprised a central core with a radius of 5.4 A which was completely disordered and a periphery of 3.3 A thickness which was partially disordered. From the magnetization-temperature curves of Cu 2 MnAl, it was considered that the disordered regions induced by the irradiation had different properties from those induced by the heat treatment. The former were the localized and comprised regions corresponding to various degrees of disorder, while the latter spread spatially in a wide range with a certain degree of disorder. The ordering by enhanced diffusion occurred simultaneously to an extent comparable to the disordering, and so it played an important role in the magnetization in the partially disordered Cu 2 MnAl. In the disordered Cu 2 MnSn, however, the ordering effect was very small. It is supposed to be difficult for the A2 structure to transform into the L2 1 structure by the enhanced diffusion. (auth.)

Cation disorder in high-dose, neutron-irradiated spinel

International Nuclear Information System (INIS)

Sickafus, K.E.; Larson, A.C.; Yu, N.; Nastasi, M.; Hollenberg, G.W.; Garner, F.A.; Bradt, R.C.

1994-08-01

The objective of this effort is to determine whether MgAl 2 O 4 spinel is a suitable ceramic for fusion applications. Here, the crystal structures of MgAl 2 O 4 spinel single crystals irradiated to high neutron fluences [>5·10 26 n/m 2 (E n > 0.1 MeV)] were examined by neutron diffraction. Crystal structure refinement of the highest dose sample indicated that the average scattering strength of the tetrahedral crystal sites decreased by ∼ 20% while increasing by ∼ 8% on octahedral sites. Since the neutron scattering length for Mg is considerably larger than for Al, this results is consistent with site exchange between Mg 2+ ions on tetrahedral sites and Al 3+ ions on octahedral sites. Least-squares refinements also indicated that, in all irradiated samples, at least 35% of Mg 2+ and Al 3+ ions in the crystal experienced disordering replacements. This retained dpa on the cation sublattices is the largest retained damage ever measured in an irradiated spinel material

Void formation by annealing of neutron-irradiated plastically deformed molybdenum

International Nuclear Information System (INIS)

Petersen, K.; Nielsen, B.; Thrane, N.

1976-01-01

The positron annihilation technique has been used in order to study the influence of plastic deformation on the formation and growth of voids in neutron irradiated molybdenum single crystals treated by isochronal annealing. Samples were prepared in three ways: deformed 12-19% before irradiation, deformed 12-19% after irradiation, and - for reference purposes -non-deformed. In addition a polycrystalline sample was prepared in order to study the influence of the grain boundaries. All samples were irradiated at 60 0 C with a flux of 2.5 x 10 18 fast neutrons/cm 2 . After irradiation the samples were subjected to isochronal annealing. It was found that deformation before irradiation probably enhanced the formation of voids slightly. Deformation after irradiation strongly reduced the void formation. The presence of grain boundaries in the polycrystalline sample had a reducing influence on the growth of voids. (author)

Seed irradiation with continuously increasing doses of thermal neutrons

International Nuclear Information System (INIS)

Uhlik, J.; Pfeifer, M.; Pittermann, P.

1977-01-01

In the 'Raman' pea cv. the biological activity of thermal neutrons was investigated after irradiation of a 780 mm column of seeds for 3000 and 4167 seconds with a flux of 5.607 x 10 9 n.cm -2 per second. For different fractions of the seed column the average density of the neutron flux was calculated. It was proved that for the described method of seed irradiation it was sufficient to determine only the dose approaching the lethal dose. If a sufficiently high column of seeds is used part of the column of seeds will be irradiated with the optimum range of doses. The advantages of the suggested method of irradiation are not only smaller time and technological requirements resulting from the need for the determination of only the critical lethal dose of radiation by means of inhibition tests performed with seedlings, but also a simpler irradiation procedure. The suggested method of irradiation is at least nine times cheaper. (author)

Anisotropic shift of the irreversibility line by neutron irradiation

International Nuclear Information System (INIS)

Sauerzopf, F.M.; Wiesinger, H.P.; Weber, H.W.; Crabtree, G.W.; Frischherz, M.C.; Kirk, M.A.

1991-09-01

The irreversibility line of high-T c superconductors is shifted considerably by irradiating the material with fast neutrons. The anisotropic and non-monotonous shift is qualitatively explained by a simple model based on an interaction between three pinning mechanisms, the intrinsic pinning by the ab-planes, the weak pinning by the pre-irradiation defect structure, and strong pinning by neutron induced defect cascades. A correlation between the cascade density and the position of the irreversibility line is observed

Aluminium alloys containing iron and nickel

International Nuclear Information System (INIS)

Coriou, H.; Fournier, R.; Grall, L.; Hure, J.; Herenguel, J.; Lelong, P.

1958-01-01

The first part of this report addresses mechanism, kinetics and structure factors of aluminium alloys containing iron and nickel in water and high temperature steam. The studied alloys contain from 0.3 to 0.7 per cent of iron, and 0.2 to 1.0 per cent of nickel. Corrosion resistance and corrosion structure have been studied. The experimental installation, process and samples are presented. Corrosion structures in water at 350 C are identified and discussed (structure of corrosion products, structure of metal-oxide interface), and then in steam at different temperatures (350-395 C). Corrosion kinetics is experimentally studied (weight variation in time) in water at 350 C and in steam at different temperatures. Reactions occurring at over-heated steam (more than 400 C) are studied, and the case of welded alloys is also addressed. The second part addresses the metallurgical mechanism and processes influencing aluminium alloy resistance to corrosion by high temperature water as it appeared that separated phases protect the solid solution through a neighbourhood action. In order to avoid deep local corrosions, it seems necessary to multiply protective phases in an as uniform as possible way. Some processes enabling this result are described. They belong to conventional metallurgy or to powder metallurgy (with sintering and extrusion)

Report of IEA workshop on reduced activation ferritic/martensitic steels

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

IEA Workshop on Reduced Activation Ferritic/Martensitic Steels under implementing agreement for program of research and development on fusion materials was held at Tokyo Yayoi Kaikan and JAERI headquarter on November 2-3, 2000. The objective of this workshop was a review of the fusion material development programs, the progress of the collaboration and the irradiation effects studies on RAF/M steels in the collaborating parties (Europe, Russia the United States, and Japan). Moreover, the development of plans for future collaboration was discussed. The present report contains viewgraphs presented at the workshop. (author)

Magnetic interactions in martensitic Ni-Mn based Heusler systems

Energy Technology Data Exchange (ETDEWEB)

Aksoy, Seda

2010-04-22

In this work, magnetic, magnetocaloric and structural properties are investigated in Ni-Mn-based martensitic Heusler alloys with the aim to tailor these properties as well as to understand in detail the magnetic interactions in the various crystallographic states of these alloys. We choose Ni{sub 50}Mn{sub 34}In{sub 16} as a prototype which undergoes a martensitic transformation and exhibits field-induced strain and the inverse magnetocaloric effect. Using the structural phase diagram of martensitic Ni-Mn-based Heusler alloys, we substitute gallium and tin for indium to carry these effects systematically closer to room temperature by shifting the martensitic transformation. A magneto-calorimeter is designed and built to measure adiabatically the magnetocaloric effect in these alloys. The temperature dependence of strain under an external magnetic field is studied in Ni{sub 50}Mn{sub 50-x}Z{sub x} (Z: Ga, Sn, In and Sb) and Ni{sub 50}Mn{sub 34}In{sub 16-x}Z{sub x} (Z: Ga and Sn). An argument based on the effect of the applied magnetic field on martensite nucleation is adopted to extract information on the direction of the magnetization easy axis in the martensitic unit cell in Heusler alloys. Parallel to these studies, the structure in the presence of an external field is also studied by powder neutron diffraction. It is demonstrated that martensite nucleation is influenced by cooling the sample under a magnetic field such that the austenite phase is arrested within the martensitic state. The magnetic interactions in Ni{sub 50}Mn{sub 37}Sn{sub 13} and Ni{sub 50}Mn{sub 40}Sb{sub 10} are characterized by using neutron polarization analysis. Below the martensitic transformation temperature, M{sub s}, an antiferromagnetically correlated state is found. Ferromagnetic resonance experiments are carried out on Ni{sub 50}Mn{sub 37}Sn{sub 13} and Ni{sub 50}Mn{sub 34}In{sub 16} to gain more detailed information on the nature of the magnetic interactions. The experimental

Irradiation creep lifetime analysis on first wall structure materials for CFETR

Energy Technology Data Exchange (ETDEWEB)

Ma, Bing; Peng, Lei, E-mail: penglei@ustc.edu.cn; Zhang, Xiansheng; Shi, Jingyi; Zhan, Jie

2017-05-15

Fusion reactor first wall services on the conditions of high surface heat flux and intense neutron irradiation. For China Fusion Engineering Test Reactor (CFETR) with high duty time factor, it is important to analyze the irradiation effect on the creep lifetime of the main candidate structure materials for first wall, i.e. ferritic/martensitic steel, austenite steel and oxide dispersion strengthened steel. The allowable irradiation creep lifetime was evaluated with Larson-Miller Parameter (LMP) model and finite element method. The results show that the allowable irradiation creep lifetime decreases with increasing of surface heat flux, first wall thickness and inlet coolant temperature. For the current CFETR conceptual design, the lifetime is not limited by thermal creep or irradiation creep, which indicated the room for design parameters optimization.

Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material