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Journal Articles

Evaluation of irradiation-induced point defect migration energy during neutron irradiation in modified 316 stainless steel

Sekio, Yoshihiro; Yamagata, Ichiro; Akasaka, Naoaki; Sakaguchi, Norihito*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 8 Pages, 2017/06

The widths of void denuded zones (VDZs) which were formed near random grain boundaries by neutron irradiation were analyzed in order to perform quantitative evaluations for the irradiation-induced point defect behavior in the modified 316 stainless steel (PNC316) having been developed by JAEA. Namely, the temperature dependence of VDZ width was investigated and vacancy migration energy of the PNC316 steel was estimated from the VDZ width analysis for the neutron-irradiated specimens. The obtained value of vacancy migration energy was estimated as 1.46 eV, which was consistent with that from the exiting method using electron in-situ examination. This indicates that VDZ analysis could be effective method to evaluate especially vacancy migration energy during irradiation, and this would be realized from not in-situ observation but post-irradiation examination in the case of neutron irradiation.

JAEA Reports

Evaluation of irradiation behavior on oxide dispersion strengthened (ODS) steel claddings irradiated in Joyo/CMIR-6

Yano, Yasuhide; Otsuka, Satoshi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Sekine, Manabu; Endo, Toshiaki; Yamagata, Ichiro; Sekio, Yoshihiro; Tanno, Takashi; Uwaba, Tomoyuki; et al.

JAEA-Research 2013-030, 57 Pages, 2013/11

JAEA-Research-2013-030.pdf:48.2MB

It is necessary to develop the fast reactor core materials, which can achieve high-burnup operation improving safety and economical performance. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, the effects of fast neutron irradiation on mechanical properties and microstructure of 9Cr-and 12Cr-ODS steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the CMIR-6 at temperatures between 420 and 835$$^{circ}$$C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.

Journal Articles

Effects of neutron irradiation on tensile properties of oxide dispersion strengthened (ODS) steel claddings

Yano, Yasuhide; Ogawa, Ryuichiro; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Tanaka, Kenya

Journal of Nuclear Materials, 419(1-3), p.305 - 309, 2011/12

 Times Cited Count:14 Percentile:76.2(Materials Science, Multidisciplinary)

The effects of fast neutron irradiation on ring tensile properties of oxide dispersion strengthened (ODS) steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the material irradiation rig at temperatures between 693 and 1108 K to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures. The experimental results showed that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 33 dpa were decreased by about 20%. On the other hand, uniform elongation after irradiation was more than 2% at all irradiation conditions. The ring tensile properties of these ODS claddings remained excellent within these irradiation conditions compared with conventional 11Cr ferritic/martensitic steel (PNC-FMS) claddings.

Journal Articles

Swelling behaviors in a fuel assembly for the wrapping wire and duct made of modified 316 austenitic stainless steel

Yamagata, Ichiro; Akasaka, Naoaki

Journal of Nuclear Science and Technology, 47(10), p.898 - 907, 2010/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The swelling behaviors in wrapping wire and duct were investigated for a fuel assembly made of modified type 316 austenitic stainless steel irradiated in a fast breeder reactor. The temperature dependence of swelling varied because the peak temperatures of swelling in the wrapping wire and the duct were different. The void distribution in the material was observed by scanning electron microscopy and transmission electron microscopy and it was confirmed that the voids grew within an area of about 100 $$mu$$m from the surface. This phenomenon seemed to be caused by a surface effect on the neutron-irradiated materials.

Journal Articles

Mechanical properties and microstructural stability of 11Cr-ferritic/martensitic steel cladding under irradiation

Yano, Yasuhide; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Shibayama, Tamaki*; Watanabe, Seiichi*; Takahashi, Heishichiro

Journal of Nuclear Materials, 398(1-3), p.59 - 63, 2010/03

 Times Cited Count:9 Percentile:56.9(Materials Science, Multidisciplinary)

The in-reactor creep rupture tests of 11Cr-0.5Mo-2W, V, Nb F/M steel were carried out in the temperature range from 823 to 943 K using Materials Open Test Assembly in the Fast Flux Test Facility and tensile and temperature-transient-to-burst specimens were irradiated in the experimental fast reactor JOYO at temperatures between 693 to 1013 K to fast neutron doses ranging from 3.5 to 102 dpa. The results of post irradiation mechanical tests showed that there was no significant degradation in tensile and transient burst strengths even after neutron irradiation below 873 K, but that there was significant degradation in both strengths at neutron irradiation above 903 K. On the other hand, the in-reactor creep rupture times were equal or greater than those of out-reactor creep even after neutron irradiation at all temperatures. This creep rupture behavior was different from that of tensile and transient burst specimens.

Journal Articles

Effect of high dose/high temperature irradiation on the microstructure of heat resistant 11Cr ferritic/martensitic steels

Yamashita, Shinichiro; Yano, Yasuhide; Tachi, Yoshiaki; Akasaka, Naoaki

Journal of Nuclear Materials, 386-388, p.135 - 139, 2009/04

 Times Cited Count:10 Percentile:60.19(Materials Science, Multidisciplinary)

The heat resistant 11Cr ferritic/martensitic steels were irradiated at 400-670 $$^{circ}$$C up to 100 dpa in FFTF and JOYO. The microstructures of unirradiated 11Cr ferritic/martensitic steels consist of laths, dislocation, and carbide. Almost of the prior austenitic boundaries (PABs) were partially decorated with carbides. It was observed from the results of post irradiation microstructural examinations that the irradiation-induced microstructures were classified into the following three types depending on irradiation temperature; (1) When irradiated at 400-450 $$^{circ}$$C, both dislocation loops and cavities with less than 30 nm in diameter were formed in the ferrite phase. On the other hand, the void swelling was about 0.05%. (2) In the case of irradiation at moderate temperature (500-600 $$^{circ}$$C), the precipitates formation M$$_{23}$$C$$_{6}$$ carbide was primarily dominated. It was a most noticeable microstructural feature that the carbides; M$$_{23}$$C$$_{6}$$ and M$$_{6}$$C grew and covered the PABs at this temperature range. (3) Finally, when irradiation temperature was above 650 $$^{circ}$$C microstructures were drastically-changed. Microstructural observations revealed that formation and growth of equi-axial grain occurred in addition to recovery of laths, growth of carbides simultaneously at high temperature. This remarkable microstructural change might be closely related to a severe degradation in the mechanical properties.

Journal Articles

Effect of fine precipitate on void swelling behavior in heavily-irradiated modified SUS316 stainless steels for fast reactor application

Yamashita, Shinichiro; Sekine, Manabu*; Akasaka, Naoaki

Materia, 47(12), P. 624, 2008/12

no abstracts in English

Journal Articles

Effects of microstructural evolution on mechanical properties of 11Cr ferritic/martensitic steel after neutron irradiation

Yano, Yasuhide; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Takahashi, Heishichiro

Materia, 47(12), P. 625, 2008/12

no abstracts in English

Journal Articles

Effects of fast reactor irradiation conditions on tensile and transient burst properties of ferritic/martensitic steel claddings

Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Watanabe, Seiichi*; Takahashi, Heishichiro

Journal of Nuclear Science and Technology, 44(12), p.1535 - 1542, 2007/12

 Times Cited Count:11 Percentile:63.76(Nuclear Science & Technology)

The effects of fast neutron irradiation have been investigated on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated in the experimental fast reactor JOYO using the PFB090 fuel test assembly. Post irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. However, these strengths for HT9M cladding tended to shift to lower values than those of as-received specimens. This different behavior of tensile and transient burst strengths was attributed to martensite structural stability which was related to the stable solid solution elements.

Journal Articles

Structure of nano-size oxides in ODS steels and its stability under electron irradiation

Oka, Keiichiro*; Onuki, Somei*; Yamashita, Shinichiro; Akasaka, Naoaki; Otsuka, Satoshi; Tanigawa, Hiroyasu

Materials Transactions, 48(10), p.2563 - 2566, 2007/10

 Times Cited Count:29 Percentile:81.12(Materials Science, Multidisciplinary)

For understanding the microstructural details of nano-size oxide particles, three types of ODS ferritic and austenitic steels were examined by high voltage electron microscopy, EDS and AP-FIM. The oxide included Y, Ti and O and showed a shell-like structure with different composition. The shell-like structure depends on crystal structure of the matrix during fabrication process. To evaluate the irradiation stability of the oxide particles, the electron irradiation was carried out to 47 dpa in the temperature range between room temperature and 923 K. During the irradiation, the oxide particles did not show obvious change in size. The irradiation behavior is discussed comparing with the results recently reported.

Journal Articles

Mechanical properties and microstructural stability of advanced ferritic/martensitic steel under irradiation

Yano, Yasuhide; Yamashita, Shinichiro; Akasaka, Naoaki; Watanabe, Seiichi*; Takahashi, Heishichiro

Proceedings of 9th China-Japan Symposium on Materials for Advanced Energy Systems and Fission & Fusion Engineering jointed with CAS-JSPS Core-university Program Seminar on Fusion Materials, System and Design Integration, p.2 - 5, 2007/10

Ferritic/martensitic (F/M) steels are expected to be prospective not only for the long life core material of fast reactors but also for the blanket materials of fusion reactor because of their superior swelling resistance. Japan Atomic Energy Agency has developed a 11Cr-0.5Mo-2W-V, Nb F/M steel (PNC-FMS) for core materials of next fast reactor. In order to evaluate the effect of structural change due to irradiation on mechanical properties of PNC-FMS, neutron irradiations were carried out between 773 and 1013 K to doses of from 11 to 102 dpa in JOYO reactor. Post irradiation tensile tests were performed at 773-1013 K with a strain rate of 0.5$$times$$10$$^{-4}$$/s. The stability of microstructures under irradiation was also compared with those of electron irradiation using HVEM.

Journal Articles

Neutron irradiation effects on $$^{11}$$B$$_{4}$$C and recovery by annealing

Donomae, Takako; Tachi, Yoshiaki; Sekine, Manabu*; Morohashi, Yuko; Akasaka, Naoaki; Onose, Shoji

Journal of the Ceramic Society of Japan, 115(1345), p.551 - 555, 2007/09

 Times Cited Count:4 Percentile:30.7(Materials Science, Ceramics)

Use of moderator materials in Fast Breeder Reactor (FBR) is effective for transmutation technology, and $$^{11}$$B$$_{4}$$C is one of the candidates. Up to now, the behavior of $$^{10}$$B$$_{4}$$C as the Control rod material is well known, but that of $$^{11}$$B$$_{4}$$C is hardly investigated. In this paper, the radiation effects of $$^{11}$$B$$_{4}$$C pellets, neutron irradiated in the experimental fast reactor JOYO were studied. From the experimental results, it was observed that no macro-cracks were recognized in the irradiated $$^{11}$$B$$_{4}$$C pellets. But, bubble nucleation was found in grain and along grain boundaries of $$^{11}$$B$$_{4}$$C. And, it was shown that the conductivity of $$^{11}$$B$$_{4}$$C was higher than that of $$^{10}$$B$$_{4}$$C. During the annealing from room temperature to 1400$$^{circ}$$C, three recovery stages were found on thermal conductivity. It was suggested that, the recovery of B$$_{4}$$C was related to the dispersion behavior of helium. Judging from these results, as $$^{11}$$B$$_{4}$$C was mechanically more stable compared with $$^{10}$$B$$_{4}$$C under irradiation, it was shown that $$^{11}$$B$$_{4}$$C had high applicability for a moderator.

Journal Articles

Tensile and transient burst properties of advanced ferritic/martensitic steel claddings after neutron irradiation

Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Takahashi, Heishichiro*

Journal of Nuclear Materials, 367-370(1), p.127 - 131, 2007/08

 Times Cited Count:10 Percentile:60.87(Materials Science, Multidisciplinary)

The effects of fast neutron irradiation on tensile and transient burst properties of advanced ferritic/martensitic steel claddings were investigated. Specimens were irradiated in the experimental fast reactor JOYO using the material irradiation rig at temperatures between 773 and 1013 K to fast neutron doses ranging from 11 to 102 dpa. The post-irradiation tensile and temperature-transient-to-burst tests were carried out. The results of mechanical tests showed that there was no significant degradation in tensile and transient burst strengths after neutron irradiation below 873 K. This was attributed to grain boundary strengthening caused by precipitates that preferentially formed on prior-austenite grain boundaries. Both strengths at neutron irradiation above about 903 K up to 102 dpa decreased due to recovery of lath martensite structures and recrystallization.

Journal Articles

Microstructural development of a heavily neutron-irradiated ODS ferritic steel (MA957) at elevated temperature

Yamashita, Shinichiro; Akasaka, Naoaki; Ukai, Shigeharu; Onuki, Somei*

Journal of Nuclear Materials, 367-370(1), p.202 - 207, 2007/08

 Times Cited Count:64 Percentile:97.19(Materials Science, Multidisciplinary)

Microstructural observation was done on a neutron-irradiated oxide dispersion strengthened (ODS) ferritic steel, MA957. Since MA957 has been investigated from various viewpoints, special emphases in this study were laid on oxide behaviors including phase stability under irradiation at elevated temperature ($$sim$$973 K). Transmission electron microscopy (TEM) observation of the Y-Ti complex oxide particles showed they were fine ($$sim$$40 nm) whereas the Ti-oxide particles were relatively coarse ($$sim$$300 nm). Dispersion parameters of oxide particles, such as mean size and number density, changed due to irradiation. This fact implies that the recoil resolution of the oxide particles. When irradiated at 973 K, some Y-Ti complex oxides were surviving and interacted with the dislocation structures, which delayed the dislocation recovery and consequently stabilized the elongated grain structures. This is the first evidence showing that oxide particles are effectively functioning as pinning points of dislocations in motion under irradiation to a dose of $$sim$$100 dpa.

Journal Articles

Effects of microstructural evolution on charpy impact properties of modified ferritic/martensitic steel after neutron irradiation

Yano, Yasuhide; Oka, Keiichiro*; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro; Onuki, Somei

Journal of Nuclear Science and Technology, 43(6), p.648 - 654, 2006/06

 Times Cited Count:4 Percentile:32.55(Nuclear Science & Technology)

The embrittlement behavior of the modified ferritic/martensitic 11Cr-0.5Mo-2W, V, Nb steel (PNC-FMS) after neutron irradiation in JOYO was investigated by Charpy impact tests and TEM and SEM observations. The impact properties of the specimens after irradiation to 4.4 dpa at 773 K were similar to the as-received PNC-FMS. The ductile-brittle transition temperature (DBTT) remarkably decreased due to irradiation to 2.8 dpa at 923 K. The precipitates formed in the martensitic lath were still stable under neutron irradiation at 773 K, however they were unstable under irradiation at 923 K. The martensitic lath structure was also stable at the former irradiation temperature, but it was significantly changed at the latter. The decrease in the upper shelf energy after irradiation was related to the precipitate distribution. The changes of DBTT due to irradiation were attributed to decreased the dislocation recovery and to increased broadening of the martensitic lath.

Journal Articles

Interaction among dislocation and complex oxide particles in ODS steels haevily-irradiated at high temperature

Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji

Zairyo Kaihatsu No Tameno Kenbikyoho To Oyo Shashinshu, P. 133, 2006/03

Oxide Dispersion Strengthened (ODS) ferritic steel dealt with this study was a MA957 (Fe-0.015C-14Cr-0.3Mo-1.0Ti-0.25Y2O3). The objectives of this study were to understand oxide particle stability of ODS steel during irradiation and interaction among dislocation and oxide particles, reflecting to advanced nuclear reactor design of next generation. Development of some nuclear energy generating systems has been proposed and supported intensively under several international collaboration programs (Generation IV International Forum (GIF), Advanced Fuel Cycle Initiative (AFCI), International Nuclear Energy Research Initiative (I-NERI) etc).Current research issue on ODS ferritic steels is considered to be poverty of experience and understanding on their practical neutron-irradiation behaviors at the temperature higher than 600C.In this research, a MA957, most familiar but primitive 14CrODS ferritic steel contained the highly textured-anisotropic grain structures, was irradiated at 500-700$$^{circ}$$C to fast fluences ranging from 19.8 to 20.8 $$times$$ 1026 n/m2 (E $$>$$ 0.1MeV) in the experimental fast reactor JOYO. The dose achieved varied from 99 to 104 dpa. TEM observation and micro-hardness measurement were carried out to clarify the irradiation effects on microstructural evolution of 14CrODS ferritic steel at elevated temperature and high dose. Microstructural examination revealed that all of the highly textured- anisotropic grain structures, following heavy irradiation at the temperature above 600$$^{circ}$$C, have not changed. In addition, large regions in all specimens have retained high dislocation density, contained negligible cavitation.

JAEA Reports

Microstructural evolution of ODS steels under neutron irradiation

Yamashita, Shinichiro; Akasaka, Naoaki

JNC TN9400 2005-018, 54 Pages, 2005/05

JNC-TN9400-2005-018.pdf:10.85MB

ODS steels have high resistance to radiation damage and superior long-term thermal-mechanical strengths at high temperature, offering a promise of high performance fuel cladding tube for advanced sodium-cooled fast reactor and being developed intensively in JNC. This report covered microstructural evolutions of ODS steels (1DS and 1DK developed in 1989 and M93, F94, F95 in 1997, respectively) irradiated in JOYO to evaluate the irradiation properties of several type of ODS steels. The new findings obtained were as the followings.1) Dislocation loops and cavities formed under neutron irradiation, depending on irradiation temperature and sink site density prior to irradiation such as dislocation density, total area of interface between precipitate and matrix, total area of grain boundary. 2) Comparison of microstructural evolution between F94 and F95, each of which were mechanically alloyed under different inert gases, indicated that argon is less significant for cavity formation (including its nucleation and growth) than helium. 3) Neither alpha prime nor shigma phases were formed in all of the ODS steels dealt in this study. In the case of high tungsten ODS steels (1DS, 1DK), Fe$$_{2}$$W type of Laves phase precipitated preferentially at grain boundaries and elevated temperatures under irradiation. As for carbide, M$$_{23}$$C$$_{6}$$ (M=Cr) was a major precipitate of the martensitic ODS steel (M93) before and after irradiation. In ferritic ODS steels (F94, F95), formation of several types of carbides coupled with Ti, W, Cr, Fe were recognized before and after irradiation, but volume of each carbides was very low, being reflected by a carbon content of the pre-irradiated ferritic ODS steels. 4) Complex oxide composed of Y and Ti elements was the majority of all oxide dispersoids. Within neutron irradiation up to 21 dpa, no significant change of individual oxide dispersoid was recognized, but its recoil resolution was implied from the statistical assessment.

Journal Articles

Mechanical behavior of oxide dispersion strengthened steels irradiated in JOYO

Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Ukai, Shigeharu; Kimura, Akihiko*

Materials Transactions, 46(3), p.493 - 497, 2005/05

 Times Cited Count:5 Percentile:46.01(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) steels, which are candidate materials for water-cooled solid breeder blankets, were fabricated with several manufacturing parameters, and then irradiated in JOYO to evaluate their irradiation behavior. Engineering stress strain curve of ODS steels irradiated at 673 K exhibited superior material response, i.e., increased tensile strength due to irradiation hardening and no loss of total elongation. Also, their temperature dependence on tensile property indicated that degradation of the tensile property at elevated temperature, which is closely related to phase stability during irradiation, could be avoided due to optimal combination of manufacturing parameters, such as chemical composition, type of inert gas during mechanical alloying, heat-treatment temperature and initial phase of the matrix.

Journal Articles

Nano-meso Structures and Ring-tensile Properties of Neutron-irradiated ODS Steels

Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Ukai, Shigeharu; Onuki, Somei

Vol.475-479(2005)1467, 0 Pages, 2005/00

Three ODS steel claddings, one martensitic phase and two ferritic ones in initial structures, were neutron-irradiated in the experimental fast reactor, JOYO, and those nano-meso structural evolutions during irradiations and those ring-tensile behaviors at as close conditions as it should be in service were evaluated. As for the structural features, no significant structural changes in macro scale such as grain morphology and dislocation structure were recognizable, but those in nano scale such as carbide and oxide dispersoid emerged in all neutron-irradiated ODS claddings. With respect to the ring-tensile properties, on the other hand, a manifested variation of an irradiation temperature dependence among all the ODS claddings was revealed, reflecting the each microstructural evolution of ODS claddings during neutron irradiation.

JAEA Reports

Evaluation of Radiation Embrittlement by Charpy Impact Tests with Miniaturized Specimens

Kurishita, Hiroaki*; Yamamoto, Takuya*; Narui, Minoru*; Yoshitake, Tsunemitsu; Akasaka, Naoaki

JNC TY9400 2004-006, 48 Pages, 2004/04

JNC-TY9400-2004-006.pdf:1.95MB

Radiation embrittlement in high-strength ferritic/martensitic steels of 2WFK and 63WFS and oxide dispersion strengthened (ODS) martensitic steel of H-35 that were irradiated in the experimental fast reactor JOYO is evaluated by instrumented Charpy impact tests for miniaturized (1.5 x 1.5 x 20 mm) and half-sized Charpy V-notch (CVN) specimens. Effects of thermal aging and microstructural evolution during irradiation on radiation embrittlement are described. Next, in order to clarify the specimen size effects on the ductile-to-brittle transition temperature (DBTT) in Charpy impact testing, a method to evaluate the plastic constraint loss for differently sized CVN specimens that may be responsible for the size effects is proposed and applied to 2 WFK. In the method, the constraint factor, a, that is an index of the plastic constraint is defined as a = s*/sy*. Here, s* is the critical cleavage fracture stress which is a material constant and sy* is the uniaxial yield stress at the DBTT at the strain rate generated in the Charpy impact test. The procedures for evaluating each of s* and sy* are described and the result of s* and sy*, thus the value of a, is presented for various types of miniaturized and full-sized CVN specimens of 2 WFK. It is shown that there is the following relationship between a and the specimen size factor, (A*/b2). a=a0-k(A*/b2)0.4 Here, A* is the critical area for cleavage fracture and b is the ligament size. a0 and k are constants depending on a /W (a is the notch depth and W is the specimen width). a increases with increasing a /W.

74 (Records 1-20 displayed on this page)