Zobrazeno 1 - 10
of 14
pro vyhledávání: '"G. E. Russcher"'
Autor:
W. E. Roake, T. T. Claudson, R. L. Simons, W. N. McElroy, L. D. Blackburn, S. D. Harkness, R. Grappel, S. G. McDonald, J. L. Straalsund, G. L. Guthrie, E. E. Bloom, J. R. Weir, W. G. Wolfer, J. P. Foster, F. A. Garner, P. J. Ring, K. D. Challenger, H. J. Busboom, L. A. Neimark, J. D. B. Lambert, W. F. Murphy, C. W. Renfro, B. F. Rubin, T. J. Black, W. K. Appleby, J. D. Stephen, R. F. Hilbert, T. D. Gulden, C. L. Smith, D. P. Harmon, W. W. Hudritsch, M. Coquerelle, J. Gabolde, R. Lesser, P. Werner, W. J. Lackey, F. J. Homan, A. R. Olsen, M. Conte, M. Mouchnino, F. K. Schmitz, B. L. Harbourn, M. S. Beck, A. Biancheria, W. H. McCarthy, K. J. Perry, G. R. Hull, J. W. Bennett, R. J. Beaver, A. E. Richt, O. M. Stansfield, G. E. Russcher, A. L. Pitner, G. L. Copeland, R. G. Donnelly, W. R. Martin, J. A. Basmajian, D. E. Mahagin, H. C. F. Ripfel, D. E. Baker, W. E. Ray, R. L. Miller, S. L. Schrock, G. A. Whitlow, E. Berkey, G. G. Sweeney, W. M. Hickam, J. C. Tobin, R. A. Moen, D. J. Ayres, T. M. Cullen, C. R. Brinkman, G. E. Korth, R. R. Hobbins, J. M. Steichen, Lee A. James, A. J. Lovell, W. E. Pennell
Publikováno v:
Nuclear Technology. 16:3-11
Autor:
G. E. Russcher, A. L. Pitner
Publikováno v:
Nuclear Technology. 16:208-215
Thirty-five sets of thermal reactor data were analyzed mathematically to derive a best fit function to predict gas release from boron carbide as a function of temperature, irradiation exposure, and...
Autor:
R. K. Marsh, L. L. King, G. M. Hesson, W. N. Rausch, G. E. Russcher, N. J. Wildung, C. L. Mohr, W. D. Bennett
Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::166a20a4600b6dcd09abc4723932a39b
https://doi.org/10.2172/1080074
https://doi.org/10.2172/1080074
Autor:
L. W. Cannon, G. E. Russcher, P. N. McDuffie, C. L. Mohr, R. L. Goodman, R. K. Marshall, C. Nealley, J. P. Pilger, L. L. King, G. M. Hesson
Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::205885ed9e3c41b17b6536d4981edee8
https://doi.org/10.2172/1079979
https://doi.org/10.2172/1079979
A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program by Pacific Northwest Laboratory (PNL). This experiment (MT-4) wa
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::463fada74d70b77d2d4c345464566f42
https://doi.org/10.2172/5870315
https://doi.org/10.2172/5870315
A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptu
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::b00d1f4070abad83af498d356fe4edaf
https://doi.org/10.2172/5402365
https://doi.org/10.2172/5402365
Autor:
R. K. Marshall, W. N. Rausch, G. M. Hesson, G W McNair, W. D. Bennett, G W Neally, C. Nealley, R. L. Goodman, L L Kirg, L. J. Parchen, G. E. Russcher, S W Heaberlin, C. L. Wilson, J. P. Pilger, R E Schreiber, W D Meitzler, N. J. Wildung
Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes u
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::1145bfe86e4b23a2ac9b51a8e6a527c7
https://doi.org/10.2172/1072978
https://doi.org/10.2172/1072978
A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the form
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::6d8238a8d6b5eff147ba96a3273d9bcc
https://doi.org/10.2172/5870336
https://doi.org/10.2172/5870336
Autor:
M. D. Wismer, G. M. Hesson, B. J. Webb, I. L. King, W. N. Rausch, G. E. Russcher, R. K. Marshall, C. L. Wilson, J. O. Barner, J. P. Pilger, N. J. Wildung, L. J. Parchen, C. L. Mohr
A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in ei
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::e977831a6848181f4c56ef145e0e4292
https://doi.org/10.2172/1084042
https://doi.org/10.2172/1084042