https://hdl.handle.net/1721.1/88179 2026-04-04T08:37:09Z https://hdl.handle.net/1721.1/90350 A simulation model for dynamic system availability analysis Deoss, D. L. (Dister LeRoy); Siu, N. O. (Nathan O.) "May, 1989."; "This report is based on the master's thesis [M.S., Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1989] of the first author."--Page iii; Includes bibliographical references (pages 85-86) 1989-01-01T00:00:00Z https://hdl.handle.net/1721.1/90349 Data analysis : operating crew characteristics and interactions during steam generator tube rupture simulation Huang, Y. (Yuhao); Siu, N. O. (Nathan O.); Lanning, David D.; Carroll, John S., 1948- This report provides an analysis of the data collected during a one-month visit to a 2-unit, non-U.S. PWR. The data consist of results from interviews (largely with plant operators and former shift engineers) and from reviews of videotapes covering crew responses to steam generator tube rupture training exercises. The interviews were aimed at indicating perceptions of individual and group skills. The analysis shows that the interview results are fairly consistent, that the time required to perform key actions does not generally correlate very well with the team quality ratings obtained from the interviews, and that the team quality ratings obtained from interviews correlate reasonably with ratings of the team performances (during the exercises) developed using the 7-dimension scale described in PNL-7250. Statement of responsibility on title page reads: Y. Huang, N. Siu, D. Lanning ,and J. Carroll; "June 1990."; Includes bibliographical references (leaf 11) 1990-01-01T00:00:00Z https://hdl.handle.net/1721.1/90348 Dynamic event trees in accident sequence analysis Siu, N. O. (Nathan O.); Acosta, C. (Crispiniano); Rasmussen, Norman C. Includes bibliographical references (leaves 14-15) 1990-01-01T00:00:00Z https://hdl.handle.net/1721.1/90347 Physical dependencies in accident sequence analysis Siu, N. O. (Nathan O.); Acosta, C. (Crispiniano); Rasmussen, Norman C. Includes bibliographical references (leaves 25-26) 1989-01-01T00:00:00Z https://hdl.handle.net/1721.1/90346 Design of a high-level waste repository system for the United States Driscoll, Michael J.; Baeza, Julio L.; Boerigter, Stephen T. (Stephen Troy); Broadbent, Gregory E. (Gregory Eric); Cabello, Ernesto David; Duran, Von B. (Von Buford); Hollaway, William Robert; Karlberg, Russell P.; Siegel, Matthew Jay; Simonson, Scott A. This report presents a conceptual design for a High Level Waste disposal system for fuel discharged by U.S. commercial power reactors, using the Yucca Mountain repository site recently designated by federal legislation. It represents the results of approximately 2000 person-hours of work by students enrolled in the combined undergraduate and graduate design subjects 22.033/22.33 of the M.I.T. Nuclear Engineering Department during Spring Term 1988.; Principal features of the resulting conceptual design include - use of unit trains (including piggyback cars for truck cask transporters where required) for periodic (once every ten years at each reactor) removal of old (cooled > 10 yrs.) spent fuel from at-reactor storage facilities - buffer storage at the repository site using dual purpose transportation/storage casks of the CASTOR V/21 type - repackaging of the spent fuel from the dual purpose transportation/storage casks directly into special-alloy disposal canisters as intact fuel assemblies, without rod consolidation - emplacement into a repository of modular design having a maximum total capacity of 150,000 MT and an annual handling capability of 4000 MT/yr - use of excavation techniques that minimize disturbance, both mechanical and chemical,; to the geologic environment - Incoloy 825 waste canisters arrayed to provide 57 kW/acre thermal loading optimized to the projected inventories - include a unit rail mounted vehicle for both the transportation and emplacement of the canister from the surface facilities to the underground repository - cost-effectiveness of the Yucca Mountain Site Criteria was studied via: a computer model, "WADCOM-II - Waste Disposal Cost Model II"; and an independent cost evaluation by the members of the design team. The total system cost (in constant 1988 dollars) was 1.9 billion dollars by WADCOM-II, and 5.3 billion dollars from the independent cost evaluation, resulting in a levelized disposal cost of 0.2 mills/kW-hr by WADCOM-II and 0.55 mills/kW-hr by the independent cost evaluation. Statement of responsibility on title-page reads: Report prepared by the following students enrolled in the combined graduate/undergraduate design subjects: 22.033, Nuclear Systems Design 22.33,- Nuclear Engineering Design for Spring Term 1988: Julio L.Baeza, Stephen T. Boerigter, Gregory E. Broadbent, Ernesto D. Cabello, Von B. Duran, William R. Hollaway, co-project manager, Russell P. Karlberg , Matthew J. Siegel, co-project manager, Scott A. Simonson; Instructor in charge: Prof. M. J. Driscoll; At head of title on cover: "Program on Nuclear Power Plant Innovation."; Includes bibliographical references 1988-01-01T00:00:00Z https://hdl.handle.net/1721.1/90345 Assessing the reliability of components and complex subsystems Siu, N. O. (Nathan O.); Pagnoni, Tommaso "Progress report for: Rockwell International Corporation."; Includes bibliographical references; Progress report 1988-01-01T00:00:00Z https://hdl.handle.net/1721.1/90344 Ammonia distillation for deuterium separation Petersen, G. T. (Gerald Thornton); Benedict, Manson "May 16, 1960."; Series statement handwritten on cover; "NYO-2347."; Also issued as a Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1960; Includes bibliographical references; Report; July, 1958 - May, 1960 1960-01-01T00:00:00Z https://hdl.handle.net/1721.1/90343 Thermal hydraulics of core/concrete interaction of severe LWR accidents Kao, Lainsu; Kazimi, Mujid S. "June 1987."; "Originally presented as the first author's thesis (Ph. D.)--Massachusetts Institute of Technology, 1987."; Includes bibliographical references 1987-01-01T00:00:00Z https://hdl.handle.net/1721.1/90342 Derivation of the diagonal 0-weighting method Zerkle, Michael L. Includes bibliographical references (leaf 24) 1987-01-01T00:00:00Z https://hdl.handle.net/1721.1/90341 Implications of Chernobyl for Seabrook Beckjord, Eric S. 1986-01-01T00:00:00Z https://hdl.handle.net/1721.1/90340 The irradiation of Santowax OMP in the M.I.T. in-pile loop Morgan, Dean T.; Mason, Edward A. (Edward Archibald), 1924- "May 1962."; Issued both as a 2 volume set, or a 2 volumes in 1 set in which p. [169] is the title-page to MITNE-22; IDO 11, 104; 11, 105; Includes bibliographical references 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/90339 A study of the spatial distributions of fast neutrons in lattices of slightly enriched uranium rods moderated by heavy water Woodruff, Gene L.; Kaplan Irving 1912-; Thompson Theos Jardin 1918-1970 "MIT-2344-5."; Includes bibliographical references (pages 201-207) 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/90338 Fuel cycle analysis in a thorium fueled reactor using bidirectional fuel movement : correction to report MIT-2073-1, MITNE-51 Stephen, James D.; Hofmann, Ferdinand. Fuel cycle analysis in a thorium fueled reactor using bidirectional fuel movements This report corrects an error discovered in the code used in the study "Fuel Cycle Analysis in a Thorium Fueled Reactor Using Bidirectional Fuel Movement," MIT-2073-1, MITNE-51. The results of the correction show considerable improvement in the conversion ratio. Although more recent cross-section data make these corrected results somewhat optimistic, the indication is that breeding on the thorium cycle may be possible with the CANDU-type reactor design. "June 1965."; "MIT-2073-4."; "MIT-2073-1."; Includes bibliographical references 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/90337 The solvent extraction of nitrosylruthenium by trilaurylamine in nitrate systems : summary report for the period July 1, 1960 to March 31, 1962 Skavdahl, Richard E. (Richard Earl), 1934-; Mason, Edward A. (Edward Archibald), 1924- "June 1, 1962."; Also issued as an Sc. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1962; Summary report; July 1, 1960 to March 31, 1962 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/90336 Examination of offsite radiological emergency protective measures for nuclear reactor accidents involving core melt Aldrich, David C.; McGrath Peter E.; Rasmussen Norman C. Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, "Melt-through" and "Atmospheric," based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for "Atmospheric" accidents are also examined in terms of their influence on the occurrence of public health effects. For "Melt-through" accidents, few, if any, early public health effects are likely, and doses in excess of Protective Action Guides (PAGs) are "confined" to areas within 10 miles of the reactor.; Evacuation appears to provide the largest reduction in whole body dose for this category. However, sheltering, particularly when basements are readily available, may be an acceptable alternative. Both evacuation and iodine prophylaxis can substantially reduce the dose to the thyroid. For "Atmospheric" accidents, PAGs are likely to be exceeded at very large distances, and significant numbers of early public health effects are possible. However, most early fatalities occur within 10 miles of the reactor. Within 5 miles, evacuation appears to be more effective than sheltering in reducing the number of early health effects. Beyond 5 miles, this distinction is less, or not, apparent. Within 10 miles, early health effects are strongly influenced by the speed and efficiency with which protective measures are implemented. Outside of 10 miles, they are not.; The projected total number of thyroid nodules is not substantially reduced unless iodine prophylaxis is administered over very large areas (distances). The qualitative effects of weather conditions on the above conclusions are also briefly discussed. "Date published: June 1978. --Reissued: October 1979."; MITNE series handwritten on title-page; "SAND78-0454."; Originally issued as a Ph. D. thesis by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1978; Originally issued as an; Includes bibliographical references 1979-01-01T00:00:00Z https://hdl.handle.net/1721.1/89787 Heavy water lattice project annual report / editors: T.J. Thompson, I. Kaplan, [and] M.J. Driscoll ; contributors J.H. Barch ... [et al.] Thompson, Theos Jardin, 1918-1970; Kaplan, Irving, 1912-; Driscoll, Michael J.; Barch, J. (J. H.); Berube, N. (N. L. J.); Bliss, Henry Edison; Bowles, Kenneth Douglas; Chase, Emery John; Cheng, Hsiang-Shou; Clikeman, Franklyn Miles; Forbes, Ian Alexander; Frech, David Franklin; Gosnell, James Waterbury; Harper, Thomas Lawrence; Harrington, Joseph, III; Hauck, Frederick Hamilton; Johnson, Malvin Gordon, Jr; Kelley, B. (Barbara), Miss; Papay, Lawrence Thomas; Pilat, E. E., 1937-; Rasmussen, Norman C.; Ricketts, Robert Lee; Sefchovich-Itzcovich, Elias; Seth, Shivaji Shrilal; Supple, A. T. Massachusetts Institute of Technology; Woodruff, Gene L. An experimental and theoretical program on the physics of heavy water moderated, slightly enriched lattices is being conducted at the Massachusetts Institute of Technology. During the past year, work was completed on studies of fast neutron distributions, lattices with added neutron absorbers, miniature lattices, two-region lattices, pulsed neutron source methods, and single,-rod experiments. In the past year, measurements were also completed on six lattices: three spacings each for 0.75-inch- and 0.387-inch-diameter, 0.947% enriched, uranium metal fuel. Statement of responsibility on title-page reads: Editors: T.J. Thompson, I. Kaplan and M.J. Driscoll; contributors: J. H. Barch, N.L. Berube, H.E. Bliss, K.D. Bowles, E J. Chase, H.S. Cheng, F.M. Clikeman, M.J. Driscoll, I.A. Forbes, D.Frech, J.W. Gosnell, T.L. Harper, J.Harrington, III, F.H. Hauck, M.G. Johnson, I. Kaplan, B. Kelley, L.T. Papay, E.E. Pilat, L.N. Price, N.C. Rasmussen, R.L. Ricketts, E. Sefchovich, S.S. Seth, A.T. Supple, T.J. Thompson, and G.L. Woodruff; "September 30, 1966"; "MIT-2344-9."; Includes bibliographical references; Annual report; September 30, 1966 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89786 Heavy water lattice project annual report / editors: T.J. Thompson ... [et al.] ; contributors J. Barch ... [et al.] Thompson, Theos Jardin, 1918-1970; Kaplan, Irving, 1912-; Clikeman, Franklyn Miles; Driscoll, Michael J.; Barch, J. (J. H.); Berube, N. (N. L. J.); Bliss, Henry Edison; D'Ardenne, Walter Herbert; Goebel, David Maxwell; Gosnell, James Waterbury; Harrington, Joseph, III; Hellman, S. P. Massachusetts Institute of Technology; Guéron, Henri Max; Kelley, B. (Barbara), Miss; Lanning, David D.; Papay, Lawrence Thomas; Pilat, E. E., 1937-; Robertson, Cloin Gentry; Sefchovich-Itzcovich, Elias; Supple, A. T. Massachusetts Institute of Technology; Woodruff, Gene L. Statement of responsibility on title-page reads: "Editors: T. J. Thompson, I. Kaplan, F.M. Clikeman, M.J. Driscoll; Contributors: J. Barch, N. Berube, H. Bliss, F.M. Clikeman, W.H. D'Ardenne, M.J. Driscoll, D.M. Goebel, J.W. Gosnell, H.M. Guéron, J. Harrington, III, S.P. Hellman, I. Kaplan, B. Kelley, D.D. Lanning, L.T. Papay, E.E, Pilat, C. Robertson, E. Sefchovich A. Supple, T.J. Thompson, and G.L. Woodruff ."; "September 30, 1965"; "MIT-2344-4."; Includes bibliographical references; Annual report; September 30, 1965 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/89785 Heavy water lattice project annual report / editors: D.D. Lanning, I/ Kaplan [and] F.M. Clickeman ; contributors J. Barch ... [et al.] Lanning, David D.; Kaplan, Irving, 1912-; Clikeman, Franklyn Miles; Barch, J. (J. H.); Berube, N. (N. L. J.); D'Ardenne, Walter Herbert; Gosnell, James Waterbury; Harrington, Joseph, III; Kelley, B. (Barbara), Miss; Malaviya, Bimal Kumar; Papay, Lawrence Thomas; Pilat, E. E., 1937-; Sefchovich-Itzcovich, Elias; Supple, A. T. Massachusetts Institute of Technology; Thompson, Theos Jardin, 1918-1970; Woodruff, Gene L. Statement of responsibility on title-page reads: "Editors: D.D. Lanning, I. Kaplan, F.M. Clikeman; Contributors: J. Barch, N. Berube, F.M. Clikeman, W.H. D'Ardenne, J.W. Gosnell, J. Harrington III, I. Kaplan, B. Kelley, D.D. Lanning, B.K. Malaviya, L.T. Papay, E.E. Pilat, E. Sefchovich, A. Supple, T.J. Thompson, and G.L. Woodruff."; "September 30, 1964."; "MIT-2344-3."; Includes bibliographical references (pages 165-168); Annual report; September 30, 1964 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89784 Heavy water lattice project annual report / editors: Irving Kaplan, D.D. Lanning, T.J. Thompson ; contributors H. E. Bliss ... [et al.] Kaplan, Irving, 1912-; Lanning, David D.; Thompson, Theos Jardin, 1918-1970; Bliss, Henry Edison; Clikeman, Franklyn Miles; D'Ardenne, Walter Herbert; Gosnell, James Waterbury; Harrington, Joseph, III; Kim, Hichull; Malaviya, Bimal Kumar; Pilat, E. E., 1937-; Profio, A. Edward, 1931-; Sefchovich-Itzcovich, Elias; Simms, Richard; Woodruff, Gene L. Statement of responsibility on title-page reads: editors: Irving Kaplan, D.D.Lanning, T.J. Thompson; contributors: H. E. Bliss, F.M. Clikeman, W.H. D'Ardenne, J.W. Gosnell, J. Harrington, III, I.Kaplan, H. Kim, D.D. Lanning, B.E.Malaviya, E.E. Pilat, A.E. Profio, E. Sefchovich. R. Simms, T.J. Thompson, G.L. Woodruff; "September 30, 1963"; Includes bibliographical references; Annual report; September 30, 1963 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89783 Heavy water lattice project annual report / editors: Irving Kaplan, A.E. Profio [and] T.J. Thompson ; contributors P.S. Brown ... [et al.] Kaplan, Irving, 1912-; Profio, A. Edward, 1931-; Thompson, Theos Jardin, 1918-1970; Brown, Paul S. (Paul Sherman); D'Ardenne, Walter Herbert; Harrington, Joseph, III; Malaviya, Bimal Kumar; Palmedo, Philip F.; Peak, John Carl; Simms, Richard; Weitzberg, Abraham; Wolberg, John R. Statement of responsibility on title-page reads: editors: Irving Kaplan A.E. Profio, T.J. Thompson; contributors: P.S. Brown, W.H. D'Ardenne, J.Harrington, III, I.Kaplan, B.K. Malaviya, P.F. Palmedo, J.C. Peak. A.E. Profio. R. Simms. T.J. Thompson. A. Weitzberg. J,R. Wolberg; "September 30, 1962"; "NYO-10, 208."; Includes bibliographical references; Annual report; September 30, 1962 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89782 In-pile loop irradiation studies of organic coolant materials : progress report, October 1, 1965 - December 31, 1965 Morgan, Dean T.; Bley, W. N. (William Norman); Timmins, Thomas Howard; Steiner, J. W. Massachusetts Institute of Technology "Issued: April 1, 1966."; "AEC Research and Development Report"--Cover; "MIT-334-48 ,Reactor Technology, Standard TID 4500."; Includes bibliographical references (leaf 28); Progress report; October 1, 1965 - December 31, 1965 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89781 In-pile loop irradiation studies of organic coolant materials : progress report, July 1- September 30, 1965 Morgan, Dean T.; Bley, W. N. (William Norman); Timmins, Thomas Howard "Issued: November 1, 1965."; "AEC Research and Development Report"--Cover; "MIT-334-33 ,Reactor Technology, Standard TID 4500."; Includes bibliographical references (leaf 35); Progress report; July 1, 1965 - September 30, 1965 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/89780 In-pile loop irradiation studies of organic coolant materials : progress report, October 1, 1963 - December 31, 1964, Mason, Edward A. (Edward Archibald), 1924-; Bley, W. N. (William Norman); Kim, Je Chul; Timmins, Thomas Howard; Terrien, Jean-François Emile Marie; Swan, Arthur Henry Statement of responsibility on title page reads: Report prepared by: E. A. Mason, Project Supervisor; Contributors: W.N. Bley, J.C. Kim, T.H. Timmins, J.F. Terrien, A.H. Swan; "Issued: February 1, 1965."; "AEC Research and Development Report"--Cover; "MIT-334-12."; Includes bibliographical references (leaves 21-22); Progress report; October 1, 1963 - December 31, 1964 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/89779 Equilibrium extraction characteristics of alkyl amines and nuclear fuel metals in nitrate systems: progress report for the period January 1 - September 31 [sic], 1963 : progress report XII Lloyd, Philip J.; Mason, Edward A. (Edward Archibald), 1924- "December 2, 1963."; Includes bibliographical references (pages 29-30); Progress report no. XII; January 1, 1963 - September 30, 1963 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89778 Equilibrium extraction characteristics of alkyl amines and nuclear fuel metals in nitrate systems: progress report for the period October 1 - May 30, 1964 : progress report XIII Lloyd, Philip J.; Mason, Edward A. (Edward Archibald), 1924- "August 3, 1964."; Includes bibliographical references (page 32); Progress report no. XIII; October 1, 1963 - May 30, 1964 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89777 Equilibrium extraction characteristics of alkyl amines and nuclear fuel metals in nitrate systems: progress report for the period June 1 - August 31, 1964 : progress report XIV Lloyd, Philip J.; Mason, Edward A. (Edward Archibald), 1924- "October 24, 1964."; Includes bibliographical references (page 21); Progress report no. XIV; June 1, 1964 - August 31, 1964 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89776 Equilibrium extraction characteristics of alkyl amines and nuclear fuel metals in nitrate systems: progress report for the period April 1- December 31, 1962 : progress report XI Mason, Edward A. (Edward Archibald), 1924-; Watanabe, Takashi "April 1, 1963."; Includes bibliographical references (page 28); Progress report no. XI; April 1, 1962 - December 31, 1962 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89775 Equilibrium extraction characteristics of alkyl amines and nuclear fuel metals in nitrate systems: progress report for the period January 1- September 30, 1960 : progress report VII Mason, Edward A. (Edward Archibald), 1924-; Skavdahl, Richard E. (Richard Earl), 1934- "November 1, 1960."; Includes bibliographical references (pages 20-21); Progress report no. VII; January 1 - September 30, 1960 1960-01-01T00:00:00Z https://hdl.handle.net/1721.1/89774 LMFBR Blanket Physics Project progress report no. 7 Driscoll, Michael J.; Aldrich, David C.; Kadiroğlu, Osman Kemal; Keyvan, Shahla; Khan, Hafeez-ur-Rehman; Lanning, David D.; Morton, R. (Rachel); Pásztor, János; Salehi, Ali Akbar; Shin, Jae In; Supple, A. T. Massachusetts Institute of Technology; Wargo, D. J. Massachusetts Institute of Technology; Wu, Shin-Shyong Statement of responsibility on title-page reads: editor: M.J. Driscoll; contributors: D.C. Aldrich, M.J. Driscoll, O.K. Kadiroglu, S. Keyvan, H.U.R. Khan, D.D. Lanning, R. Morton, J. Pasztor, T.J. Reckart, A.A. Salehi, J.I. Shin, A.T. Supple, D.J. Wargo, and S.S. Wu; Includes bibliographical references; Progress report; September 30, 1976 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89773 LMFBR Blanket Physics Project progress report no. 5 Driscoll, Michael J.; Brown, Gilbert Jay; Chan, Joseph Kwok-Kwong; Choong, Philip Tsi-shien; Kadiroğlu, Osman Kemal; Kalra, Manjeet Singh; Lukic, Yovan D.; Leveckis, Algis Stephen; Masterson, Robert Edward; Miethe, V. A. (Virginia A.); Scheinert, Paul Albert; Supple, A. T. Massachusetts Institute of Technology; Shin, Jae In; Tagishi, Akinori; Yeung, Man-kit Includes bibliographical references; Progress report; June 30, 1974 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89772 LMFBR Blanket Physics Project progress report no. 4 Driscoll, Michael J.; Lanning, David D.; Kaplan, Irving, 1912-; Supple, A. T. Massachusetts Institute of Technology; Alvim, Antonio Carlos Marques; Brown, Gilbert Jay; Choong, Philip Tsi-shien; Ducat, Glenn Alexander; Forbes, Ian Alexander; Gregory, Michael Vladimir; Ho, Simon Yeung-Nam; Hove, Craig Moller; Kadiroğlu, Osman Kemal; Kennerley, Robert John; Lazewatsky, Joel L.; Lederman, Luis; Miethe, V. A. (Virginia A.); Scheinert, Paul Albert; Thompson, Alan Martin; Todreas, Neil E.; Tzanos, C. P.; Wood, Paul Joseph Statement of responsibility on title-page reads: editors: M.J. Driscoll, D.D. Lanning, I. Kaplan, A.T. Supple ; contributors: A. Alvim, G.J. Brown, J.K. Chan, T.P. Choong, M.J. Driscoll, G. A. Ducat, I.A. Forbes, M.V. Gregory, S.Y. Ho, C.M. Hove, O. K. Kadiroglu, R.J. Kennerley, D.D. Lanning, J.L. Lazewatsky, L. Lederman, A.S. Leveckis, V.A. Miethe, P. A. Scheinert, A.M. Thompson, N.E. Todreas, C.P. Tzanos, and P.J. Wood; Includes bibliographical references; Progress report; June 30, 1973 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89771 LMFBR Blanket Physics Project progress report no. 3 Driscoll, Michael J.; Lanning, David D.; Kaplan, Irving, 1912-; Brewer, Shelby Templeton; Brown, Gilbert Jay; De Laquil, Pascal; Ducat, Glenn Alexander; Forbes, Ian Alexander; Gregory, Michael Vladimir; Ho, Simon Yeung-Nam; Kalra, Manjeet Singh; Kang, Chʻang-sun; Kim, L. T. Massachusetts Institute of Technology; Lazewatsky, Joel L.; Mason, Edward A. (Edward Archibald), 1924-; Ortiz, Nestor Ruben; Rasmussen, Norman C.; Rickard, I. C. Massachusetts Institute of Technology; Roberson, K. D. Massachusetts Institute of Technology; Supple, A. T. Massachusetts Institute of Technology; Thompson, Alan Martin; Tzanos, C. P. Statement of responsibility on title-page reads: editors: M.J. Driscoll, D.D. Lanning, I. Kaplan; contributors: S. T. Brewer, G.J. Brown, P. Delaquil, M.J. Driscoll, G.A. Ducat, I.A. Forbes, M. V. Gregory, S.Y. Ho, M.S. Kalra, C.S. Kang, L.T. Kim, D.D. Lanning, J.L. Lazewatsky, T.C. Leung, E.A. Mason, N.R. Ortiz, N.C. Rasmussen, I.C. Rickard, K.D. Roberson, A.T. Supple, A.M. Thompson, and C.P. Tzanos; Includes bibliographical references; Progress report ; June 30, 1972 1972-01-01T00:00:00Z https://hdl.handle.net/1721.1/89770 LMFBR Blanket Physics Project progress report no. 2 Forbes, Ian Alexander; Driscoll, Michael J.; Rasmussen, Norman C.; Lanning, David D.; Kaplan, Irving, 1912-; Brewer, Shelby Templeton; Brown, Gilbert Jay; De Laquil, Pascal; Forsberg, Charles Winfield; Gyftopoulos, E. P.; Hendrick, Peter Ladd; Kang, Chʻang-sun; Klucar, James Leonard; Leung, Timothy Chung-tim; Mason, Edward A. (Edward Archibald), 1924-; Ortiz, Nestor Ruben; Passman, Neil Alan; Rickard, I. C. Massachusetts Institute of Technology; Rogers, V. C. (Vern Child), 1941-; Sullivan, G. E. Massachusetts Institute of Technology; Supple, A. T. Massachusetts Institute of Technology; Tzanos, C. P.; Westlake, William James Statement of responsibility on title-page reads: Editors: I.A. Forbes, M.J. Driscoll, N.C. Rasmussen, D.D. Lanning and I. Kaplan; Contributors: S.T. Brewer, G.J. Brown, P.DeLaquil, III, M.J. Driscoll, I.A. Forbes, C.W. Forsberg, E.P. Gyftopoulos, P.L. Hendrick, C.S. Kang, I. Kaplan, J.L. Klucar, D.D. Lanning, T.C. Leung, E.A. Mason, N.R. Ortiz, N.A. Passman, N.C. Rasmussen, I.C. Rickard, V.C. Rogers, G.E. Sullivan, A.T. Supple, and C. P. Tzanos; Includes bibliographical references; Progress report; June 30, 1971 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89769 LMFBR Blanket Physics Project progress report no. 1 Forbes, Ian Alexander; Driscoll, Michael J.; Lanning, David D.; Rasmussen, Norman C.; Ali, Syed Jameel, 1978-; Brewer, Shelby Templeton; Choi, Dug Kwang; Clikeman, Franklyn Miles; Corcoran, W. R. (William Richard); Forsberg, Charles Winfield; Ho, Sung Ling; Kang, Chʻang-sun; Kaplan, Irving, 1912-; Klucar, James Leonard; Leung, Timothy Chung-tim; McFarland, Emerson Lee; Mertens, Paul Gustaaf; Ortiz, Nestor Ruben; Pant, Aniket; Passman, Neil Alan; Sheaffer, Michael Kell; Shupe, Dwight Ardan; Sullivan, G. E. Massachusetts Institute of Technology; Supple, A. T. Massachusetts Institute of Technology; Synan, Joseph William; Tzanos, C. P.; Westlake, William James Statement of responsibility on title-page reads: Editors: I.A. Forbes, M.J. Driscoll, D.D. Lanning, I. Kaplan, N.C. Rasmussen; Contributors: S.A. Ali, S.T. Brewer, D.K. Choi, F.M. Clikeman, W.R. Corcoran, M.J. Driscoll, I.A. Forbes, C.W. Forsberg, S.L. Ho, C.S. Kang, I. Kaplan, J.L. Klucar, D.D. Lanning, T.C. Leung, E.L. McFarland P.G. Mertens, N.R. Ortiz, A. Pant, N.A. Passman, N.C. Rasmussen, M.K. Sheaffer, D.A. Shupe, G.E. Sullivan, A.T. Supple, J.W. Synan, C.P. Tzanos, W.J. Westlake; "MIT-4105-3."; Includes bibliographical references; Progress report; June 30, 1970 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89768 Activities in nuclear engineering at M.I.T. "List of graduate theses (September 1989 to June 1991)"--Pages [132]-[140]; Progress report; September 1991 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89767 Activities in nuclear engineering at M.I.T. "List of graduate theses (September 1987 to June 1989) "--Pages [129]-[134]; Progress report; August 1989 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89766 Activities in nuclear engineering at M.I.T. "List of theses (February 1986-June 1987)"--Pages [133]-[138]; Progress report; August 1987 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89765 Activities in nuclear engineering at M.I.T. "List of theses"--Pages 130-133; Progress report; December 1985 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89764 Activities in nuclear engineering at M.I.T. "List of theses"--Pages 134-143; Progress report; September 1984 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89763 Activities in nuclear engineering at M.I.T. "List of theses"--Pages 120-123; Progress report; September 1981 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89762 Activities in nuclear engineering at M.I.T. "List of theses"--Pages 120-123; Progress report; January 1980 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89761 Activities in nuclear engineering at M.I.T. "List of theses"--Pages 157-167; Progress report; October 1978 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89760 Activities in nuclear engineering at M.I.T. "List of theses"--Pages 143-148; Progress report; September 1976 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89759 Activities in nuclear engineering at M.I.T. "List of thesis"--Pages 126-129; Progress report; February 1975 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89758 Activities in nuclear engineering at M.I.T. "List of thesis 1972-1973"--Pages 109-112; Progress report; August 1973 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89757 Technical specifications for the MIT Research Reactor Cover title; "June 1967."; Includes bibliographical references (pages 116-118) 1967-01-01T00:00:00Z https://hdl.handle.net/1721.1/89756 Fuel cycle code, "FUELMOVE III" Sovka, Jerry Alois; Benedict, Manson Further modifications to the fuel cycle code FUELMOVE are described which were made in an attempt to obtain results for reflected reactors operated under batch, outin, and bidirectional fueling schemes. Numerical methods used to obtain solutions to the condensed two-group diffusion equation are presented. Results indicated that the method for obtaining solutions for the thermal flux distribution in reflected reactors using this condensed two-group formulation appears to be inadequate in certain cases in which the reactor is treated explicitly as a separate region. A recommendation is made for one additional evaluation of this technique with a further recommendation that subsequent studies of the fuel cycle behavior of reflected reactors be made using the full two-group diffusion formulation. "This work was done in part at the M.I.T. Computation Center, Cambridge, Massachusetts."; "MIT-2073-3."; Includes bibliographical references (leaf 34); Progress report; September 1964 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89755 Geometrical effects on axial & azimuthal variations of heat flux to coolant in asymmetrically heated channels Wang, Chunyun, 1968-; Lanza, Nina. Massachusetts Institute of Technology This report summarizes analyses of the effects of heat conduction in a copper block on the heat flux to a coolant flowing axially in the block. Heat is assumed to be added through one side of the block corresponding to conditions that may arise in fusion reactors or particle accelerator targets. It is found that three dimensional analysis of the heat transport will be required to accurately describe the heat flux at the wall of the coolant channel. The effects of axial and azimuthal heat conduction in the copper block depend on the block width to channel diameter ratio and the BIOT number of the channel. "July 1998." 1998-01-01T00:00:00Z https://hdl.handle.net/1721.1/89754 Models and evaluation of human-machine systems Bayout Alvarenga, Marco Antonio The field of human-machine systems and human-machine interfaces is very multidisciplinary. We have to navigate between the knowledge waves brought by several areas of the human learning: cognitive psychology, artificial intelligence, philosophy, linguistics, ergonomy, control systems engineering, neurophysiology, sociology, computer sciences, among others. At the present moment, all these disciplines seek to be close each other to generate synergy. It is necessary to homogenize the different nomenclatures and to make that each one can benefit from the results and advances found in the other. Accidents like TMI, Chernobyl, Challenger, Bhopal, and others demonstrated that the human beings shall deal with complex systems that are created by the technological evolution more carefully. The great American writer Allan Bloom died recently wrote in his book 'The Closing of the American Mind' (1987) about the universities curriculum that are commonly separated in tight departments. This was a necessity of the industrial revolution that put emphasis in practical courses in order to graduate specialists in many fields. However, due the great complexity of our technological world, we feel the necessity to integrate again those disciplines that one day were separated to make possible their fast development. This Report is a modest trial to do this integration in a holistic way, trying to capture the best tendencies in those areas of the human learning mentioned in the first lines above. I expect that it can be useful to those professionals who, like me, would desire to build better human-machine systems in order to avoid those accidents also mentioned above. "September 1993."; "Prepared for: International Atomic Energy Association [sic], Wagramerstrasse 5, P. 0. Box 100 A-1400 Vienna, Austria."; Part of appendix A and bibliography missing; Includes bibliographical references 1993-01-01T00:00:00Z https://hdl.handle.net/1721.1/89753 Natural convection in high heat flux tanks at the Hanford Waste Site / [by] Mark van der Helm and Mujid S. Kazimi Van der Helm, Mark Johan, 1972-; Kazimi, Mujid S. A study was carried out on the potential for natural convection and the effect of natural convection in a High Heat Flux Tank, Tank 241-C-106, at the Hanford Reservation. To determine the existence of natural convection, multiple computations based on analytical models were made knowing the tank geometry and contents' thermal characteristics. Each computation of the existence of natural convection was based on the determination of the onset of natural convection generalizing the tank as a 1-D porous medium. Computations were done for a range of permeabilities considering the porous medium alone, with a superposed fluid layer, and with a salt gradient. Considering only the porous medium, the higher permeability value, 3.2 *10-10 ft2, allowed convection, though the lower permeability, 2.6*10-14 ft2, did not. The presence of the superposed layer induced convection throughout the porous medium for the full range of permeabilities.; Considering the effect of the salt gradient and superposed layer together, the effect of the superposed layer is expected to induce convection despite the stabilizing salt gradient. Therefore, natural convection is expected to exist in Tank 241-C-106. Secondly, because temperature measurements indicated lower temperatures at a location near the center of the tank, a thermal model was used to compute the local effects of a convective annulus around a thermocouple tree at that location. A conduction model of the tank and surroundings was used to bound the local model. The local model allowing convection in the annulus set the size of the annulus based on the known temperature measurements of the thermocouple tree and the boundary conditions set by the conduction model. Previous published calculations on Tank 241-C-106, allowing for only conduction within the tank, reported a steam region at the bottom of the tank with an approximately 24 foot radius.; In the present analysis, using the computer code, TEMPEST, it is found that the cooling effect of the annulus creates a region with a 12 foot radius surrounding the thermocouple tree in which the temperature is suppressed below the saturation temperature due to the effects of the convective annulus. The annulus gap width for matching temperatures and the boundary conditions is on the order of 1 inch. "February 1996."; Series statement handwritten on title-page; Page 118 blank; Also issued as an M.S. thesis written by the first author, and supervised by the second author, MIT Dept. of Nuclear Engineering; Includes bibliographical references (pages 115-117) 1996-01-01T00:00:00Z https://hdl.handle.net/1721.1/89752 The role of nuclear energy in the global context of the 21st century Häfele, Wolf; Dave Rose Memorial Lecture (1994 : Massachusetts Institute of Technology) "April 20, 1994."; "Text of the David J. Rose Lecture in Nuclear Technology."; Includes bibliographical references (pages 23-25) 1994-01-01T00:00:00Z https://hdl.handle.net/1721.1/89751 Computational and visualization techniques for Monte Carlo based SPECT Dobrzeniecki, Andrew B. (Andrew Bruno); Yanch, Jacquelyn C. Nuclear medicine imaging systems produce clinical images that are inherently noisier and of lower resolution than images from such modalities as MRI or CT. One method for improving our understanding of the factors that contribute to SPECT image degradation is to perform complete photon-level simulations of the entire imaging environment. We have designed such a system for SPECT simulation and modelling (SimSPECT), and have been using the system in a number of experiments aimed at improving the collection and analysis of SPECT images in the clinical setting. Based on Monte Carlo techniques, Sim- SPECT realistically simulates the transport of photons through asymmetric, 3-D patient or phantom models, and allows photons to interact with a number of different types of collimators before being collected into synthetic SPECT images. We describe the design and use of SimSPECT, including the computational algorithms involved, and the data visualization and analysis methods employed. "25 January 1993."; Includes bibliographical references (pages 40-42) 1993-01-01T00:00:00Z https://hdl.handle.net/1721.1/89750 Investigations of ORNL pressurized thermal shock experiments Jerng, Dong Wook; Carter, Robert G. Yankee Atomic Electric Company "March 30, 1992."; Includes bibliographical references (page 50); Research report ; February 12, 1992 - March 25,1992 1992-01-01T00:00:00Z https://hdl.handle.net/1721.1/89749 Narrow channel turbulence modeling project : final report Meyer, John E.; Kwak, Sangman, Sc. D. Massachusetts Institute of Technology; Shubert, T. D. (Tyler D.) "March 1992."; Includes bibliographical references; Final report https://hdl.handle.net/1721.1/89748 Risk impact of maintenance program changes Credit, Kimberly A. (Kimberly April); Ouyang, Meng; Siu, N. O. (Nathan O.) This study quantifies the change in one measure of plant risk, the frequency of loss of long-term decay heat removal, due to changes in maintenance at the James A. Fitzpatrick (JAF) plant. Quantification is accomplished in two steps. First, the effects of maintenance are quantified in terms of changes in: a) the frequency of common cause failure of residual heat removal (RHR) pumps and b) the frequency with which operators fail to correctly restore the RHR system following maintenance. These parameters are selected as the result of an importance analysis for the plant. Second, the changes in these two parameters are propagated through a simple plant model to obtain the associated change in plant risk. Based on this study's assessment of the current maintenance program at JAF, it appears that the potential for significant risk reduction due to improved maintenance is not extremely large; an optimal program might lead to an 80% reduction. The optimal program would place a stronger emphasis on predictive maintenance, and would employ improved procedures for RHR pump maintenance. There is potential for significant risk increase (around a factor of 70) if the maintenance program is significantly degraded (e.g., if post-maintenance is deemphasized). This study shows how, at a simple level, maintenance program changes can be quantified without explicit modeling of the details of a plant's management and organizational structure. However, such modeling may be required: a) to more strongly justify the quantitative factors used in the analysis and b) to quantify the effect of other program changes not yet treated (e.g., the strengthening of program elements ensuring feedback of information to organization). In addition, failure data specific to the JAF plant are also needed to increase the confidence in the quantitative results of this study. "January 1992."; Includes bibliographical references (pages 129-132); Final report, "Operating and maintenance cost reduction using probabilistic risk assessment (PRA)" 1992-01-01T00:00:00Z https://hdl.handle.net/1721.1/89747 Modeling control room crews for accident sequence analysis Huang, Y. (Yuhao); Siu, N. O. (Nathan O.); Lanning, David D.; Carroll, John S., 1948-; Dang, Vinh Ngoc This report describes a systems-based operating crew model designed to simulate the behavior of an nuclear power plant control room crew during an accident scenario. This model can lead to an improved treatment of potential operator-induced multiple failures, since it deals directly with the causal factors underlying individual and group behavior. It is intended that the model, or more advanced developments of the model, will be used in the human reliability analysis portion of a probabilistic risk assessment study, where careful treatment of multiple, dependent failures is required. The model treats the members of the control room crew as separate, reasoning entities. These entities receive information from the plant and each other, process that information, perform actions that affect the plant, and provide information to the other crew members.; The information retrieval, processing, and output activities are affected by the characteristics of the individual operator (e.g., his technical ability) and his relationship (measured in terms of "confidence level") with his fellow operators. Group behavior is modeled as the implicit result of individual operator behavior and the interactions between operators. The model is applied towards the analysis of steam generator tube rupture (SGTR) accidents at a non-U.S. pressurized water reactor, using the SIMSCRIPT 11.5 programming language. Benchmark runs, comparing the model predictions with videotaped observations of the performances of three different crews during SGTR training exercises, are performed to tune a small number of model parameters. The tuned model is then applied in a blind test analysis of a fourth crew. In both the benchmarking and blind test runs, the model performs quite well in predicting the occurrence, ordering, and timing of key events.; The model is also employed in a number of sensitivity analyses that demonstrate the robustness of the model (it generates plausible results even when the model parameters are assigned values not representative of observed crews) and the model's usefulness in investigating key issues (e.g., the effect of stress buildup on crew performance). i Statement of responsibility on title page reads: Y. Huang, N. Siu, D. Lanning, J. Carroll, and V. Dang; "December 1991."; Includes bibliographical references (pages 153-155); Final report: "A systems model for dynamic human error during accident sequences" 1991-01-01T00:00:00Z https://hdl.handle.net/1721.1/89746 An experiment to simulate the heat transfer properties of a dry, horizontal spent nuclear fuel assembly Lovett, Phyllis Maria; Todreas, Neil E. Nuclear power reactors generate highly radioactive spent fuel assemblies. Initially, the spent fuel assemblies are stored for a period of several years in an on-site storage facility to allow the radioactivity levels of the assemblies to decay. As the radioactive fission product isotopes in the fuel decay, they generate significant amounts of thermal energy producing high temperatures in the spent fuel. The spent fuel from nuclear power plants will eventually have to be transferred to a federal geologic repository in a spent fuel transportation casks. The purpose of this research project is to characterize the relative importance of the heat transfer mechanisms of radiation, conduction, and convection in a dry horizontally-oriented nuclear spent fuel assembly, for eventual application in spent fuel transportation cask design.; To determine the relative importance of each heat transfer mode, an experiment was designed and operated to characterize the heat transfer in an 8x8 square heater rod array (similar to a Boiling Water Reactor fuel assembly) in a horizontal orientation. The experimental apparatus was operated with the following variable parameters and their ranges: Power to Heater Rods (Controlling Temperatures from 40'C to 250'C); Heater Transfer Medium (Air, Nitrogen, Argon, and Helium); Pressure of the Heat Transfer Medium (15 psig, 0 psig, 24 inches of mercury); Power to Boundary Condition Box (not controlled). The experiment was designed, fabricated, and operated under the Sandia National Laboratories-approved MIT Nuclear Engineering Department Quality Assurance Program developed in this work specifically for this project. The test data obtained from the experimental apparatus was analyzed with the lumped keff/hedge model developed by R.D.; Manteufel at MIT, in related work on this research project, and the Wooten-Epstein relationship developed at Battelle Memorial Institute. The test data was used to validate the lumped keff/hedge model. Good agreement was found between the lumped keff/hedge model and the test data in each Test Campaign with the exception of Below Atmospheric Pressure data. Both experimental and theoretical sources for the discrepancy are discussed. However, the full reason for the deviation is not know. "September 1991."; At head of title: Final report - experimental; Also issued as an M.S. thesis written by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1991; Includes bibliographical references (pages 113-115) 1991-01-01T00:00:00Z https://hdl.handle.net/1721.1/89745 Choice of discount rate for cost levelization Correia, F. Fundação Getúlio Vargas; Driscoll, Michael J. "September 20, 1991."--Foreword; Includes bibliographical references (page 32) 1991-01-01T00:00:00Z https://hdl.handle.net/1721.1/89744 FRANTIC 5 (a version of FRANTIC II) : a computer code for evaluating system aging effect Dimitrijevic, Vesna B.; Rasmussen, Norman C.; Vesely, W. E. The FRANTIC 5 code is a modification of the FRANTIC II code for time dependent unavailability analysis. FRANTIC 5 is specially adapted for modeling the aging effects on system and component performance. The FRANTIC 5 code uses the linear aging model, i.e., based on the assumption that component failure rates increase linearly in time. The constant failure rate and the aging acceleration rate for a component can be changed during the plant life, which allows the creation of different time scales for components as a function of the replacement or any significant maintenance or repair action on the component. FRANTIC 5 preserves most of the unique features of FRANTIC II, for example the modeling of periodic testing. The output from FRANTIC 5 consists of the system mean unavailabilities, tables of the system unavailabilities at designated time points and the system mean unavailabilities between consecutive tests. The code is applied for evaluation of aging effects of the Auxiliary Feedwater System of Arkansas Nuclear Unit 1. The usefulness of the method will depend upon the availability of the component aging data needed to develop the model parameters. "November 1986."; Includes bibliographical references 1986-01-01T00:00:00Z https://hdl.handle.net/1721.1/89743 M.I.T. LMFBR Blanket Physics Project : final summary report Driscoll, Michael J. This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at M.I. T. in the period 1969-1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the M.I. T. Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly -heterogeneous blanket configurations, is documented for the record. "August 1983."; "DOE/ET/37241-54."; Includes bibliographical references (pages 49-58); Final summary report ; 1969-1983 1983-01-01T00:00:00Z https://hdl.handle.net/1721.1/89742 Availability of the THERMIT thermal hydraulic reactor computer codes through M.I.T. Parsons, Donald Kent; Todreas, Neil E.; Kazimi, Mujid S.; Lanning, David D. Three of the THERMIT thermal hydraulic reactor computer codes are available through the Department of Nuclear Engineering at MIT. The three available codes are THERMIT-2, for LWR subchannel analysis, THIOD, for BWR analysis and NATOF-2D, for LMFBR sodium boiling analysis. Descriptive summaries and sample results are given for each code. In addition, a list of THERMIT references is given. Cover title; Statement of responsibility on title reads: D. Kent Parsons, Neil E. Todreas, Mujid S. Kazimi, David D. Lanning; "April 1981."; Includes bibliographical references (pages 19-21) 1981-01-01T00:00:00Z https://hdl.handle.net/1721.1/89741 Analyzing the safety impact of containment inerting at Vermont Yankee Heising, Carolyn D. (Carolyn DeLane), 1952-; Lepervanche-Valencia, José Gregorio; Pilat, E. E., 1937-; Slifer, Bruce C. Post-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control probability (air dilution: .917 - .989; nitrogen inerting: .987 - .998). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify "unidentified" leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations related disbenefits. Includes bibliographical references; Final Report; July 1980 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89740 An Introduction to the THERMIT thermal hydraulic reactor computer codes at M.I.T. Parsons, Donald Kent; Todreas, Neil E.; Kazimi, Mujid S.; Lanning, David D. The THERMIT thermal hydraulic reactor computer codes developed at MIT are described. The codes include THERMIT-2, THIOD, NATOF-2D, THERMIT-3, THERMIT-2D-PLENUM, THERFLIBE, THERLIT, THERMIT (sodium) and THERMIT-SIEX. Descriptive code summaries and sample code results from each THERMIT version are given. Finally, a complete THERMIT bibliography is presented. Cover title; Statement of responsibility on title reads: D. Kent Parsons, Neil E. Todreas, Mujid S. Kazimi, David D. Lanning; "April 1981."; Includes bibliographical references (pages 29-31) 1981-01-01T00:00:00Z https://hdl.handle.net/1721.1/89739 Modeling of fuel-to-steel heat transfer in core disruptive accidents Smith, Russell Charles; Rohsenow, Warren M.; Kazimi, Mujid S. A mathematical model for direct-contact boiling heat transfer between immiscible fluids was developed and tested experimentally. The model describes heat transfer from a hot fluid bath to an ensemble of droplets of a cooler fluid that boils as it passes through the hot fluid. The mathematical model is based on single bubble correlations for the heat transfer and a drift-flux model for the fluid dynamics. The model yields a volumetric heat transfer coefficient as a function of the initial diameter, velocity and volume fraction of the dispersed component. An experiment was constructed to boil cyclopentane droplets in water. The mathematical and experimental results agreed reasonably well. The results were applied to investigate the possibility of steel vaporization during a hypothetical core disruptive accident in a liquid metal fast breeder reactor. The model predicts that substantial steel vaporization may occur in core disruptive accidents, if the steel reaches its saturation temperature rapidly enough. The potential importance of steel vaporization is dependent on the accident scenario. "June 1980."; Also issued as a Ph. D. thesis by the first author, MIT Dept. of Nuclear Engineering, 1980; Includes bibliographical references (pages 110-111) 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89738 An evaluation of the fast-mixed spectrum reactor Loh, Wee Tee; Driscoll, Michael J.; Lanning, David D. An independent evaluation of the neutronic characteristics of a gas-cooled fast-mixed spectrum reactor (FMSR) core design has been performed. A benchmark core configuration for an early FMSR design was provided by Brookhaven National Laboratory, the originators of the concept. The results of the evaluation were compared with those of BNL. Points of comparison included system reactivity and breeding ratio, and region-wise power densities and isotopic compositions as a function of burnup. The results are in sufficiently good agreement to conclude that the neutronic feasibility of the FMSR concept has been independently validated. Significant differences, primarily in higher plutonium isotope concentrations, occur only in regions of low neutronic importance, and plausible reasons for the differences are advanced based on sensitivity studies and comparison of spectral indices. While both M.I.T. and BNL calculations tend to predict that the benchmark design is slightly subcritical, at the beginning of equilibrium cycle, the margin to k = 1.0 is close enough (Ak < 0.03) that the situation can be remedied. Establishment of a consensus fission product cross section set was identified as an objective of merit, since non-negligible differences were found in results computed using various extant sets (BNL, LIB-IV, Japanese). Non-fission heating by gamma and neutron interactions was evaluated for the reference core design using a coupled neutron/gamma cross section set and SN calculations. In the unfueled regions of the core, moderator elements in particular, the non-fission heating rate was found to be significant (averaging about 6 kw/liter), but posed no obvious problems. In fueled regions the common assumption of local deposition of all energy at the point of fission was verified to be a good approximation for most engineering purposes. "February 1980."; Also issued as an M.S. thesis written by the first author and supervised by the second and third authors, MIT Dept. of Nuclear Engineering, 1980; Includes bibliographical references (pages 145-147) 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89737 Interfacial effects in fast reactors Saidi, Mohammad Said; Driscoll, Michael J. The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed to measure U-238 capture rates near the blanket-reflector interface in the MIT Blanket Test Facility. Prior MIT experiments on a thorium-uranium interface in a blanket assembly were also reanalyzed. Extremely localized fertile capture rate increases of on the order of 50% were measured immediately at the interfaces relative to extrapolation of asymptotic interior traverses, and relative to state-of-the-art (LIB-IV, SPHINX, ANISN/2DB) calculations which employ infinite-medium self-shileding throughout a given zone. A method was developed to compute a spatially varying background scattering cross section per absorber nucleus, a , which takes into account both homogeneous and heterogeneous effec~s on the interface flux transient. This permitted use of the standard self-shielding factor method (Bondarenko f-factors) to generate modified cross sections for thin layers near the interfaces. Calculations of the MIT experiments using this approach yielded good agreement with the measured data. "May 1979."; Also written as a Ph. D. thesis by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1979; Includes bibliographical references (pages 191-193) 1979-01-01T00:00:00Z https://hdl.handle.net/1721.1/89736 Experimental investigation of the thermal-hydraulics of gas jet expansion In a two-dimensional liquid pool Rothrock, Ray Alan; Kazimi, M. S. Gas jet blowdown in a two-dimensional liquid pool has been experimentally investigated. Two sets of experiments were performed: a set of hydrodynamic experiments, where a non-condensible gas is injected into a subcooled liquid pool; and a set of thermal-hydraulic experiments, where a non-condensible heated gas is injected into a near saturated liquid pool. Liquid entrainment by the gas, bubble growth characteristics, and the potential for vaporization, were investigated for a variety of experimental pressures (3 to 10 bars) and two liquid types (water and R-113). Liquid entrainment increased with increasing pressure. The fraction of the jet volume which is liquid is relatively the same for all pressures and decreases with time of expansion. A Taylor instability mechanism for entrainment is found to under predict the entrained volume. In the initial stages of the expansion, higher entrainment is experienced for more dense fluids. For the same fluid, the entrainment rate was slightly higher for the heated experiments compared to the unheated experiments. Both lateral and vertical growth rates increased with pressure. Vaporization may have occurred for the 4 bar initial pressure, 12 °C superheat condition in freon R-113. "October 1978."; Also issued as an M.S. thesis by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1978; Includes bibliographical references (pages 115-117) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89735 PL-MODT and PL-MODMC : two codes for reliability and availability analysis of complex technical systems using the fault tree modularization technique Modarres, M. (Mohammad); Wolf, Lothar The methodology used in the PL-MOD code has been extended to include the time-dependent behavior of the fault tree components. Four classes of components are defined to model time-dependent fault tree leaves. Mathematical simplifications are applied to predict the time-dependent behavior of simple modules in the fault tree from its input components' failure data. The extended code, PL-MODT, handles time-dependent problems based on the mathematical models that have been established. An automatic tree reduction feature is also incorporated into this code. This reduction is based on the Vesely-Fussell importance measure that the code calculates. A CUT-OFF value is defined and incorporated into the code. Any module or component in the fault tree whose V-F importance is less than this value will automatically be eliminated from the tree. In order to benchmark the PL-MODT code, a number of systems are analyzed. The results are in good agreement with other codes, such as FRANTIC and KITT. The computation times are comparable and in most of the cases are even lower for the PL-MODT code compared to the others. In addition, a Monte-Carlo simulation code (PL-MODMC) is developed to propagate uncertainties in the failure rates of the components to the top event of a fault tree. An efficient sorting routine similar to the one used in the LIMITS code is employed in the PL-MODMC code. Upon modularization the code proceeds and propagates uncertainties in the failure rates through the tree. Large fault trees such as the LPRS fault tree as well as some smaller ones have been analyzed for simulation, and the results for the LPRS are in fair agreement with the WASH-1400 predictions for the number of simulations performed. The codes PL-MODT and PL-MODMC are written in PL/l language which offers the extensive use of the list processing tools. First experience indicates that these codes are very efficient and accurate, specifically for the analysis of very large and complex fault trees "November 1978."; Includes bibliographical references 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89734 Sensitivity study of the assembly averaged thermal-hydraulic models of the MEKIN computer code in power transients / by Thomas Rodack [and] Lothar Wolf Rodack, Thomas; Wolf, Lothar The thermal-hydraulic (T-H) models and solution schemes employed by the MEKIN computer code have been examined. The effects of T-H input parameters on- predicted fuel temperatures and coolant densities were determined in transient analyses. Consideration was limited primarily to a simulated PWR control rod ejection transient. Limitations to the use of MEKIN that arise because of simplifying assumptions in the T-H models are discussed. Computation time may be reduced without altering the results of a transient analysis if appropriate MEKIN options are selected. Guidelines are presented to facilitate the selection of these options. Suggestions for improvement of the code are also ma:de. Cover title; "August 1977."; Also issued as a Nucl. E. and M.S. thesis by the first author and supervised by the second author, MIT Depts. of Nuclear and Mechanical Engineering, 1977; Includes bibliographical references (pages 242-246) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89733 Description of the two-dimensional nodal codes "NRMS," a multigroup static code, and "NRMT," a one- and two-group transient code for solution of the neutron diffusion equation in rectangular geometry Sims, Randall Nee "August 1977."; Series statement handwritten on spine; Includes bibliographical references 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89732 Description of the computer code 2DTD Shober, Robert Anthony "November 1976."; Includes bibliographical references (page 13) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89731 Nonlinear methods for solving the diffusion equation Shober, Robert Anthony; Henry, Allan F. This thesis is concerned with methods for the transient solution of the neutron diffusion equations in one or two energy groups. Initially, nonlinear methods for solving the static diffusion equations using the finite element method were investigated. By formulating a new eigenvalue equation, some improvement in the solution efficiency was obtained. However, the transient solution of the diffusion equation using the finite element method was considered to be overly expensive. An analytic method for solving the one-dimensional diffusion equation was then developed. Numerical examples confirmed that this method is exact in one dimension. The method was extended to two dimensions, and results compared employing two different approximations for the transverse leakage. The method based on a flat approximation to the leakage was found to be superior, and it was extended to time-dependent problems. Results of time-dependent test problems show the procedure to be accurate and efficient. Comparisons with conventional finite difference techniques (such as TWIGL or MEKIN) indicate that the scheme can be an order of magnitude more cost effective. "November, 1976."; Also issued as a Ph. D. thesis written by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1977; Includes bibliographical references (pages 112-116) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89730 Design of central irradiation facilities for the MITR-II research reactor Meagher, Paul Christopher; Lanning, David D. Design analysis studies have been made for various in-core irradiation facility designs which are presently used, or proposed for future use in the MITR-II. The information obtained includes reactivity effects, core flux and power distributions, and estimates of the safety limits and limiting conditions for operation. A finite-difference, diffusion theory computer code was employed in two and three dimensions, and with three and fifteen group energy schemes. The facilities investigated include the single-element molybdenum sample holder, a proposed double-element irradiation facility and a proposed central irradiation facility design encompassing most of the area of the three central core positions. In addition, a comparison of the effects of various absorber materials has been made for a core configuration which includes three solid dummies. Flux levels in the molybdenum holder facility and in the beam ports were calculated for both three and five dummy cores. Flooding the sample tube in these cases was found to increase the safety and operating limits, but not to unacceptable levels for an 8 inch blade height. For the five dummy case, the operating limit in the C-ring was predicted to reach its maximum allowed value at a blade bank height of 13.6 inches. The reactivity effect of flooding was calculated to be 0.19%AK for the five dummy case, in direct agreement with the measured value. Flooging the large sample channel in the double element facility was found to increase the reactivity by 1.5 6%AK ff and also to cause an unacceptable power-peaking. The proposed central irradiation facility is a thermal flux-trap which could produce thermal flux values of up to 2.0 x 1014 n/cm 2 sec. Computer estimates show that flooding this facility's central sample tube would increase this value to 2.5 x 1014 n/cm2 sec, without resulting in an unacceptable power peak. "September 1976."; Also issued as a Ph. D. thesis by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1977; Includes bibliographical references (pages 94-95) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89729 Reactor physics calculation of BWR fuel bundles containing gadolinia Morales, Diego; Lanning, David D.; Pilat, E. E., 1937- A technique for the calculation of the neutronic behavior of BWR fuel bundles has been developed and applied to a Vermont Yankee fuel bundle. The technique is based on a diffusion theory treatment of the bundle, with parameters for gadolinia bearing pins generated by transport theory, and converted to effective diffusion- theory values by means of blackness theory. The method has been used to examine the dependence of various bundle average parameters on control rod insertion history. "January 1977."; "YAEC-1126."; Includes bibliographical references (pages 142-144) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89728 CHD : a two-dimensional, multigroup, rectangular geometry, cubic hermite, finite element diffusion code Kautz, Frederick A.; Deppe, Lothario Olavo "June 1975."; Includes bibliographical references (pages [157A]-[157B]) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89727 The replacement of reflectors by Albedo-type boundary conditions Kalambokas, Panagiotis Constantinos; Henry, Allan F. The idea of the representation of reflectors by boundary conditions in static and dynamic reactor calculations is investigated. Analytical, group-diffusion theory boundary conditions relating neutron flux to normal current at core-reflector interfaces in terms of the reflector parameters are derived and applied for the implicit treatment of reflectors. Kirchhoff's formula of optics, translated into the neutron-diffusion language, is the general relation from which boundary conditions for various reflector configurations (slab, wedge, elbow), geometrically separable or not, are obtained. The integral form of these boundary conditions in space and/or time generally being unswited for computer implementation, a reduction to simle, approximate algebraic forms is carried out. The finite diffusion length and/or the finite neutron lifetime in various reflector materials are exploited in the above reduction.; The bulk of the applications of the reflector-replacement method are for two-dimensional, light water-reflected reactor models with typical, step-like core-reflector interfaces, in steady state. With these models it is found that, if the reflector is viewed as consisting of many, narrow channels perpendicular to the interface and the transchannel leakage is neglected, (so that the one-dimensional- slab boundary conditions can be used to represent the reflector), the accuracy of the method, is extremely good for large, shrouded cores, decreases with core-size and is poor when the surface of a small, unshrouded core contains reentrant corners.; In the latter case the accuracy is improved by either the explicit treatment of a light water buffer zone, about 2 cm thick, adjacent to the core and the representation of the rest of the reflector by one-dimensional-slab boundary conditions, or the employment of numerical corrections generated internally during the computation or the employment of simple, analytical corrections derived from the infinite-wedge-shaped reflector configuration. The replacement of light water reflectors in transients and the replacement of heavy water or graphite reflectors in steady state or transients are studied in a preliminary way. Excellent results are obtained from the implicit treatment of light water reflectors of one-dimensional-slab geometry in fast transients. Heavy water or graphite reflectors are found to be more difficult to replace by boundary conditions because of their longer diffusion lengths and neutron lifetimes.; In all cases, the computational cost is reduced by about 40% when the reflector is treated implicitly. "November, 1975."; Also issued as an Sc. D. thesis by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1976; Includes bibliographical references (pages 221-224) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89726 Neutronic analysis of a proposed plutonium recycle assembly Solan, George Michael; Lanning, David D.; Momsen, Bruce William Foster; Pilat, E. E., 1937- A method for the neutronic analysis of plutonium recycle assemblies has been developed with emphasis on relative power distribution prediction in the boundary area of vastly different spectral regions. Such regions are those of mixed oxide (Pu0 2 in natural U02 ) fuel pins relative to enriched uranium pins, or water regions relative to fuel pin regions. The basic analytical methods for determination of spectrum averaged constants are given in the following descriptions: (1) Generalized Mixed Number Density (GMND) group constants (based on Breen's Mixed Number Density Method) are generated by a modified version of the spectrum code LASER, called LASER-M. (2) THERMOS Corrected LASER-M (TCL) group constants are based on mixed oxide- uranium oxide and water region boundary modeling in one dimensional (slab) geometry with the integral transport code THERMOS.; The LASER-M model, as modified by addition of ENDF/B-II thermal cross sections for the plutonium isotopes, is used to predict the criticality of experimental lattices of U02 - 2 w/o Pu0 2, and fair agreement is shown. LASER-M unit cell depletion calculations with Yankee Core I data (3.4 w/o U-235) to 40,000 MWD/MT and Saxton Core II data (6.6 w/o Pu02 in natural U02) to 20,000 MWD/MT show good isotopic agreement. Saxton Critical Reactor Experiment (CRX) lattice cores (19 x 19 rod array) consisting of a single fuel type region (mixed oxide or uranium oxide) or multiregions of both pin types were analyzed for relative power distribution comparisons. Cores with water slot regions were included. LASER-M Normal, LASER-M GMND and TCL two group constants were used with PDQ-7 in the calculations. GMND results were in excellent agreement compared to the good agreement of TCL for these cases of isolated spectral disturbances in an asymptotic core region.; The methods were applied to a proposed plutonium recycle "island design" assembly in which a large control rod water region is in close proximity to a zoned mixed oxide region. The TCL method yielded significantly greater power peaking and mixed oxide region average power owing to the spectral influence of the water region explicitly accounted for in this method. Such a result is consistent with published calculations. It is concluded that infinite lattice spectrum calculations are insufficient to deal with spectrum effects more complex than those in the Saxton CRX experiments. Statement of responsibility on title-page reads: George M. Solan David D. Lanning Bruce F. Momsen, and Edward E. Pilat; "August 1975."; Also issued as a Nucl. E. thesis, MIT Dept. of Nuclear Engineering, 1975; Includes bibliographical references (pages 271-275) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89725 Gas-cooled fast breeder reactor fuel element thermal-hydraulic investigations : final report Eaton, Thomas Eldon; Lanning, David D.; Todreas, Neil E. Experimental and analytical work was performed to determine the influence of rod surface roughening on the thermal-hydraulic behavior of rod array type, nuclear fuel elements. Experimental data was obtained using a grid-spaced, 37-rod hexagonal test section with both a smooth and a rough rod array. The rods were 0.331 inch (8.41 mm) diameter with a pitch/diameter of 1.30. The roughened surface used trapezoidal ribs 6-mils (0.15 mm) high with a rib pitch/height of 12. Velocity profiles taken at the flow exit plane indicated that when comparing the rough array results with the smooth, the gap velocities were lower, the peak-to-average velocities were higher: and the peripheral subchannel velocities were higher. Axial static pressure profiles. were used to determine rod array friction factors and grid loss coefficients. The friction factor results were in agreement with predictions using tube data.; The friction factor multipliers were strongly Reynolds number dependent and grid losses were apparently 10% higher in the rough rod array. Detailed pressure profiles were taken in the axial vicinity of the grid spacers. Coolant mixing data using a salt solution tracer was obtained for smooth and rough arrays. Flow scattering at the spacers was responsible for most of the smooth array tracer dispersion. In the rough array, turbulent interchange was considerably higher. The grid-spaced, rough array, dimensionless mixing coefficient was estimated to be 0.C20 + 0.005. Flow scattering at the grids prevented the determination of geometry and Reynolds number effects, as well as, the smooth array mixing coefficient. By neglecting coolant mixing and radial pressure gradients, an equation was developed to determine the flow rate in the subchannels of a nuclear fuel element with roughened surfaces and gas-cooling.; Relative subchannel flow rates were influenced by flow regime, fuel element geometry, fuel rod surface roughening, Reynolds number and coolant property variations. Two simple models were discussed which estimate the "equivalent friction factor" in partially roughened flow passages. Computational results obtained using the RUFHYD code showed that fuel element thermal-hydraulics are influenced by both rod array design parameters and operating conditions. Calculational results included axial subchannel flow distributions, optimum subchannel design estimates, and peripheral subchannel flow sensitivities to changes in rod-to-wall gap. "August 1975."; "Prepared for the General Atomic Company P.O. Box 81608 San Diego, California 92138."; Also issued as an Sc. D. thesis by the first author and supervised by the second and third author, MIT Dept. of Nuclear Engineering, 1975. -- Technical report has the following sub-title: Final report; Includes bibliographical references (pages 295-301); Final report; August 1975 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89724 The effect of reactor size on the breeding economics of LMFBR blankets Tagishi, Akinori; Driscoll, Michael J. The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MWe were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that at a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's. "February, 1975."; Also issued as an Sc. D. thesis written by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1975; Includes bibliographical references (pages 344-346) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89723 An analysis of plutonium recycle fuel elements in San Onofre-1 Momsen, Bruce William Foster; Lanning, David D.; Pilat, E. E., 1937- A method has been developed to allow independent assessment of the use of plutonium recycle assemblies in operating reactors. This method utilizes Generalized Mixed Number Density (GMND) cross sections (based on Breen's Mixed Number Density cross sections) and the spectrum code LASER. LASER is modified to form LASER-M by adding ENDF/B-II thermal cross sections for the plutonium isotopes; adding edits to output G-aND cross sections, approximate microscopic removal and transport cross sections; and increasing LASERs compatibility with commonly used diffusion theory codes such as PDQ. Plutonium critical experiments for a number of lattices of 1.5 w/o and 6.6 w/o plutonium are analysed with LASER-M which is found to give better criticality agreement than LASER (without the ENDF/B-II plutonium cross sections) and other published data. Unit assembly power distributions are calculated for a uranium assembly and a constant and graded enrichment plutonium assembly both surrounded by uranium assemblies. The use of LASER-M with GMND cross sections is found to give excellent agreement with the published calculations of power distributions for the uranium assembly and good agreement for the plutonium assemblies. A quarter core depletion calculation of the San Onofre reactor containing four plutonium recycle demonstration assemblies is performed using the diffusion theory computer code PDQ-7. Use of PDQ-7 with GMND cross sections from LASER-M is shown to give excellent agreement with quasi experimental power distributions at cycle burnups of 0 MWD/MTM, 3342 MWD/MTM, and 6045 MWD/'MTM. Also, the calculated value of k-eff versus cycle burnup is determined to be in excellent agreement with the actual operating condition of k-eff = 1 .000. "May 1974."; Also issued as an Nucl. E. thesis by the first author, MIT, Dept. of Nuclear Engineering, 1974; Includes bibliographical references (pages 167-171) 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89722 An experimental investigation of acoustic cavitation as a fragmentation mechanism of molten tin droplets in water Bjørnard, Trond Arnold A series of experiments were performed where single molten tin droplets of known size, shape and temperature were dropped from a low height into a pool of distilled water. The pressure waves emanating from the hot droplets were recorded by a transducer in the coolant for varying initial droplet and pool temperatures. The results obtained show well defined patterns of pressure frequency and magnitude behavior. Application of the results to the acoustic cavitation theory of fragmentation shows that the pressure excursions within the molten tin are considerably less severe than was predicted. The likelihood that acoustic cavitation caused the observed fragmentation is therefore considerably diminished. Further, the results strongly suggest that spontaneous nucleation of the coolant did not cause the observed fragmentation, while it appears that another mechanism linked to the observed dwell time behavior was responsible. Further theoretical and experimental work is required to establish the nature of this mechanism. "May 30, 1974."; Also issued as an M.S. thesis, MIT, Dept. of Nuclear Engineering, 1976, (c1975); Includes bibliographical references (pages 110-112) 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89721 Utility system integration and optimization models for nuclear power management Deaton, Paul Ferris; Mason, Edward A. (Edward Archibald), 1924- A nuclear power management model suitable for nuclear utility systems optimization has been developed for use in multi-reactor fuel management planning over periods of up to ten years. The overall utility planning model consists of four sub-models: (1) Refueling and Maintenance Model (RAMM), (2) System Integration Model (SIM), (3) System Optimization Model (SOM), and (4) CORE Simulation and Optimization Models (CORSOM's). The SIM and SOM sub-models were developed in this study and are discussed in detail; full-scale computerized versions of each (SYSINT and SYSOPT, respectively) are evaluated as part of the methods development research. The RAMM generates feasible, mutually exclusive nuclear refueling-fossil maintenance schedules. These are evaluated in detail by the rest of the model. Using the Booth-Baleriaux probabilistic utility system model, the SIM integrates the characteristics of the utility's plants into a representation which meets the necessary operating constraints.; Scheduling of system nuclear production and detailed fossil production is done for each time period (few weeks) making up the multi-year planning horizon. Utilizing a network programming model, the SOM optimizes the detailed production schedules of the nuclear units so as to produce the required system nuclear energy at minimum system cost. CORSOM's are utilized to optimize reload parameters (batch size and enrichment) and to generate the individual reactor fuel costs and nuclear incremental costs. These incremental costs are then used by the SOM's iterative gradient optimization technique known as the method of convex combinations. The SYSINT model is shown to be remarkably fast, performing the Booth-Baleriaux simulation for a single time period on a system with over 45 generating units in less than 2.5 seconds on an IBM-370 model 155 computer. SYSOPT converged to optimum solutions in roughly ten iterations.; Immediate reduction of iterations by roughly half is estimated by merely increasing piecewise-linearization of the network objective function. Overall model computational requirements are limited by available CORSOM's, which require 99% of the computational effort (over 3 minutes per reactor per SOM iteration). Nuclear incremental costs (~ 0.8-1.6 $/MWH) are shown to be less than fossil incremental costs (> 2.0 $/MWH) for the foreseeable future. Thus, nuclear power should always be operated so as to supply customer demands with a minimum use of the more expensive fossil energy. For the same reason, the lengthening of nuclear irradiation cycles (in terms of both energy and time) more than pays for itself by reducing the total cost of fossil replacement energy. Idealized nuclear production schedules yield constant nuclear incremental costs regardless of reactor unit and time. One of the key input parameters is the fossil thermal energy cost. "Issued: June 1973."; Vita; Also issued as a Ph. D. thesis by the first author and supervised by the second author, MIT, Dept. of Nuclear Engineering, 1973; Includes bibliographical references (pages 578-582) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89720 Incremental costs and optimization of in-core fuel management of nuclear power plants Watt, Hing Yan; Benedict, Manson; Mason, Edward A. (Edward Archibald), 1924- This thesis is concerned with development of methods for optimizing the energy production and refuelling decision for nuclear power plants in an electric utility system containing both nuclear and fossil-fuelled stations. The objective is to minimize the revenue requirements for refuelling the power plants during the planning horizon; the decision variables are the energy generation, reload enrichment and batch fraction for each reactor cycle; the constraints are that the customer's load demand, as well as various other operational and engineering requirements be satisfied. This problem can be decomposed into two sub-problems. The first sub-problem is concerned with scheduling energy between nuclear reactors which have been fuelled in an optimal fashion. The second sub-problem is concerned with optimizing the fuelling of nuclear reactors given an optimized energy schedule.; These two sub-problems when solved iteratively and interactively, would yield an optimal solution to the original problem. The problem of optimal energy scheduling between nuclear reactors can be formulated as a linear program. The incremental cost of energy is required as input to the linear program. Three methods of calculating incremental cost are considered: the Rigorous Method, based on the definition of partial derivativesis accurate but time consuring; the Inventory Value Method and the Linearization Method, based respectively on equations of inventory evaluation and linearization, are less accurate, but efficient. The latter two methods are recommended for the early stages of optimization. The problem of optimizing the fuelling of nuclear reactors has been solved for two cases: the special case of steady state operation, and the general case of nonsteady- state operation. The steady-state case has been solved by simple graphic techniques.; The results indicate that reactors should be refuelled with as small a batch fraction as allowed by burnup constraints. The non-steady case has been solved by polynomial approximation, in which the objective function as well as the constraints are approximated by a sum of polynomials. The results indicate that the final selection of an optimal solution from a set of sub-optimal solutions is primarily based on engineering considerations, and not on economics considerations. "Issued: February 1973."; Also issued as an Sc. D. thesis by the first author and supervised by the second and third authors, MIT, Dept. of Nuclear Engineering, 1973; Includes bibliographical references (pages 250-251) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89719 Heterogeneous effects in fast breeder reactors Gregory, Michael Vladimir; Driscoll, Michael J.; Lanning, David D. Heterogeneous effects in fast breeder reactors are examined through development of simple but accurate models for the calculation of a posteriori corrections to a volume-averaged homogeneous representation. Three distinct heterogeneous effects are considered: spatial coarse-group flux distribution within the unit cell, anisotropic diffusion, and resonance self-shielding. An escape/transmission probability theory is developed which yields region-averaged fluxes, used to flux-weight cross sections. Fluxes calculated by the model are in substantial agreement with S 8 discrete ordinate calculations. The method of Benoist, as applied to tight lattices, is adopted to generate anisotropic diffusion coefficients in pin geometry. The resulting perturbations from a volume-averaged homogeneous representation are interpreted in terms of reactivities calculated from first order perturbation theory and an equivalent "total differential of k" method.; Resonance self-shielding is treated by the f-factor approach, based on correlations developed to give the self-shielding factors as functions of one-group constants. Various reference designs are analyzed for heterogeneous effects. For a demonstration LMFBR design, the whole-core sodium void reactivity change is calculated to be 90e less positive for the heterogeneous core representation compared to a homogeneous core, due primarily to the effects of anisotropic diffusion. Parametric studies show at least a doubling of this negative reactivity contribution is attainable for judicious choices of enrichment, lattice pitch and lattice geometry (particularly the open hexagonal lattice). The fuel dispersal accident is analyzed and a positive reactivity contribution due to heterogeneity is identified. The results of intra-rod U-238 activation measurements in the Blanket Test Facility are analyzed and compared to calculations.; This activation depression is found to be of the order of 10% (surfaceto- average) for a typical LMFBR blanket rod and is ascribed to the effect of heterogeneous resonance self-shielding of U-238. Heterogeneous effects on the breeding ratio are studied with the conclusions that accounting for resonance self-shielding reduces the total breeding ratio by over 10%, but heterogeneous effects are not important for breeding ratio calculations. "January, 1973."; Also issued as a Ph. D. thesis written by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1973; Includes bibliographical references (pages 259-266) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89718 Variational derivation of modal-nodal finite difference equations in spatial reactor physics Bailey, Patrick Gage; Henry, Allan F. A class of consistent coarse mesh modal-nodal approximation methods is presented for the solution of the spatial neutron flux in multigroup diffusion theory. The methods are consistent in that they are systematically derived as an extension of the finite element method by utilizing general modal-nodal variational techniques. Detailed subassembly solutions, found by imposing zero current boundary conditions over the surface of each subassembly, are modified by piecewise continuous Hermite polynomials of the finite element method and used directly in trial function forms. Methods using both linear and cubic Hermite basis functions are presented and discussed. The proposed methods differ substantially from the finite element methods in which homogeneous nuclear constants, homogenized by flux weighting with detailed subassembly solutions, are used. However, both schemes become equivalent when the subassemblies themselves are homogeneous. One-dimensional, two-group numerical calculations using representative PWR nuclear material constants and 18-cm subassemblies were performed using entire subassemblies as coarse mesh regions. The results indicate that the proposed methods can yield comparable if not superior criticality measurements, comparable regional power levels, and extremely accurate subassembly fine flux structure with little increase of computational effort in comparison with existing coarse mesh methods. "July 1972."; Vita; Also written by the first author as a Ph. D. thesis, MIT, Dept. of Nuclear Engineering, 1972; Includes bibliographical references (pages 134-137) 1972-01-01T00:00:00Z https://hdl.handle.net/1721.1/89717 The Reactivity and transient analysis of MITR-II Yarman, Nuh Tolga; Henry, Allan F.; Lanning, David D.; Gosnell, James Waterbury The two-dimensional, time dependent, three-group diffusion equations for the proposed designed core of the MIT reactor are written with an extra source term accounting for the photoneutrons generated in the D20 reflector. An analytical expression is developed for this term. Then an approximate flux composed of two spatial shapes chosen beforehand, each having an unknown time coefficient, is inserted into the time dependent multigroup equations and the weighted residual criteria is applied. This yields multimode kinetics equations with generalized definitions for the conventional matrix parameters: generation time, reactivity, delayed neutron (and photoneutron) fractions matrices. Computational methods for these parameters are presented. An accident concerning the withdrawal of the shim rods is examined with the code OZAN written for the purpose of-the computations required by the present work. This study suggests that a space-dependent analysis is required to analyse the accident postulated. Statement of responsibility on title page reads: Nuh Tolga Yarman A.F. Henry, D.D. Lanning, and J.W. Gosnell; "July 1972."; Originally written by the first author as a Ph. D. thesis, MIT, Dept. of Nuclear Engineering, 1972; Includes bibliographical references (pages [191]-193) 1972-01-01T00:00:00Z https://hdl.handle.net/1721.1/89716 Development of utility system simulation model : progress report no. 2 Mason, Edward A. (Edward Archibald), 1924-; Deaton, Paul F.; Kearney, Joseph P. (Joseph Patrick); Rieck, Terrance Arthur Statement of responsibility on title-page reads: Edward A. Mason, Paul F. Deaton, Joseph P. Kearney and Terrance A. Rieck; "July 30, 1971."; "Worked preformed for Commonwealth Edison Company, Chicago, Illinois."; Includes bibliographical references (pages 138-139); Progress report no. 2; January 1, through June 30, 1971 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89715 The economics of fuel depletion in fast breeder reactor blankets Brewer, Shelby Templeton A fast breeder reactor fuel depletion-economics model was developed and applied to a number of 1000 MWe UMBR case studies, involving radial blanket-radial reflector design, radial blanket fuel management, and sensitivity of energy costs to changes in the economic environment. Choice of fuel cost accounting philosophy, e.g. whether or not to tax plutonium revenue, was found to have significant effect on absolute values of energy costs, without, however, distorting design rankings, comparative results, and irradiation time optimization. A single multigroup physics computation, to obtain the flux shape and local spectra for depletion calculations, was found to be sufficient for preliminary design and sensitivity studies. The major source of error in blanket depletion results was found to be the assumption of a fixed flux snape over an irradiation cycle; spectrum hardening in the radial blanket with irradiation is of minor importance. The simple depletion-economics model was applied to several 1000 NWe L1FBR case studies. Advantages of a moderating reflector were found to increase as blanket thickness was reduced. For a 45 cm radial blanket, a beryllium metal reflector offered little improvement, in blanket fuel economics, over sodium; for a 15 cm blanket, beryllium increased net blanket revenue by about 60%. An improvement of about 30% in net blanket revenue resulted when each radial blanket annular region was assumed to be expose" to its own local optimum irradiation time. Optimum radial blanket irradiation time and the net blanket revenue (mills/Khiie) at this optimum were found to be approximately linear in the unit fuel cycle costs, i.e. fabrication and reprocessing costs ($/4gld) and fissile market value ($/kg). "November, 1972."; Also issued as a Ph. D. thesis by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering 1973; Includes bibliographical references (pages 333-342) 1972-01-01T00:00:00Z https://hdl.handle.net/1721.1/89714 Aspects of a seismic study of the MITR Allen, G. C. The design version of the Massachusetts Institute of Technology research reactor (MITR-II) was analyzed subject to earthquake forces. The problem was divided into three major areas. First, the reactor core tank and support structure were studied. The reactor can be adequately cooled and shutdown if the core tank remains undamaged. Using a SABOR-5 computer program, the peak accelerations required to cause yielding of the core tank were calculated to be well above potential earthquake accelerations. Second, the possibilities of potential damage to miscellaneous reactor systems were studied, The miscellaneous systems were studied to see if earthquake accelerations, resonance response, or differential motions would result in damage leading to major radioactive releases. No major potential hazards were discovered. Third, the possibility of earthquake damage to the reactor stack was studied. An approximate analysis of the stack subject to dynamic earthquake shear and a 100 mile per hour wind was made. A case of a fallen stack was modeled to determine its effect on the containment building. The conservative calculations indicate that it is unlikely that the stack will fall and even if it were to fall onto the containment shell, it would not cause damage to the reactor core tank. Within the scope of this report, it appears that the design MITR-II is adequate to provide required protection even in the event of the maximum expected earthquake motions. "May, 1971."; Series statement handwritten on spine. -- Missing MITNE cover and title-page. -- Title-page is the thesis title page; Also issued as an M.S. thesis, MIT, Dept. of Nuclear Engineering, 1971; Includes bibliographical references (pages 117-118) 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89713 Evaluation of the parfait blanket concept for fast breeder reactors Ducat, Glenn Alexander; Driscoll, Michael J.; Todreas, Neil E. An evaluation of the neutronic, thermal-hydraulic, mechanical and economic characteristics of fast breeder reactor configurations containing an internal blanket has been performed. This design, called the parfait blanket concept, employs a layer of axial blanket fuel pellets at the core midplane in the fuel pins of the inner enrichment zone; otherwise, the design is the same as that of the conventional LMFBR's to which the parfait configuration was compared. Two significant advantages were identified for the parfait blanket concept relative to the conventional design. First, the parfait configuration has a 25% smaller peak fast flux which reduces wrapper tube dilation by 37% and fuel element elongation by 29%; and second, axial and radial flux flattening contribute to a 7. 6% reduction in the peak fuel burnup. Both characteristics significantly diminish the problems of fuel and metal swelling. Other advantages identified for a typical parfait design include: a 25% reduction in the burnup reactivity swing, which reduces control rod requirements; a 7% greater overpower operating margin; an increased breeding ratio, which offsets the disadvantage of a higher critical mass; and more favorable sodium voiding characteristics which counteract the disadvantage of an 8% smaller power Doppler coefficient. All other characteristics investigated were found to differ insignificantly or slightly favor the parfait design. "January 1974."; Also issued as a Ph. D. thesis by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1974; Includes bibliographical references (pages 261-264) 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89712 Theoretical studies on some aspects of molten fuel-coolant thermal interaction Kazimi, Mujid S. Rapid generation of high pressures and mechanical work may result when thermal energy is transferred from the hot molten nuclear fuel to the coolant in an LMFBR accident. Such energetic thermal interactions are facilitated by the large heat transfer area created when molten fuel is fragmented in the coolant. Two aspects of the molten fuel coolant interaction problem are investigated: (1) the effects of gas/vapor blanketing of the fuel on post-fragmentation generation of pressure and mechanical work, and (2) the mechanism of the fragmentation of the molten fuel as it contacts the coolant. A model developed at Argonne National Laboratory to analyze fragmentation-induced energetic fuel-coolant interactions is modified to allow for gas/vapor blanketing of the fuel. The modified model is applied to a. hypothetical accident involving an FFTF subassembly. The results indicate that high shock pressures are not necessarily precluded by gas/vapor blanketing of the fuel. However, the generation of mechanical work is greatly reduced. A model is developed to simulate the dynamic growth of the vapor film around a hot spherical particle which has been suddenly immersed in a coolant. The model is applied to various cases of hot spheres in water and in sodium. A fragmentation mechanism based on the ability of the pressure pulsations of the vapor film to induce internal cavitation in the molten material is shown to predict the reported fragmentation behavior of drops of several hot molten materials in water and sodium. "May 1973."; Also issued as a Ph. D. thesis, MIT, Dept. of Nuclear Engineering, 1973; Includes bibliographical references (pages 258-269) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89711 Solution of the space-dependent reactor kinetics equations in three dimensions Ferguson, Donald Ross; Hansen, Kent F. A general class of two-step alternating-direction semi-implicit methods is proposed for the approximate solution of the semi-discrete form of the space-dependent reactor kinetics equations. An exponential transformation of the semi-discrete equations is described which has been found to significantly reduce the truncation error when several alternating-direction semi-implicit methods are applied to the transformed equations. A subset of this class is shown to be a consistent approximation to the differential equations and to be numerically stable. Specific members of this subset are compared in one- and two-dimensional numerical experiments. An "optimum" method, termed the NSADE (Non-Symmetric Alternating-Direction Explicit) method is extended to three-dimensional geometries. Subsequent three-dimensional numerical experiments confirm the truncation error, accuracy, and stability properties of this method. "August, 1971."; "MIT-3903-4."; Also issued as a Ph. D. thesis by the first author and supervised by the second author, MIT, Dept. of Nuclear Engineering, 1971; Includes bibliographical references (pages 86-88) 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89710 Time integration methods for reactor kinetics Nóbrega, José de Anchieta Wanderley da; Henry, Allan F. A technique based on the Padé approximations is applied to the solution of the point kinetics equations. The method consists of treating explicitly the roots of the inhour formula which would make the Padé approximations inaccurate. Also, an analytic method is developed which permits a fast inversion of polynomials of the point kinetics matrix and has direct applicability to the Pads approximations. Results are presented for several cases using various options of the method. It is concluded that the technique provides a fast and accurate computational method for the point kinetics equations. Also, an implicit solution method for the time-dependent multigroup diffusion equations known as the "theta method" is studied. Both the usual method and a variation of it, derived from the precursor integrated equations, are considered. Several properties of both versions of the theta method are demonstrated. An attempt is made to find better integration parameters (thetas) for the method, based on corresponding point kinetics calculations. Calculations are done for several test cases, leading to the conclusion that the improvements obtained are of limited value. "December 1971."; Includes bibliographical references (pages 104-107) https://hdl.handle.net/1721.1/89709 Analysis of conventional and plutonium recycle unit-assemblies for the Yankee (Rowe) PWR Mertens, Paul Gustaaf An analysis and comparison of Unit Conventional UO2 Fuel-Assemblies and proposed Plutonium Recycle Fuel Assemblies for the Yankee (Rowe) Reactor has been made. The influence of spectral effects, at the watergaps -and spectral-and transport effects at the UO2 - Mixed Oxide interface, on the powerpeaking has been determined. Two one thermal group methods have been developed for the calculation of powerpeaking in the two dimensional assemblies. The accuracy of the LEOPARD code and LASER code (thermal cut off 1.855 ev) for the calculation of the powerpeaking in conventional and plutonium recycle assemblies has been evaluated. The power distribution and local power peaking factors during burnup, including spectral effects, were also calculated with a macroscopic depletion model. Power gradients inside the peak UO2 rod and peak mixed oxide rod were also determined, and the variations in the heat flux, at the pellet and cladding surface, around these peak pins were calculated. Finally preliminary comparisons of engineering factors for the peak U02 rod and the peak mixed oxide rod have been made. Cover title; At head of title, "Draft report."; "November 1971."; Includes bibliographical references (pages 299-303); Draft report; November 1971 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89708 Use of gamma spectroscopy for neutronic analysis of LMFBR Blankets Kang, Chʻang-sun; Rasmussen, Norman C.; Driscoll, Michael J. It was the purpose of the present investigation to extend and apply Ge(Li) gamma-ray spectroscopy to the study of fast reactor blankets. The focal point for this research was the Blanket Test Facility at the MITR and Blanket No. 2, a realistic mockup of the blanket reflector region of a large liquid metal cooled fast breeder reactor. It was found that Ge(Li) detectors can be simultaneously used as both high energy neutron spectrometers and continuous gamma-ray spectrometers. The broadened internal conversion spectral line at 691.4 KeV has been analyzed for the former purpose, and the Compton recoil continuum has been analyzed and unfolded for the latter. This development makes the Ge(Li) spectrometer an extremely valuable shield analysis tool. The moisture content of the sodium chromate used in the blanket mockup has been confirmed to be less than 0.1 w/o by prompt activation analysis. Prompt capture and inelastic gamma, and decay gamma spectra emitted by the blanket were also analyzed to perform a neutron balance with mixed results. The inability to resolve U-238 prompt capture gammas made it necessary to use the low energy Np-239 decay gammas, with the attendant uncertainties due to large self-shielding corrections. Lack of data on the variation of prompt gamma yield with neutron energy for all blanket constituents also contributed to the uncertainties, which together made it impossible to develop this method to the point where reliable practical application can be recommended. "November, 1971."; Also issued as an Sc. D. thesis by the first author and supervised by the second and third author, MIT Dept. of Nuclear Engineering, 1972; Includes bibliographical references (pages 158-162) https://hdl.handle.net/1721.1/89707 Instrumental methods for neutron spectroscopy in the MIT blanket test facility Ortiz, Nestor Ruben; Rickard, I. C. Massachusetts Institute of Technology; Driscoll, Michael J.; Rasmussen, Norman C. U.S. Atomic Energy Commission contract; The energy spectrum of the neutron flux in a realistic mockup of the blanket region of a large liquid-metal-cooled fast breeder reactor was measured using three different spectrometers: He-3 and Li-6 semiconductor detectors and a Proton-Recoil proportional counter. The He-3 detector was operated in the sum and difference modes, and the Li-6 detector in the sum, difference and triton modes. The experimental data was unfolded using direct, integral and derivative techniques. Methods were developed or perfected to enable use of the He-3 detector over the neutron energy range from 10 keV to 1.3 MeV and the Li-6 detector from 10 keV to 3.1 MeV; the Proton-Recoil detector was operated in the region from 2 keV to 1.5 MeV. In general, good agreement was found between the experimental measurements for all detector types, modes of operation and methods of unfolding, except for the low-energy He-3 data. The present experimental results and previously reported data obtained using a method based on gamma line broadening are in relatively good agreement in the high energy region above 0.8 MeV. The measured neutron spectrum is also similar in shape to neutron spectra measured at ANL in critical assembly mockups of large LMFBR cores, but systematically softer, as expected.. However, there is a large discrepancy in the energy region from 10 keV to 50 keV between the present results and either spectra unfolded from foil data or those numerically calculated using the 1-D ANISN code in the S8 option with 26 energy groups. "May, 1972."; Also issued as an Sc. D. thesis by the first author and supervised by the third and fourth author, MIT Dept. of Nuclear Engineering, 1972; Includes bibliographical references (pages 223-225) 1972-01-01T00:00:00Z https://hdl.handle.net/1721.1/89706 Reactor physics project final report Driscoll, Michael J.; Kaplan, Irving, 1912-; Lanning, David D.; Rasmussen, Norman C.; Agarwala, Vinay Kumar; Clikeman, Franklyn Miles; Hukai, Yoshiyuti; Izzo, Lawrence Leonard; Kazimi, Mujid S.; Leung, Timothy Chung-tim; McFarland, Emerson Lee; Seth, Shivaji Shrilal; Sullivan, G. E. Massachusetts Institute of Technology; Supple, A. T. Massachusetts Institute of Technology This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, 1968 through September 30, 1970. In addition to summarizing results for the entire contract period, this report also serves as the final annual report; thus, work completed in the period of October 1, 1969 through September 30, 1970 is dealt with in more detail than the earlier work. Methods were developed to measure the heterogeneous parameters 17, [Gamma] [eta] and [Alpha] for single fuel elements immersed in moderator in an exponential tank using foil activation measurements external to the fuel. These methods were applied to clustered fuel rods in D 20 moderator and single fuel rods in H 20 moderator, and the results were extended to and compared with data on complete multi-element lattices reported by other laboratories. Advanced gamma spectrometric methods using Ge(Li) detectors were applied to the analysis of both prompt and fission product decay gammas for the nondestructive analysis of the fuel used in this work. The latter includes both simulated burned fuel containing plutonium and actual burned fuel irradiated to 20,000 MWD/T in the Dresden BWR. "September 30, 1970."; Statement of responsibility on title-page reads: Editors, M. J. Driscoll, I. Kaplan, D. D. Lanning, N. C. Rasmussen. Contributors: V. K. Agarwala, F. M. Clikeman, M. J. Driscoll, Y. Hukai, L. L. Izzo, I. Kaplan, M. S. Kazimi, D.D. Lanning, T.C. Leung, E.L. McFarland, N.C. Rasmussen, S.S. Seth, G.E. Sullivan, and A.T. Supple; Includes bibliographical references; Final report; January 1, 1968 to September 30, 1970 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89705 The reactor physics of the Massachusetts Institute of Technology reactor redesign Addae, Andrews Kwasi; Lanning, David D.; Thompson, Theos Jardin, 1918-1970 An H20 cooled compact MITR-II core, reflected by D20 has been designed for the MITR to increase the reflector thermal neutron flux at tips of beam ports by a factor of 3 or better, without changing the operating power level of the reactor. The diffusion approximation to the neutron transport equation has been used. A three neutron energy group scheme, that retains essential spatial effects, used in the studies has yielded satisfactory agreement with measured data. The factors which affect the intensity as well as the quality of the reflector thermal neutron flux have been studied. These studies show that the permanent features of the MITR limit the maximum power densities in the MITR-II core to factors between 4.5 and 12 below the corresponding values in reactors employing a similar core concept Nevertheless, the predicted unperturbed reflector thermal neutron flux of 1.lXlO14 n/cm 2-sec in MITR-II yields a reflector flux per unit power that is competitive with the corresponding values available in reactors of its type and a factor of 5.0 higher than that in MITR-I. "August, 1970."; Also written as a Ph. D. thesis by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1970; Includes bibliographical references (pages 284-289) 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89704 Some applications of Ge(Li) gamma-ray spectroscopy to fuel element assay Hukai, Yoshiyuti; Driscoll, Michael J.; Rasmussen, Norman C. It was the object of this work to study the gamma rays emitted by the products of the interaction of thermal neutrons with the nuclei of U-238, Th-232, U-235 and Pu-239 during and after irradiation and to explore some applications mainly to fuel element assay. An irradiation facility and a Ge(Li) detector cryostat were constructed for this purpose. A new method of assaying a fuel rod containing a mixture of plutonium and uranium oxide, based on the difference in the observed yield of the fission products 1-135 and Sr-92, has been developed. The energies and intensities of the thermal neutron capture gamma rays for U-238 and Th-232 were determined. Four new lines have been found in the energy region previously unexplored for U-238. For Th-232, 66 certain lines were found, compared to 7 lines in the literature. Many prompt gammas emitted 'by the highly excited fission products following the fission of U-235 and Pu-239 were resolved in the energy region above 1.4 MeV. For U-235 fissions, 57 lines were found, and for Pu-239, 51 certain lines were recorded. The use of prompt gammas for assaying fuel rods was investigated. An accuracy of about ± 7% was obtained for the analysis of U-238 content; ± 10% to ± 20% accuracy was obtained for U-235 analysis in the range of 1% to 2% enrichment; and ± 35% accuracy for the analysis of 0.25% Puenriched rods. It has been found that Ge(Li) detectors can be operated as fast neutron detectors and used to determine the relative neutron yield. With this method, the enrichment of uranium rods can be found with an accuracy of ± 1% to ± 2% in the range from 116 to 2% enrichment. Finally, some considerations were given to the use of prompt gamma rays for measuring the initial conversion ratio C and the neutron yield parameter [eta]. "MIT-3944 -5."; Also issued as a Ph. D. thesis by the first author and supervised by the second and third author, MIT Dept. of Nuclear Engineering, 1970; Includes bibliographical references (pages 195-198) 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89703 Study of gamma rays from neutron inelastic scattering Hui, Bertram Ho Wai; Rasmussen, Norman C. The energy and intensity of the inelastic gamma rays of twenty low atomic number elements: Li, C, N, 0, Mg, Al, Na, Si, S, Cl, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, and Pb, are measured with a 30 cc Ge(Li) detector using an unmoderated 5 curie Pu-Be neutron source. The value of the production cross section for the Pu-Be spectrum is calculated. The results show they are mostly less than one barn. Cover title; "February 1970."; "Prepared for United States Department of the Interior Bureau of Mines Morgantown Research Center Morgantown, West Virginia."; Includes bibliographical references (page 59) 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89702 Reactor physics project progress report no. 2 Driscoll, Michael J.; Kaplan, Irving, 1912-; Lanning, David D.; Agarwala, Vinay Kumar; Clikeman, Franklyn Miles; Donohew, Jack Norman; Hamilton, George Thomas; Harper, Thomas Lawrence; Hukai, Yoshiyuti; Kelley, Thomas James; Leung, Timothy Chung-tim; McFarland, Emerson Lee; Rasmussen, Norman C.; Seth, Shivaji Shrilal; Sicilian, J. M., B.S. Massachusetts Institute of Technology; Sullivan, G. E. Massachusetts Institute of Technology; Supple, A. T. Massachusetts Institute of Technology; Thompson, Theos Jardin, 1918-1970 This is the second annual report in an experimental and theoretical program to develop and apply single and few element heterogeneous methods for the determination of reactor lattice parameters. During the period covered by the report, October 1, 1968 through September 30. 1969, work was primarily devoted to measurement of the heterogeneous fuel element parameters (F, rl and A) of 19- and 31- rod clusters of plutonium-containing fuel. Methods development research focused on determination of the epithermal absorption constant, A. Calculations and an analysis of data reported in the literature were made to assess the applicability of heterogeneous methods to H 20- moderated systems. Advanced gamma spectrometric methods using Ge(Li) detectors were applied to the analysis of prompt and delayed gamma spectra from fertile and fissile materials and from fuel elements. These methods were used successfully for nondestructive analysis of the composition of fuel elements. A feasibility study was performed on an in-pile gamma spectrometer. Two fuel pins irradiated to a burnup of approximately 20,000 MWD/MT in the Dresden reactor were received and preparations made for their analysis and use in reactor physics experiments. Statement of responsibility on title page reads: Editors: M.J. Driscoll, I. Kaplan, D.D. Lanning; Contributors: V. Agarwala, F.M. Clikeman, J.N. Donohew, M.J. Driscoll, G. T. Hamilton, T.L. Harper, Y. Hukai, I. Kaplan, T. J. Kelley, D.D. Lanning, T.C. Leung, E.L. McFarland, N.C. Rasmussen, S.S. Seth, J.M. Sicilian, G.E. Sullivan, A.T.Supple and T.J. Thompson; "September 30, 1969."; "MIT-3944-4."; Includes bibliographical references; Progress report no. 2; October 1, 1968 through September 30. 1969 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89701 Investigation of elemental analysis using neutron-capture gamma ray spectra Hamawi, John Nicholas; Rasmussen, Norman C. This thesis evaluated the potential of neutron-capture gamma rays in elemental analysis. A large portion of the work was devoted to the development of a method for the analysis of weak peaks in gamma ray spectra. This was based on equations developed for the standard deviation in the measurement of the various peak parameters, consideration being also given to the reduction in the statistical fluctuations obtained by smoothing the data with the use of Fourier transforms. Two methods of peak area determination were considered end their relative effectiveness examined. An equation was then derived for the minimum weight of an element needed for reliable quantitative analysis. The equations were verified using both real and pseudo-experimental data constructed with the use of a computer. Experiments were carried out using the MIT Reactor with samples positioned La) in a high neutron flux next to the reactor tank (2xl01- n/sq.cm sec), and (b) in an external neutron beam facility of relatively lower but well thermalized flux (2xl0 n/sq.cm sec). Capture gamma ray spectra were obtained with a three-crystal system capable of operating in the free mode, the Compton suppression mode and as a pair spectrometer. The results were used to examine the relative analytical sensitivity of the internal and external sample arrangements and the various gamma detection modes. The minimum measurable weights of 75 elements were evaluated for a stainless steel sample. For these computations use was made of the listing of capture gamma ray spectra recently established by the MIT gamma spectroscopy group. 'In a majority of the cases the detection limits range between 0.1 percent and 10 percent. Equations were developed for extending the results to different samples and different. experimental arrangements. "September 1969."; "Prepared for United States Department of the Interior Bureau of Mines Morgantown Research Center Morgantown, West Virginia."; Also issued as a Ph. D. thesis, written by the first author and supervised by the second author, MIT, Dept. of Nuclear Engineering, 1969; Includes bibliographical references (pages 307-312) 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89700 GAMANL : a computer program applying Fourier transforms to the analysis of gamma spectral data Harper, Thomas Lawrence; Inouye, Tamon; Rasmussen, Norman C. GAMANL, a computer code for automatically identifying the peaks in a complex spectra and determining their centers and areas, is described. The principal feature of the method is a data smoothing technique employing Fourier transforms. The smoothing eliminates most of the random fluctuations without effecting the spectral resolution and makes identification of maxima using a zero slope criterion possible. Using the same Fourier transform with different constants it is possible with a second transformation to improve the spectral resolution. The computer program has been written in FORTRAN IV for the M.IT. IBM 360 model 65 computer and also for the Toshiba Electric Company G.E. 635 computer. The complete analysis of a 4096 channel spectrum containing one hundred twenty peaks requires about 75 seconds of computation time. "August 1968."; "MIT-3944-2."; Includes bibliographical references (pages 86-87) 1968-01-01T00:00:00Z https://hdl.handle.net/1721.1/89699 Determination of thermal neutron capture gamma yields Harper, Thomas Lawrence; Rasmussen, Norman C. A method of analysing Ge(Li) thermal neutron capture gamma spectra to obtain total gamma yields has been developed. Tie method determines both the yields from the well resolved gamma peaks in a spectrum as well as the gamma yields from the unresolved gamma lines which appear in the continuum portion of a spectrum. Accounting for the unresolved continuum enables a large fraction of total emitted energy to be observed, and values of 100% +/- 15% are obtained for the cases studied. The techniques used involve the determination of a peak response function suitable for the Ge(Li) pair spectrometer spectra being studied. The response function is used to strip off the effects of the peaks upon the background and unresolved data continuum. The continuum, which is due only to the unresolved lines, is broken into energy bins of 210 keV width and the gamma yield per bin is calculated. Results of the analysis and normalized yields are given for the rare earth samples of: Nd, Sm, Eu, Gd, and Er. The capture data were obtained by using the MITR 4th irradiation facility operated with a Ge(Li) pair spectrometer. "July 1969."; Vita; Also issued as a Ph. D. thesis by the first author and supervised by the second author, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1969; Includes bibliographical references (volume 2, pages 383-385) 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89698 Research and educational activities at the MIT Research Reactor : Fiscal year 1968 A report of research and educational activities which utilized the Massachusetts Institute of Technology, five-megawatt, heavy water, research reactor during fiscal year 1968 has been prepared for administrative use at MIT and for presentation to the U. S. Atomic Energy Commission. The latter action is required by Contract AT(30-1)-1967 under which the AEC provides the fully-enriched uranium-235 fuel and the heavy-water moderator-coolant for the reactor. Research projects at MIT which make significant use of the MITR are described, and principal participating personnel are named. Listings are provided of theses, reports, journal articles, and conference papers resulting from these projects during fiscal year 1968. A comprehensive bibliography of earlier publications was contained in Report No. MITNE-91 "Research and Educational Activities at the MIT Research Reactor To and Including Fiscal Year 1967". That report lists essentially all documents of these types which are concerned with the design, operation, or research use of the MITR from the time its construction was first contemplated at the Institute until June 30, 1967. In addition to the educational value derived from the many research activities by the students who participated in them, training in several areas of nuclear technology is imparted through formal courses designed to make use of the reactor or its research projects. The courses are briefly described, and attendance figures are given. Detailed information concerning the research activities of the numerous other universities, hospitals, and commercial companies which have used the MITR for irradiations is not available, but these organizations and also the materials irradiated are listed. The reactor, its purpose, its organization, and a summary of operations are briefly described in order to provide a more complete understanding of the MITR program. "December 1968."; Includes bibliographical references (pages 141-143); Activities progress report; July 1, 1967 to June 30, 1968 1968-01-01T00:00:00Z https://hdl.handle.net/1721.1/89697 Effect of reactor irradiation on Santowax OM and WR Mason, Edward A. (Edward Archibald), 1924-; Lee, Min-Li; Brewer, Shelby Templeton; Bley, W. N. (William Norman) Statement of responsibility on title page reads: E.A. Mason, M.L. Lee, S.T. Brewer, and W.N. Bley; "Issued: June, 1968."; Includes bibliographical references 1968-01-01T00:00:00Z https://hdl.handle.net/1721.1/89696 Technical specifications for the MIT Research Reactor "August 16, 1965."; Includes bibliographical references (pages 105-106) 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/89695 A generalized study of the breeding potential of large heavy water moderated power reactors fueled with thoria and urania Richardson, Max Cotner; Benedict, Manson; Mason, Edward A. (Edward Archibald), 1924- "January 1967."; "MIT-2073-5."; Also issued as a Ph. D. thesis by the first author and supervised by the second and third authors, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1966; Includes bibliographical references (pages 324-328) 1967-01-01T00:00:00Z https://hdl.handle.net/1721.1/89694 Proposed methods for determining the effect of U-236 and Np-237 on the value of uranium as feed for pressurized water power reactors / by D.A. Goellner and M. Benedict Goellner, Donald A., 1936-; Benedict, Manson Cover title; "June 1966."; Includes bibliographical references (pages 47-48) 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89693 Friction factor and heat transfer correlation for irradiated organic coolants Swan, Arthur Henry; Mason, Edward A. (Edward Archibald), 1924-; Bley, W. N. (William Norman); Kim, Je Chul; Morgan, Dean T. "September 1965."; Series statement handwritten on cover; "MIT-334-23 Chemistry."; Also written as an M.S. theses written by the first author and advised by the second author, Massachusetts Institute of Technology, Dept. of Mechanical Engineering, 1966; Includes bibliographical references (pages 163-165) 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/89692 Use of a moments method for the analysis of flux distributions in subcritical assemblies Cheng, Hsiang-Shou; Kaplan, Irving, 1912-; Thompson, Theos Jardin, 1918-1970; Driscoll, Michael J. A moments method has been developed for the analysis of flux distributions in subcritical neutron-multiplying assemblies. The method determines values of the asymptotic axial and radial buckling, and of the extrapolated height and radius, from foil activation data, in terms of flux moments defined in the usual sense. Analytic expressions are derived for the axial and radial buckling and extrapolated dimensions in terms of the flux moments. These expressions have clear physical meaning and are suitable for the interpretation of conventional buckling measurements. The method treats the moment index as a variable parameter and allows freedom in the choice of the locations of the first and last data points used in the analysis. These degrees of freedom make it possible to reduce the effects of source neutrons, flux transients, and higher harmonics. As a result, the moments method can be applied successfully to very small lattices ("miniature lattices") as well as to large exponential assemblies. The moments method has been tested, in comparison with the conventional least-squares curve-fitting method, by applying the two methods to the analysis of measurements made in several uranium heavy water, and uranium oxide-heavy water lattices investigated at the M. I. T. Lattice Project. In the case of large exponential assemblies, the moments method yielded more consistent results than the curve-fitting method. In the case of miniature lattices, the moments method made it possible for the first time to determine values of axial and radial buckling and extrapolated dimensions. Statement of responsibility on title-page reads: H.S. Cheng, I. Kaplan, T.J. Thompson, M.J. Driscoll; "MIT-2344 -11."; Includes bibliographical references (pages 347-352) 1968-01-01T00:00:00Z https://hdl.handle.net/1721.1/89691 Computer simulation of neutron capture therapy Olson, Arne Peter Analytical methods are developed to simulate on a large digital computer the production and use of reactor neutron beams f or boron capture therapy of brain tumors. The simulation accounts for radiation dose distributions in tissue produced by fast neutrons and by neutron capture reaction products such as gamma rays, C -particles, protons, and heavy particles. These techniques are applied to optimize the effectiveness of the M.I.T. Reactor Medical Therapy Facility through a survey of the effects of neutron filters and of modifications to the beam collimation system. Neutron beams reflected from thin slabs of hydrogenous materials are shown to have an improved ability to effectively irradiate a deep tumor without destroying normal tissue above it because relatively few fast neutrons are reflected. Considerable improvements in thermal neutron distribution in tissue are shown to result from surrounding the head with a neutron-reflecting annulus to reduce lateral neutron leakage. A new numerical solution is obtained for the problem of neutron transport in finite thickness slabs with isotropic scattering. Gaussian quadratures are used to evaluate the neutron transport integral equations, yielding transmission, absorption, and reflection probabilities, and fluxes, as a function of collision number. Collision history correlations are devised which use only five paraeters to predict the fate of neutrons incident on an infinite slab having arbitrary thickness and neutron cross sections. A very fast multigroup neutron spectrum calculation is developed by combining collision history correlations with single-collision group transfer probabilities to directly obtain transmission and reflection matrices for multi-slab shielding problems. "August 1967."; "Prepared for Physics Research Laboratory Massachusetts General Hospital Boston, Massachusetts."; Also issued as an Sc. D. thesis, MIT, Dept. of Nuclear Engineering, 1967; Includes bibliographical references (pages 340-343) 1967-01-01T00:00:00Z https://hdl.handle.net/1721.1/89690 Study of thermal neutron capture gamma rays using a lithium-drifted germanium spectrometer / [by] Victor John Orphan [and] Norman C. Rasmussen Orphan, V. J.; Rasmussen, Norman C. A gamma-ray spectrometer, using a 30 cc coaxial Ge(Li) detector, which can be operated as a pair spectrometer at high energies and in the Compton suppression mode at low energies provides an effective means of obtaining thermal neutron capture gamma spectra over nearly the entire capture gamma energy range. The energy resolution (fwhm) of the spectrometer is approximately 0.5% at 1 MeV and 0.1% at 7 MeV. Capture gamma-ray energies can be determined to an accuracy of about 1 keV. The relatively high efficiency of this spectrometer allows the use of an external neutron beam geometry, which simplifies sample changing. Using a 4096 channel pulse height analyzer, the capture gamma spectrum of an element may be obtained in about one day. Low cross section (order of 0.1 b) elements with many weak intensity gammas may be studied. Over 100 gamma rays have been identified in the spectrum of one such element, Zr. The spectra of Be, Sc, Fe, Ge, and Zr are presented. "January 1967."; "AFCRL-67-0104."; Also issued as an Sc. D. thesis by the first author and advised by the second author, MIT, Dept. of Nuclear Engineering, 1967; Includes bibliographical references (pages 199-203); Scientific report, interim; January 1967 1967-01-01T00:00:00Z https://hdl.handle.net/1721.1/89689 The Use of experiments on a single fuel element to determine the nuclear parameters of reactor lattices Pilat, E. E., 1937-; Driscoll, Michael J.; Kaplan, Irving, 1912-; Thompson, Theos Jardin, 1918-1970 The nuclear parameters of a reactor lattice may be determined by critical experiments on that lattice, by theoretical calculations in which only cross sections are used as input, or by methods which combine theory and experiment. Of those methods which combine theory and experiment, the Single Element Method, abbreviated SEM, is shown to have great usefulness. As used here, the method combines experiments on the smallest meaningful unit of fuel - a single fuel element - with a theory which relates the behavior of a lattice of such elements to the experimentally determined behavior of the single element. This particular division of the problem into theory and experiment is useful for at least three reasons.; First, several parameters which characterize a reactor lattice - the thermal utilization and resonance escape probability, for example - often depend strongly and in a complicated manner on the properties of individual fuel elements, but only depend weakly or in a simple manner on interactions between the fuel elements. In the Single Element Method, the largest contribution to these parameters is determined by measurements on a single fuel element, and only a relatively small correction to account for the presence of the rest of the fuel elements need be estimated theoretically. Second, the determination of lattice parameters in this way represents a desirable saving of time, money, effort, and material over their determination in critical or exponential experiments.; Third, it is shown that the method provides an excellent way of correlating the results of experimental measurements, since it shows what pertinent variables must be used to express the quantity of interest in a linear or nearly linear fashion. Values obtained by the SEM for the thermal utilization of lattices of uranium rods in heavy water are accurate to about 0.3 percent (by comparison with THERMOS). Values of P28, 28, and C* are obtained by the SEM for the same lattices to an accuracy of between five and ten percent (by comparison with experiment). The same method yields values of 28 with are equally accurate in lattices moderated by light water. In addition, the theoretical development of the SEM predicts that P28, 28, C*, and 625 should vary nearly linearly with the inverse of the unit cell volume (for a fixed size of fuel element). This explains the experimentally observed behavior and provides an important tool for the rational correlation of experimental results. Statement of responsibility on title page reads: Edward E. Pilat, M.J. Driscoll, I. Kaplan, and T.J. Thompson; "February 1967."; Includes errata; "MIT-2344 -10"; Also issued as a Ph. D. thesis by the first author (Pilat) and supervised by the last author (Thompson), Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1967; Includes bibliographical references 1967-01-01T00:00:00Z https://hdl.handle.net/1721.1/89688 PULSE : an IBM 7094 program for calculation of fast neutron kinetics by Monte Carlo. Progress report, October, 1963 Profio, A. Edward, 1931- Progress report; October, 1963 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89687 The effects of reactor irradiation on Santowax OMP at 610°F and 750°F Sawyer, Craig D. (Craig Delany) Santowax OMP has been irradiated in the M.I.T. In-Pile Loop Facility at 610°F and at 750°F. At both temperatures the loop was operated in a transient phase and a steady-state-HB phase. In the transient phase, unirradiated material was allowed to degrade to 60 w/o DP. In the steadystate- HB phase, the HB content of the coolant was maintained constant at about 33 w/o by the removal and distillation of samples and the replacement of the HB by unirradiated material before returning the samples to the loop. Neutron and gamma ray doses were measured with adiabatic calorimeters and foil monitors. The average dose rate to the coolant in the core region of the in-pile section was about 0. 5 watts/gm, of which 37% was due to fast neturon interactions and 63% to gamma ray interactions. Terphenyl concentrations were measured by gas chromatography and HB concentrations by distillation.; Analysis of the transient phase terphenyl concentration and absorbed dose data showed that first order kinetics provided an adequate description of the degradation rate of the terphenyls. At 610°F no significant difference in the stabilities of the terphenyl issomers was found and the overall degradation rate of the coolant was G*(-omp) = G(-omp)/Comp = 0. 26 ± 0. 01 molecules of terphenyl degraded per 100 ev absorbed in the terphenyls. At 750°F the terphenyl isomer stabilities were in the order para>meta>ortho. After corrections for out-of-pile pyrolysis the overall degradation rate of the coolant was G*(-omp) = 0. 49 ± 0. 02. The results are compared to those of other investigations. For the 610°F irradiation the radiolytic gas generation rate was G(total gas) = 0.037 ± 0.003 molecules of gas produced per 100 ev absorbed in the coolant mixture, the principal product being hydrogen.; During the 750°F irradiation the generation rate was G(total gas) = 0.105 ± 0.008, with a marked increase in the evolution of methane. Physical property measurements included density, viscosity, specific heat, thermal conductivity, number average molecular weight, gas solubility, carbon-hydrogen content and ash content. The increase in viscosity with increasing DP concentration was significantly less for the 750°F irradiation than for the 610*F irradiation. Heat transfer measurements showed that standard correlations could be used to determine the heat transfer rates using the physical properties of the irradiated coolant. The correlation obtained for the data of both irradiations was Nu = 0. 0079(Re)0.9(Pr) 0 . 4 ± 10%. No evidence of scale buildup on the heat transfer surfaces was observed over the entire period of operation of the experiment. The results of preliminary measurements with a fouling probe are also reported. "September 1963."; "IDO-11, 107."; Also issued by the first author as an Sc. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1964; Includes bibliographical references 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89686 Measurements of the spatial and energy distribution of thermal neutrons in uranium, heavy water lattices Brown, Paul S. (Paul Sherman); Thompson, Theos Jardin, 1918-1970; Kaplan, Irving, 1912-; Profio, A. Edward, 1931- Intracell activity distributions were measured in three natural uranium, heavy water lattices of 1. 010 inch diameter, aluminum clad rods on triangular spacings of 4. 5 inches, 5. 0 inches, and 5. 75 inches, respectively, and in a uranium, heavy water lattice of 0. 25 inch diameter, 1. 03% U 2235, aluminum-clad rods on a triangular spacing of 1. 25 inches. The distributions were measured with bare and cadmium-covered foils of gold, lutetium, and europium. The gold was used as a 1/v absorber to measure the thermal neutron density distribution. Because the activation cross sections of lutetium and europium depart considerably from 1/v behavior, their activation depends strongly on the thermal neutron energy spectrum. Hence, they were used to make integral measurements of the change in the neutron energy spectrum with position in the lattice cell. A method was developed for treating the partial absorption, by cadmium covers, of neutrons at the 0.; 46 ev europium resonance, and it was found possible to correct the europium activations to energy cutoffs just above and just below the resonance. The measured activity distributions were compared with those computed with the THERMOS code. In the natural uranium lattices, THERMOS gave excellent agreement with the measured gold activity distributions and very good agreement with the lutetium and europium distributions, indicating that THERMOS gives a very good estimate of the spatial and energy distribution of thermal neutrons in these lattices. In the enriched lattice, THERMOS gave a large overestimate of the activity dip in the fuel for all three detectors. The discrepancy was attributed to a breakdown in the Wigner-Seitz cylindrical cell approximation at small cell radii.; However, the measured ratios of lutetium and europium activity to gold activity were in good agreement with the THERMOS values, indicating that THERMOS still gave a good estimate of the degree of spectral hardening. Neutron temperature calculations were made from the data by using Westcott effective cross sections. The temperature changes so calculated agreed well with those predicted by THERMOS. Disadvantage factors calculated by the Amouyal-Benoist-Horowitz (ABH) method were in excellent agreement with the measured values in the natural uranium lattices. The agreement was not as good in the enriched lattice because of an expected breakdown in the ABH method at small cell radii. Values of the thermal utilization obtained from experiment, from THERMOS, and with the ABH method were in excellent agreement for all the lattices studied.; Radial and axial buckling measurements made with lutetium were in excellent agreement with similar measurements made with gold, indicating that the thermal neutron spectrum was uniform throughout the lattice tank. Measurements of intracell gold activity distributions made in off-center cells differed only slightly from those made in the central cell of the lattice, indicating that the radial flux distribution was almost completely separable into a macroscopic Jo and a microscopic cell distribution. "August 20, 1962."; Statement of responsibility on title-page reads: P. S. Brown, T. J. Thompson, I. Kaplan, A. E. Profio; Also issued by the first author as a Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1962; "NYO-10205."; Includes bibliographical references (pages 185-210) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89685 Equilibrium extraction characteristics of alkyl amines and nuclear fuel metals in nitrate systems: progress report for the period July 1 - December 31, 1961 : progress report X Mason, Edward A. (Edward Archibald), 1924-; Skavdahl, Richard E. (Richard Earl), 1934- "February 15, 1962."; Includes bibliographical references (page 62); Progress report no. 10; July 1 to December 31, 1961 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89684 First annual report : organic moderator-coolant in-pile irradiation loop for the MIT nuclear reactor : October 1, 1958 to October 1, 1959 Mason, Edward A. (Edward Archibald), 1924-; Bley W. N. (William Norman); Morgan Dean Thomas Includes bibliographical references (leaf 34); First annual report; to October 1, 1958 to October 1, 1959 1961-01-01T00:00:00Z https://hdl.handle.net/1721.1/89683 The Effect of fuel and poison management on nuclear power systems McLeod, Norman Barrie; Benedict, Manson; Uematsu, Kunihiko; Witting, Harald L. (Harald Ludwig); Ram, K. S. Statement of responsibility on title page reads: N.B. McLeod, M. Benedict, K. Uematsu, H.L. Witting, and K.S. Ram; "September 15, 1961."; Submitted by the first author as a Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1962; "NYO-9715, TID 4500 Category, UC-80 Reactor Technology."; "This work was done in part at the MIT Computation Center."; Includes bibliographical references (p. 492-496); Report; June, 1959 - September, 1961 2014-09-16T00:00:00Z https://hdl.handle.net/1721.1/89682 A review of recent analytical and experimental studies applicable to LMFBR fuel and blanket assembly design / by E. Khan and N. Todreas Khan, Eshan Ullah; Todreas, Neil E. "September, 1973."; Includes bibliographical references (pages [40]-[44]) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89681 A coarse-mesh nodal diffusion method based on response matrix considerations Henry, Allan F.; Sims, Randal Nee The overall objective of this thesis is to develop an economical computational method for multidimensional transient analysis of nuclear power reactors. Specifically, the application of nodal methods based on the multigroup diffusion theory approximation to reactors composed of regular arrays of large homogeneous (or homogenized) zones was investigated. A nodal scheme is formulated using the response matrix approach as a conceptual basis. Solutions of equivalent sets of coupled one dimensional problems are used to treat the local multidimensional response problems. Polynomial expansions in conjunction with weighted residual procedures are employed to obtain approximate solutions of the one-dimensional problems. A linear set of nodal equations expressed in terms of nodal average fluxes and interface average partial currents is obtained. Applications to two-dimensional few-group, static and transient problems demonstrate that the nodal scheme can be an order of magnitude more computationally efficient than conventional finite difference methods. "March 1977."; Originally issued as the 2nd author's Sc. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1976; "Electric Power Research Institute."; Includes bibliographical references (pages 152-155) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89680 A core reload pattern and composition optimization methodology for pressurized water reactors Sauer, Ildo Luis; Driscoll, Michael J. The primary objective of this research was the development of a comprehensive, rapid and conceptually simple methodology for PWR core reload pattern and fuel composition optimization, capable of systematic incorporation of constraints, in which cycle burnup is defined as the optimality criterion. A coarse mesh nodal method for PWR core analysis was formulated by coupling the one-and-one-half-group diffusion theory model for spatial power calculations with the linear reactivity versus burnup model (LRM) for depletion calculations. The accuracy and suitability of this model was determined through comparisons of its results with those of state-of-the-art core analysis methods.; The simplicity of the LRM-based core model allowed the direct analytical computation of the derivatives necessary in the steepest gradient type optimization methods applied in the present work, and its versatility permitted use of the analytical and computational methods for a variety of applications, ranging from core reload pattern searches to burnable poison (BP) and composition optimization. Algorithms for identification of unconstrained maximum-burnup core reload patterns and for optimal BP allocation were successfully implemented and tested, and the basis for systematic incorporation of constraints on power peaking was developed. The potential application of the methodology to fuel composition optimization was also examined. Most of the methodological developments have been embodied in the LRM-NODAL code which was programmed in the course of this research.; From the numerical and analytical results it was found that the optimal core configurations are arranged such as to produce power histories and profiles in which the most reactive assemblies are at their highest allowable power at EOC (thus maximizing their importance) and where the converse applies to the least reactive; these preferred profiles also produce relatively higher leakage at EOC, evolving to the lowest possible leakage at EOC, but always consistent with the maximization of the core reactivity importance. "March 1985."; Originally issued by the first author as a Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1985; Includes bibliographical references (pages 277-283) 1985-01-01T00:00:00Z https://hdl.handle.net/1721.1/89679 Boiling heat transfer for high velocity flow of highly subcooled water Lekakh, Boris; Kazimi, Mujid S.; Meyer, John E. "October 1998."; Includes bibliographical references (p. 27-28) 1999-01-01T00:00:00Z https://hdl.handle.net/1721.1/89678 Review of applicable U.S. Department of Energy and U.S. Nuclear Regulatory Commission activities : Project task 1 Apostolakis, G.; Golay, M.; Chaniotakis, E. A. (Emmanouil A.); Borgonovo, Emanuele, 1970-; Felder, Frank Andrew; Ghosh, Suchandra Tina, 1973-; Rempe, Joy (Joy L.); Leahy, Timothy. INEEL; Knudson, Darrell Lee At head of title: Regulatory excellence project : performance-based regulatory framework for U.S. Department of Energy facilities; Statement of responsibility on title-page reads: George Apostolakis (Principal Investigator), Michael Golay (Principal Investigator), Emmanuel Chaniotakis, Emanuele Borgonovo, Frank Felder, S. Tina Ghosh and Yu Sui; INEEL: Joy Rempe, Timothy Leahy and Darrell Knudson; "June 1999."; Includes bibliographical references (pages 44-48) 1999-01-01T00:00:00Z https://hdl.handle.net/1721.1/89677 Processing incentives for Russian weapons-grade plutonium Sylvester, Kory William Budlong Includes bibliographical references (page 10) 1997-01-01T00:00:00Z https://hdl.handle.net/1721.1/89676 Dynamic event tree analysis method (DETAM) for accident sequence analysis Acosta, C. (Crispiniano); Siu, N. O. (Nathan O.) Includes bibliographical references (pages 133-138); Final report 1991-01-01T00:00:00Z https://hdl.handle.net/1721.1/89674 The 1994 MIT NED-DOE intern program Meyer, John E. Includes bibliographical references 1994-01-01T00:00:00Z https://hdl.handle.net/1721.1/89673 Uncertainty evaluations for MITR-3 applications : final report for MITNE-305 Sardy, Sylvain Includes bibliographical references (leaves 44-45); Final report 1994-01-01T00:00:00Z https://hdl.handle.net/1721.1/89672 The Chernobyl accident revisited : source term analysis and reconstruction of events during the active phase Sich, Alexander Roman; Borovoi, Aleksandr A.; Rasmussen, Norman C. Includes bibliographical references (pages 489-495) 1994-01-01T00:00:00Z https://hdl.handle.net/1721.1/89631 Thermal neutron capture gamma-ray spectra of the elements Rasmussen, Norman C.; Hukai, Yoshiyuti; Inouye, Tamon; Orphan, V. J. "January 1969."; Statement of responsibility on title-page reads: Norman C. Rasmussen, Yoshiyuti Hukai, Tamon Inouye, Victor J. Orphan; "Prepared for Air Force Cambridge Research Laboratories, Office of Aerospace Research, United States Air Force, Bedford, Mass."; "AFCRL-69-0071."; Includes bibliographical references 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89630 A comparative assessment of the LMFBR and advanced converter fuel cycles with quantification of relative diversion resistance Heising, Carolyn D. (Carolyn DeLane), 1952-; Saragossi, Isi Issac; Sharafi, Mohammad Prepared for Electric Power Research Institute, Palo Alto California; Includes bibliographical references; Final report; September 1979 1979-01-01T00:00:00Z https://hdl.handle.net/1721.1/89629 Quantitative methods for assessing nuclear fuel cycle diversion resistance Heising, Carolyn D. (Carolyn DeLane), 1952-; Miller, Marvin M. "Prepared for Department of Energy, Nuclear Alternatives System Assessment Program, Washington, D.C."; Includes bibliographical references; Final report; October 1979 1979-01-01T00:00:00Z https://hdl.handle.net/1721.1/89628 An assessment of internal blankets for gas-cooled fast reactors Lancaster, Dale Burkham; Driscoll, M. J. Originally presented as first author's thesis (Ph. D.--Massachusetts Institute of Technology), 1980; Includes bibliographical references (pages 338-346) 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89627 Generic nuclear safety issues : methods of analysis Heising, Carolyn D. (Carolyn DeLane), 1952-; Gordon, Ethel Sherry; Lepervanche-Valencia, José Gregorio; Dykes, Andrew Arthur "Prepared for: Nuclear Safety Analysis Center."; Includes bibliographical references (leaves 223-231) 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89626 Transient response of a single heated channel Lee, Min; Kazimi, Mujid S. Includes bibliographical references (page 20) 1984-01-01T00:00:00Z https://hdl.handle.net/1721.1/89625 Modeling of corium/concrete interaction Lee, Min; Kazimi, Mujid S. Includes bibliographical references (pages 232-237) 1985-01-01T00:00:00Z https://hdl.handle.net/1721.1/89624 A time-dependent, two-phase, thermal hydraulic feedback model for the Nodal code QUANDRY Khalil, Hussein Shoukry 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89623 A static, two-phase, thermal hydraulic feedback model for the Nodal code QUANDRY Khalil, Hussein Shoukry Includes bibliographical references (leaf 20) 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89622 Demonstration of methods for analytical measurement of natural circulation flow in EBR-II Witt, R. J. (Robert James); Meyer, John E.; Choi, J. I. (Jung I.); Lanning, David D.; Schor, A. L.; Wittmeier, Richard Dean Statement of responsibility on title page reads: R. J. Witt and J. E. Meyer, Includes MIT technical contributions from J. I. Choi, D. D. Lanning, J. E. Meyer, A. L. Schor, R. J. Witt and R. D. Wittmeier."; "February, 1986."; Includes bibliographical references (leaf 44); Final project report 1986-01-01T00:00:00Z https://hdl.handle.net/1721.1/89621 Laminar film condensation in zero gravity Stenning, Alan H. (Alan Hugh); Hooper, Richard Jon; Korn, Donald Harry, 1935-; Yoshitani, Yutaka, 1928- Statement of responsibility on title-page reads, A. H. Stenning, R. J. Hooper, D. H. Korn, and Y. Yoshitani; "June 1960."; Series statement handwritten on cover; Includes bibliographical references (page 25) 1960-01-01T00:00:00Z https://hdl.handle.net/1721.1/89620 MITR-II start-up report "February 14, 1977."; Report; January 1977 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89619 Irradiation loop studies of organic reactor coolants Mason, Edward A. (Edward Archibald), 1924-; Bley W. N. (William Norman) Includes bibliographical references (leaf 26); Quarterly report; July 1-September 30, 1961 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89618 Computer techniques for sensor validation during EBR-II natural circulation Witt, R. J. (Robert James); Meyer, J. E. "November, 1984."; "Includes MIT technical contributions from D.D. Lanning, J.E. Meyer, A.L. Schor, R.J. Witt and R.D. Wittmeier."; "U.S. Dept. of Energy Breeder Technology Program, Division of Educational Programs, Argonne National Laboratory."; Includes bibliographical references (leaf 17); Final project report; November, 1984 1984-01-01T00:00:00Z https://hdl.handle.net/1721.1/89617 Neutron thermalization at a temperature discontinuity Carlson, Roger Willard Based on the author's Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1966; Includes bibliographical references (leaves 81-82) 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89616 Development of utility system simulation model Mason, Edward A. (Edward Archibald), 1924-; Deaton Paul Ferris; Kearney Joseph P. (Joseph Patrick) "Worked preformed for Commonwealth Edison Company, Chicago, Illinois."; Includes bibliographical references (leaf 28) 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89615 Reactor physics project progress report Driscoll, Michael J.; Thompson, Theos Jardin, 1918-1970; Clikeman, Franklyn Miles; Donohew, Jack Norman; Eckard, Joseph D., 1942-; Harper, Thomas Lawrence; Hukai, Yoshiyuti; Kaplan, Irving, 1912-; Kim, Chang Hyo; LeFevre, Yves-Marie Etienne André; Leung, Timothy Chung-tim; Ortiz, Nestor Ruben; Rasmussen, Norman C.; Rim, Chang Saeng; Seth, Shivaji Shrilal; Supple, A. T. Massachusetts Institute of Technology; Takahata, Chuzo Statement of responsibility on title page reads: Editors: M.J. Driscoll and T.J. Thompson; Contributors: F.M. Clikeman, J.N. Donohew, M.J. Driscoll, J.D. Eckard, T.L. Harper, Y. Hukai, I. Kaplan, C.H. Kim, Y.-M. Lefevre, T.C. Leung, N.R. Ortiz, N.C. Rasmussen, C.S. Rim, S.S. Seth, A.T. Supple C. Takahata, and T.J. Thompson; "MIT-3944-1."; Progress report; September 30, 1968 1968-01-01T00:00:00Z https://hdl.handle.net/1721.1/89614 The effect of uranium-236 and neptunium-237 on the value of uranium as feed for pressurized water power reactors Goellner, Donald A., 1936-; Benedict, Manson; Mason, Edward A. (Edward Archibald), 1924- "December 1967."; Volume 2 contains appendices; MIT-2073-6; Includes bibliographical references 1967-01-01T00:00:00Z https://hdl.handle.net/1721.1/89613 Analytical and experimental investigations of the behavior of thermal neutrons in lattices of uranium metal rods in heavy water Simms, Richard; Kaplan, Irving, 1912-; Thompson, Theos Jardin, 1918-1970; Lanning, David D. Measurements of the intracellular distribution of the activation of foils by neutrons were made in lattices of 1/4-inch diameter, 1.03% U-235, uranium rods moderated by heavy water, with bare and cadmium-covered foils of gold, depleted uranium, lutetium, europium and copper. The measurements were made in the M.I.T. Heavy Water Lattice Facility with source neutrons from the M.I.T. Reactor. Two lattices were studied in detail in this work. The more closely packed lattice had a triangular spacing of 1.25 inches, and the less closely packed lattice had a triangular spacing of 2.5 inches. The results of the experiments were compared to one-dimensional, 30-energy group, THERMOS calculations based on the available energy exchange kernels. The comparison indicated that the approximation that the hexagonal cell may be replaced by an equivalent circular cell (the Wigner-Seitz approximation) can lead to serious discrepancies in closely packed lattices moderated b! y heavy water.; A modified one-dimensional, and a two-dimensional, calculation were shown to predict the intracellular activation distribution in the closely packed lattice. An analytical treatment of the problem of the flux perturbation in a foil was developed and compared to the experimental results obtained by using gold foils of four different thicknesses in the lattice cell; the method was shown to be adequate. An analytical method to treat the effect of leakage from an exponential assembly was formulated; the results indicated that only in small exponential assemblies would leakage be a significant problem in intracellular flux measurements. A method was developed to predict the cadmium ratio of the foils used in the lattice cell; comparison with available measurements with gold foils indicated good agreement between theory and experiment, except for a lattice having very large ratios of moderator volume, to fuel volume, e.g., 100:1.; Calculations of the fuel disadvantage factor by the method of successive generations for gold, lutetium and europium detector foils were compared to the results of THERMOS calculations, because THERMOS was shown to predict the experimental distributions. The comparison indicated that the method of successive generations is a good alternative to the THERMOS calculation, if all that is required is 17 and the thermal utilization. Statement of responsibility on title-page reads: R. Simms, I. Kaplan, T. J. Thompson, D. D. Lanning; "October 11, 1963."; "NYO-10211."; Also issued by the first author as a Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1964; Includes bibliographical references (leaves 193-199) 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89612 Spatial distribution of the neutron flux on the surface of a graphite-lined cavity Madell, J. T.; Thompson, Theos Jardin, 1918-1970; Profio, A. Edward, 1931-; Kaplan, Irving, 1912- Statement of responsibility on title-page reads: J.T. Madell, T.J. Thompson, A.E. Profio, and I. Kaplan; "April 1, 1962."; "NYO-9657."; Includes bibliographical references (leaves 315-316) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89611 Use of neutron absorbers for the experimental determination of lattice parameters in subcritical assemblies Harrington, J. (John); Lanning, David D.; Kaplan, Irving, 1912-; Thompson, Theos Jardin, 1918-1970 Statement of responsibility on title-page reads: J. Harrington, D. D. Lanning, I. Kaplan, T. J. Thompson; "February 1966."; AEC Research and Development Report; MIT-2344-6; Includes bibliographical references (leaves 188-192) 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89610 In-pile loop irradiation studies of organic coolant materials : quarterly progress report, July 1-September 30, 1963 Mason, Edward A. (Edward Archibald), 1924-; Bley, W. N. (William Norman); Sawyer, Craig D. (Craig Delany); Swan, Arthur Henry; Chin, Roy A.; Casey, J. P. Massachusetts Institute of Technology; Terrien, Jean-François Emile Marie; Nullens, Gilles C. H. Statement of responsibility on title page reads: Report prepared by: E. A. Mason, Project Supervisor W. N. Bley, Project Engineer; Contributors: C. D. Sawyer A. H. Swan, R. A. Chin, J. P. Casey J. F. Terrien G. C. Nullens; "Issued: December 15, 1963."; "AEC Research and Development Report"--Cover; "SRO-85."; Includes bibliographical references (leaf 24); Quarterly progress report; July 1-September 30, 1963 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89609 Organic moderator-coolant in-pile irradiation loop for the MIT nuclear reactor Mason, Edward A. (Edward Archibald), 1924-; Bley W. N. (William Norman); Morgan Dean T. Includes bibliographical references (leaf 34); Progress report; to July 1, 1961 1961-01-01T00:00:00Z https://hdl.handle.net/1721.1/89608 Heavy Water Lattice Project final report Thompson, Theos Jardin, 1918-1970; Kaplan, Irving, 1912-; Driscoll, Michael J. MIT-2344-12; AEC Research and Development Report; Bibliography: leaves 181-192 1967-01-01T00:00:00Z https://hdl.handle.net/1721.1/89607 Studies of reactivity and related parameters in slightly enriched uranium, heavy water lattices Malaviya, Bimal Kumar; Kaplan, Irving, 1912-; Lanning, David D.; Profio, A. Edward, 1931-; Thompson, Theos Jardin, 1918-1970 Statement of responsibility as it appears on title page reads: B. K. Malaviya, I. Kaplan, D. D. Lanning, A. E. Profio , T. J. Thompson; "May 25, 1964."; MIT-2344-1; Includes bibliographical references 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89606 Prompt activation analysis of coal and iron ore Schaefer, Robert W. (Robert Walter); Rasmussen, Norman C. Prepared for U.S. Dept. of Mines, Bureau of Mines; Also issued as a M.S. thesis in the Dept. of Nuclear Engineering, 1970; Includes bibliographical references (leaf 47) 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89605 Final report, U.S. Bureau of Mines Contract number H0180895 Rasmussen, Norman C. Includes bibliographical references (leaf 10); Final report 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89604 Optimization of material distributions in fast breeder reactors Tzanos, C. P.; Gyftopoulos E. P.; Driscoll Michael J. "MIT-4105-6."; Based on a Sc. D. thesis submitted by C.P. Tzanos to the Dept. of Nuclear Engineering, 1971; Includes bibliographical references (pages 188-190) 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89603 The decontamination of radioactive ion exchange resins using neutral salts as elutriants Baron, Joseph S. (Joseph Sigmund); Mason Edward A. (Edward Archibald) 1924-; Olson N. Thomas Includes bibliographical references (leaves 124-126) 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89602 Finite element methods for space-time reactor analysis Kang, Chang Mu; Hansen Kent F. "MIT-39-3-5."; Also issued as a Sc. D. thesis by Chang Mu Kang in the Dept. of Nuclear Engineering, 1971; Includes bibliographical references (leaves 147-150) 1971-01-01T00:00:00Z https://hdl.handle.net/1721.1/89601 Pressurized water reactor loss-of-coolant accidents by hypothetical vessel rupture Doan, Phung Lien; Lanning, David D.; Rasmussen, Norman C. Also issued by the 1st author as an Sc. D. thesis, Massachusetts Institute of Technology. Dept. of Nuclear Engineering, 1972; Includes bibliographical references (leaves 331-349) 1972-01-01T00:00:00Z https://hdl.handle.net/1721.1/89600 Simulation and optimization techniques of nuclear in-core fuel management decisions Kearney, Joseph P. (Joseph Patrick); Mason Edward A. (Edward Archibald) 1924- Also issued as a Ph. D. thesis in the Dept. of Nuclear Engineering, MIT, 1973; Includes bibliographical references (leves 290-293) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89599 GAKIN II : a one-dimensional multigroup diffusion theory reactor kinetics code Mason, J. H. (John Herbert); Hansen, Kent F. Cover and added title-page read: By K.F. Hansen, J.H. Mason; Also issued as a M.S. thesis in the Dept. of Nuclear Engineering, MIT, 1973; Includes bibliographical references (leaf 204) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89598 Design of a cold neutron source for the MIT reactor Sanders, Robert Charles; Lanning David D.; Thompson Theos Jardin 1918-1970 Also issued as a Sc. D. thesis in the Dept. of Nuclear Engineering, MIT, 1970; Includes bibliographical references (leaves 211-212) 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89597 Finite difference techniques for the solution of the reactor kinetics equations Reed, William Hudmon; Hansen Kent F. Also issued as a Sc. D. thesis in the Dept. of Nuclear Engineering, MIT,1969; "MIT-3903-2."; Bibliography: leaves 70-71 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89596 Numerical solution of the two-dimensional time-dependent multigroup equations McCormick, W. T. (William Thomas); Hansen, Kent F. Also issued as a Ph. D. thesis in the Dept. of Nuclear Engineering, MIT, 1969; "MIT-3903-1."; Includes bibliographical references (leaves 60-61) 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89595 Neutron and gamma-ray spectroscopy and activation analysis : final report Rasmussen, Norman C.; Thompson, Theos Jardin, 1918-1970 "AFCRL-66-135."; Includes bibliographical references (pages 167-171); Final report 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89594 Measurements of reactor parameters in subcritical and critical assemblies : a review Kaplan, Irving, 1912- "NYO 10, 207."; Includes bibliographical references (pages 50-58) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89593 Summary description of the fuel depletion code, CELL Goellner, D.; Beaudreau, James Joseph "MIT-2073-8."; Includes bibliographical references (pages 85-66) 1968-01-01T00:00:00Z https://hdl.handle.net/1721.1/89592 The nature of the high boiler degradation products from irradiated Santowax OMP Bley, W. N. (William Norman); Mason, Edward A. (Edward Archibald), 1924- Based on the 1st author's M.S. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1965; "MIT-334-11."; Includes bibliographical references (pages 89-92) 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/89591 Effect of reactor irradiation on Santowax WR : irradiations from 425° F to 800° F at 40% fast neutron fraction Timmins, Thomas Howard; Mason, Edward A. (Edward Archibald), 1924-; Morgan, Dean Thomas "MIT-334-34."; Includes bibliographical references 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89590 Relative roles of pyrolysis and radiolysis in the degradation of terphenyls Terrien, Jean-François Emile Marie; Mason, Edward A. (Edward Archibald), 1924- "SRO-87."; Also issued as an M.S. thesis by the first author and supervised by the second author, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1964 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89589 Prompt activation analysis for boron and lithium Clark, Lincoln; Rasmussen, Norman C. "AFCRL-63-575."; Includes bibliographical references (pages 91-92) 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89588 PULSE : an IBM 7094 program for calculation of fast neutron kinetics by Monte Carlo. Addendum no. 1, May 1964 Profio, A. Edward, 1931- Addendum no. 1; May 1964 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89587 Measurement of gamma-ray spectra from thermal-neutron capture Neill, John Muir; Rasmussen, Norman C.; Thompson, Theos Jardin, 1918-1970 "AFCRL-63-341."; Originally presented as the first author's Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1963 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89586 The effects of changing economic conditions on energy costs in zircaloy 4 clad pressurized water reactors : a report to East Central Nuclear Group Richardson, Max C. Includes bibliographical references (page 78) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89585 Recalculation of power costs for the CANDU reactor Hsi, Ching-Foo George; Benedict, Mason "NYO-9716." 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89584 A study of the fast fission effect in lattices of uranium rods in heavy water Wolberg, John R.; Thompson, Theos Jardin, 1918-1970; Kaplan, Irving, 1912- NYO-9661; Includes bibliographical references (p. 168-171) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89583 Theory and use of small subcritical assemblies for the measurement of reactor parameters Peak, John Carl; Kaplan Irving 1912-; Thompson Theos Jardin 1918-1970 "NYO-10204 ."; "TID-4500, 17 ed. UC-34 Physics."; Includes bibliographical references (pages 124-126) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89582 Dynamics and control of nuclear rocket engines Smith, Harold Palmer; Stenning, Alan Hugh Errata of 2 p. inserted; Originally issued as the 1st author's Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1960; Includes bibliographical references (pages 181-182); Final report 1960-01-01T00:00:00Z https://hdl.handle.net/1721.1/89581 BWR depletion methods using quandry and the response matrix technique of assembly homogenization Khalil, Hussein Shoukry Cover title; Includes bibliographical references (leaves 52-53) 1981-01-01T00:00:00Z https://hdl.handle.net/1721.1/89580 Application of time dependent unavailability analysis to standby safety systems Dykes, Andrew Arthur; Rasmussen Norman C.; Vesely W. E. "Prepared for Brookhaven National Laboratory."; Includes bibliographical references (p. 280-284) 1982-01-01T00:00:00Z https://hdl.handle.net/1721.1/89579 Nuclear power plant design innovation for the 1990s : a preliminary assessment Lester, Richard K. (Richard Keith), 1954-; Driscoll, Michael J.; Golay, M.; Lanning, David D.; Lidsky, L. M. (Lawrence Mark); Rasmussen, Norman C.; Todreas, Neil E. Statement of responsibility on title-page reads: Richard K. Lester, Michael J. Driscoll, Michael W. Golay, David D. Lanning, Lawrence M. Lidsky, Norman C. Rasmussen and Neil E. Todreas; "September 1983."; Includes bibliographical references 1983-01-01T00:00:00Z https://hdl.handle.net/1721.1/89578 Impact of thermal constraints on the optimal design of high-level waste repositories in geologic media : topical report Malbrain, C. (Carl); Lester, Richard K. (Richard Keith), 1954- "Prepared for the U.S. Department of Energy."; Includes bibliographical references (p. 49) 1982-01-01T00:00:00Z https://hdl.handle.net/1721.1/89577 Analysis of forced convection degraded core cooling in light water reactors Mohammed, S. M.; Kazimi Mujid S. Includes bibliographical references (pages 135-141) 1981-01-01T00:00:00Z https://hdl.handle.net/1721.1/89576 Station blackout : an opportunity for formulating less prescriptive nuclear safety regulation Manno, V. Includes bibliographical references 1984-01-01T00:00:00Z https://hdl.handle.net/1721.1/89575 Analytical investigation of post-accident containment atmospheric stratification Manno, V.; Golay, M. Bibliography: leaf 21 1984-01-01T00:00:00Z https://hdl.handle.net/1721.1/89574 The use of burnable poison to improve uranium utilization in PWRs Loh, Wee Tee; Driscoll Michael J.; Lanning David D. A methodology based on the linear reactivity model of core behavior has been developed and employed to evaluate fuel management tactics for improving uranium utilization in Pressurized Water Reactors in a once-through fuel cycle mode on a consistent basis. A major focus has been on the benefit of using burnable poison in conjunction with low-leakage fuel management schemes. Key features in the methodology, such as power weighting of batch reactivity values and correlation of neutron leakage effects with peripheral assembly power, were verified against results generated using detailed state-of the- art computer analyses. A relation between batch power fraction and batch reactivity was derived from a 1/2-group diffusion theory model, and similarly validated. These prescriptions have been used in two ways: to develop analytical models which allow quick scoping calculations; and, programmed into a code, to facilitate more rigorous applications. The methodo! logy has been applied to evaluate fuel management schemes of contemporary interest, such as the use of burnable poison to shape the power history profile, the use of low-leakage fuel loading patterns, and extended cycle length/ burnup, and combinations of these individual schemes. It was found that shaping of the power,history profile in a low-leakage assembly pattern by means of burnable poison, even after accounting for the anticipated residual poison reactivity penalty, has the potential of increasing PWR discharge burnup, and hence uranium utilization by roughly 1%. The overall improvement in uranium utilization for a low-leakage loading over that for the current out-in/scatter scheme, was about 3.6% for current cycle lengths (3-batch, discharge burnup ' 30,000 MWD/MT), and approximately 11.1% for extended cycle operation (3-batch, discharge burnup u 50,000 MWD/MT). Also presented as author's dissertation in substantially the same form. (Nuc. Eng., Ph. D., 1982); Includes bibliographical references (pages 246-249) 1982-01-01T00:00:00Z https://hdl.handle.net/1721.1/89573 Uranium from seawater research : final progress report, FY 1982 Borzekowski, J.; Driscoll Michael J.; Best F. R. During the FY '82 campaign 14 new ion exchange resin formulations, prepared by the Rohm & Haas Company, were tested by MIT at the Woods Hole Oceanographic Institution. The best of these chelating resins was again of the acrylic amidoxime type; it picked up approximately 100 ppm uranium in seven days' exposure to seawater, which represents a factor of better than two improvement over thq seven-day results for the best FY '81 candidate (which saturated at roughly 100 ppm U after 30 days' exposure). Saturation was not reached and, within experimental accuracy, uranium accumulated at a constant rate over the seven-day period; it is speculated that a useful capacity of over 300 ppm U would be achieved. All resins of the styrenic amidoxime type were found to be an order of magnitude lower in their effective capacity for uranium in seawater than the best of the acrylic forms. Particle size effects, which were found to be less than expe! cted from theoretical computations of both fluid and solid side mass transfer resistance, can not account for this difference. Scanning electron microscope examination by R & H scientists of ion exchange resin beads from beds subjected to seawater flow for 30 days in MIT's WHOI columns showed that the internal pores of the macro reticular- type resins become filled with debris (of undetermined nature and effect) during exposure. Includes bibliographical references; Final progress report; FY 1982 1982-01-01T00:00:00Z https://hdl.handle.net/1721.1/89572 Final report on improved uranium utilization in PWRs Driscoll, Michael J. This is the final summary progress report on a research program carried out within the MIT Energy Laboratory/Nuclear Engineering Department under the US Department of Energy's program to increase the effectiveness of uranium utilization in light water reactors on the once-through fuel cycle. Two major themes, methodology and applications, characterize the research. A simple built accurate set of algorithms, designated as "the linear reactivity method" were developed to permit self-consistent evaluations of a broad spectrum of changes in core design and fuel management tactics. More than a dozen suggested improvements were then evaluated, focusing primarily on retrofittable modifications and pressurized water reactors. In common with the findings of many other investigators, high burnup and routine end-of-cycle coastdown were identified as preferred options. Includes bibliographical references 1982-01-01T00:00:00Z https://hdl.handle.net/1721.1/89571 The selective use of thorium and heterogeneity in uranium-efficient pressurized water reactors Kamal, Altamash; Driscoll Michael J.; Lanning David D. Systematic procedures have been developed and applied to assess the uranium utilization potential of a broad range of options involving the selective use of thorium in Pressurized Water Reactors (PWRs) operating on the once-through cycle. The methods used rely on state-of-the-art physics methods coupled with batch-wise core depletion models based on the "group-and-one-half" theory. The possible roles for thorium that were investigated are: as internal and radial blanket material, as thorium pins dispersed within uranium fuel assemblies, its use in PWRs operating on spectral shift control, and its reconstitution and reinsertion as radial blanket assemblies. The use of smaller assemblies in PWRs (for cores with and without thorium) was also investigated, as well as options which can be regarded as reasonable substitutes for employing thorium. The analyses were performed for both current (3-batch, discharge burnup n 30 GWD/MT) and high-burnup (5! -batch, discharge burnup% 50 GWD/MT) PWR cores in their steady-state. It was found that except for special circumstances (dry lattices and/or high burnup), the use of thorium does not save uranium compared to the conventional all-uranium PWRs. When savings are achieved (typically 1-3%, but as high as 9% in special circumstances), they can be, for the most part, equaled or exceeded by easier means: in particular, by the re-use of spent fuel. On the other hand, up to 15 or 20% thorium could be added into PWRs without significant losses in uranium utilization, if policies called for the build up of a U-233 inventory for later use in the recycle mode. It was also found that, regardless of the deployment of thorium, the use of smaller fuel assemblies with the concurrent deployment of radial blankets is an effective uranium conservation strategy, with accompanying power-shaping advantages. Includes bibliographical references (pages 315-318) 1982-01-01T00:00:00Z https://hdl.handle.net/1721.1/89570 Delayed neutron assay to test sorbers for uranium-from-seawater applications Nitta, Cynthia K.; Best F. R.; Driscoll Michael J. Includes bibliographical references (pages 106-107); Final Report of the Uranium from Seawater Project ; FY 1981 1982-01-01T00:00:00Z https://hdl.handle.net/1721.1/89569 Global energy futures and CO₂-induced climate change Rose, David J.; Miller, Marvin M.; Agnew, Carson E. "15 November 1983."; Also issued as a 3 volume set; "Report prepared for Division of Policy Research and Analysis, National Science Foundation."; Includes bibliographical references (pages 224-234) 1983-01-01T00:00:00Z https://hdl.handle.net/1721.1/89568 Measurement of the ratio of fissions in U²³⁸ to fissions in U²³⁵ using 1.60 mev gamma rays of the fission product La¹⁴⁰ Wolberg, John R.; Thompson Theos Jardin 1918-1970; Kaplan Irving 1912- "August 19, 1963."; "NYO-10210. " 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89566 The effects of changing economic conditions on fuel cycle costs in pressurized water reactors : a report to East Central Nuclear Group, Inc. Benedict, Manson; Fenech, Henri, 1925-; Richardson, Max C. 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89565 Deuterium concentration by chemically-refluxed ammonia-hydrogen exchange : Final report Mason, Edward A. (Edward Archibald), 1924-; Benedict, Manson; Chow, Edward Raymond; Baron, Joseph S. (Joseph Sigmund) Statement of responsibility on title-page reads E.A. Mason, M. Benedict, E.R. Chow, J.S. Baron; "June 1969."; "MIT-D14."; Includes bibliographical references (leaves 89-92); Final report; June 1969 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89564 Deuterium concentration by chemically-refluxed ammonia-hydrogen exchange : Supplementary reports Mason, Edward A. (Edward Archibald), 1924-; Benedict, Manson; Chow, Edward Raymond; Baron, Joseph S. (Joseph Sigmund) Statement of responsibility on title-page reads M. Benedict, E.A. Mason, E.R. Chow, J.S. Baron; "June 1969."; "MIT-D15."; Includes bibliographical references 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89563 A method of short-range system analysis for electric utilities containing nuclear plants Eng, Raymond Lehman Also issued as a Ph. D. thesis in the Dept. of Nuclear Engineering, M.I.T., 1975; Includes bibliographical references (pages 460-462) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89562 Design, construction and evaluation of a facility for the simulation of fast reactor blankets Forbes, Ian Alexander; Driscoll, Michael J.; Thompson, Theos Jardin, 1918-1970; Kaplan, Irving, 1912-; Manning, David D. A facility has been designed and constructed at the MIT Reactor for the experimental investigation of typical LMFBR breeding blankets. A large converter assembly, consisting of a 20-cm-thick layer of graphite followed by a 17.5-cm-thick U0 2 fuel region, is used to convert thermal neutrons into fast neutrons to drive a blanket mockup. Operating at 55 watts, the converter generates blanket fluxes at an equivalent LMFBR core power of about 350 watts, with as little as one tenth of the blanket material required for a critical assembly. Calculations show that the converter leakage spectrum is a close approximation to the core leakage spectrum from reference LMFBR designs, and that the axial distribution of the neutron flux in the blanket assembly simulates that in the radial blanket of a large LMFBR when the effective height and width of the blanket assembly are correctly chosen. Testing of the completed facility with a blanket composed of 50 v/o iron and ! 50 v/o borax showed that the lateral flux distributions were cosine-shaped, and that lateral spectral equilibrium was achieved in a large central volume of the blanket. Backscattering from concrete shielding surrounding the experiment was found to affect no more than the outer 30 cm of the blanket assembly, confirming the results of two-dimensional multigroup calculations. Measurements of the axial activity of gold and indium show good agreement with 16- group, S8 ANISN calculations. "February 1970."; Statement of responsibility on title-page reads: I.A. Forbes, M.J. Driscoll, T.J. Thompson, I.Kaplan and D.D. Lanning; Also issued as a Ph. D. thesis in the Dept. of Nuclear Engineering, MIT,1970; "MIT-4105-2."; Bibliography: leaves 123-125 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89561 Nondestructive analyses of irradiated MITR fuel by gamma ray spectroscopy Sovka, Jerry Alois; Rasmussen Norman C. "AFCRL-65-787."; Also issued as a Sc. D. thesis in the Dept. of Nuclear Engineering, 1966; Prepared by Air Force Cambridge Research Laboratories; Includes bibliographical references (pages 243-248) 1965-01-01T00:00:00Z https://hdl.handle.net/1721.1/89560 Research and educational activities at the M.I.T. Research Reactor, fiscal years 1969 and 1970 Publications: pages 155-173; Includes bibliographical references (pages 189-191) 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89559 Studies of two-region subcritical uranium heavy water lattices Gosnell, James Waterbury; Driscoll Michael A.; Thompson Theos Jardin 1918-1970 Reactor physics parameters were measured in eleven two-region subcritical assemblies moderated by heavy water. The regions of the assemblies consisted of nine different lattices of various fuel rod size, U235 enrichment, and spacing. The following parameters were measured in the assemblies: bare and cadmium-covered gold foil radial traverses; bare gold foil axial traverses; the ratio of epicadmium to subcadmium capture r tes in U238 (P28); and the ratio of fissions in U238 to fissions in U233(628)- From analysis of axial traverses at various radial positions, it was determined that the axial buckling was independent of radial position in the assemblies. A method was developed to apply the age equation to the experimental gold foil traverses. This analysis yielded the quantity [ ... ] for each region of the assemblies. Calculated values of [ ... ] were used to obtain values of the infinite multiplication factor from this parameter.; For assemblies of sufficiently large inner regions, the values of km so found agreed within experimental uncertainty with independent determinations. The slowing-down spectra. arising from the age theory analysis were used to extrapolate two-region assembly measurements of P28 to critical assembly values. General agreement was found between these extrapolated values and the results of measurements made in full, single region lattices. The heterogeneous expressions for uncollided flux derived by Pilat were extended to two-region assemblies and used to determine single rod values of 628 from two-region assembly measurements. The theory was also used to predict values of 628 for each of the lattices composing the two-region assemblies. Both the single rod values and the full lattice predictions agreed within experimental error with previously reported results. The determination of material buckling from two-region subcritical assemblies is also discussed.!; Because of the nature of the assemblies investigated, satisfactory measurements of the material buckling could not be made. "February 1969."; "MIT-2344-13."; Some technical reports have the series numbering of MITNE-84; Substantially the same as J.W. Gosnell's Ph. D. thesis in the Dept. of Nuclear Engineering, 1969; Includes bibliographical references (leaves [212]-215) 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89558 Irradiation loop studies of organic reactor coolants : Progress report Mason, Edward A. (Edward Archibald), 1924- Prepared by E.A. Mason, Project supervisor; Progress report; Jan.-Sept. 1962 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89557 The effect of refueling decisions and engineering constraints on the fuel management for a pressurized water reactor Rieck, Terrance Arthur; Benedict, Manson; Mason, Edward A. (Edward Archibald), 1924- Also issued as a Ph. D. thesis in the Dept. of Nuclear Engineering, 1974; Work sponsored by Commonwealth Edison Company; Includes bibliographical references (pages 383-385) 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89556 Modifications to fuel cycle code "FUELMOVE" Sovka, Jerry Alois; Benedict Manson "NYO-9717."; Includes bibliographical references (leaf 37) 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89555 Summary report on heavy water, natural uranium research Kaplan, Irving, 1912-; Lanning David D.; Profio A. Edward 1931-; Thompson Theos Jardin 1918-1970 NY0-10209; Summary report 1963-01-01T00:00:00Z https://hdl.handle.net/1721.1/89554 The Effect of uranium-236 and neptunium-237 on the value of uranium used as feed for thermal power reactors Benedict, Manson; Bauhs, David John; Golden, Terence Cashin; Mason, Edward A. (Edward Archibald), 1924- Statement of responsibility on title page reads: Manson Benedict, David J. Bauhs, Terence C. Golden and Edward A. Mason; "June 30, 1968."; "MIT-2073-7."; Includes bibliographical references (pages 183-184) 1968-01-01T00:00:00Z https://hdl.handle.net/1721.1/89553 A Single-element method for heterogeneous nuclear reactors Seth, Shivaji Shrilal; Driscoll, Michael J.; Kaplan, Irving, 1912-; Thompson, Theos Jardin, 1918-1970; Lanning, David D. Statement of responsibility on title-page reads, S.S. Seth, M.J. Driscoll, I. Kaplan, T.J. Thompson and D.D. Lanning; "May 1970."; "MIT-3944-3."; Also issued by the first author and supervised by the second and third author as a Sc. D. thesis in the Dept. of Nuclear Engineering, MIT, 1970; Includes bibliographical references (leaves 169-176) 1970-01-01T00:00:00Z https://hdl.handle.net/1721.1/89552 Neutron dosimetry, spectrometry, and neutron activation analysis : final report Thompson, Theos Jardin, 1918-1970; Rasmussen Norman C.; Schwartz Daniel AFCRL-62-1; Includes bibliographical references (leaf 25); Final report 1961-01-01T00:00:00Z https://hdl.handle.net/1721.1/89551 Fast neutron shielding studies Wilensky, Samuel; Beghian Leon Edward; Clikemann Franklin Includes bibliographical references (leaf 6) 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89550 Measurements of neutron capture in U²³⁸ in lattices of uranium rods in heavy water Weitzberg, Abraham; Kaplan Irving 1912-; Thompson Theos Jardin 1918-1970 "NYO-9659."; Also issued as a Ph. D. thesis in the Dept. of Nuclear Engineering, 1962; Includes bibliographical references (pages 119-145) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89549 Neutron capture gamma rays of 75 elements listed in terms of increasing gamma-ray energy Hamawi, John Nicholas Prepared for U.S. Dept. of the Interior Bureau of Mines; Includes bibliographical references (pages 4-5) 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89548 Studies of epithermal neutrons in uranium, heavy water lattices D'Ardenne, Walter Herbert; Thompson, Theos Jardin, 1918-1970; Lanning, David D.; Kaplan, Irving, 1912- Measurements related to reactor physics parameters were made in three heavy water lattices. The three lattices studied consisted of 0.250-inch-diameter, 1.03 w/o U2 3 5 uranium fuel rods arranged in triangular arrays and spaced at 1.25, 1.75, and 2.50 inches. The following quantities were measured in each of the three lattices studied: the ratio of the average epicadmium U2 3 8 capture rate in the fuel rod to the average subcadmium U2 3 8 capture rate in the fuel rod ([sigma]28); the ratio of the average epicadmium U2 3 o fission rate in the fuel rod7 to the average subcadmium U 35 fission rate in the fuel rod (625); the ratio of the average U2 3 8 capture rate in the fuel rod to the average U2 3 5 fission rate in the fuel rod (C ); the ratio of the average U2 3 8 fission rate in the fuel rod to the average U2 3 5 fission rate in the fuel rod (628); and the effective resonance integral of U2 3 8 in a fuel rod (ER12 8 ).; The results of an investigation of systematic errors associated with these measurements have-led to many changes and adjustments in the experimental techniques and procedure which have improved the general precision of the experimental results. A new method was developed to measure the ratio C * which simplified the experiment, significantly reduced the experimental uncertainty associated with the measurement, and avoided systematic errors inherent in the method used to measure C* in earlier work. The value of ER12 8 was also measured by a new method in which the results of measurements made in an epithermal flux which had a 1/E energy dependence are combined with the results of measurements made in a lattice.; The experimental results were combined with theoretical results obtained from the computer programs THERMOS and GAM-I to determine the following reactor physics parameters for each of the three lattices studied: the resonance escape probability, p; the fast fission factor, E; the multiplication factor for an infinite system, k [infinity]; and the initial conversion ratio, C. Methods were developed to measure that portion of the activity of a foil which is due to neutron captures in the resonances in the activation cross section of the foil material. The resonance escape probability was determined by a new method, using the resonance activation date, in which the use of cadmium is not necessary. Statement of responsibility on title-page reads: W. H. D'Ardenne, T. J. Thompson, D. D. Lanning and I. Kaplan; "August 24, 1964."; "MIT-2344-2."; Also issued as a Ph. D. thesis by the first author, MIT Dept. of Nuclear Engineering, 1964; Includes bibliographical references (leaves 167-170) 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89547 Multi-group diffusion methods Hansen, Kent F. "NYO-10, 206."; Includes bibliographical references (page 71) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89546 Fuel cycle analysis in a thorium fueled reactor using bidirectional fuel movement Hofmann, Ferdinand "MIT-2073-1."; Includes bibliographical references (pages 131-132) 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89545 Process system requirements of the MIT reactor at five megawatts Devoto, William Robert Also issued as a Nuclear Engineer thesis in the Dept. of Nuclear Engineering, 1962; Includes bibliographical references (leaves 168-169) 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89544 Rate of exchange of deuterium between water and dissolved hydrogen in presence of diethylamine Ishida, Takanobu; Benedict Manson Originally issued as the 1st author's Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1964; "MIT-2249-1." 1964-01-01T00:00:00Z https://hdl.handle.net/1721.1/89543 The measurement of reactor parameters in slightly enriched uranium, heavy water moderated miniature lattices Sefchovich-Itzcovich, Elias; Kaplan, Irving, 1912-; Thompson, Theos Jardin, 1918-1970 "MIT-2344-8." 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89542 Use of a pulsed neutron source to determine nuclear parameters of lattices of partially enriched uranium rods in heavy water Bliss, Henry Edison; Kaplan, Irving, 1912-; Thompson, Theos Jardin, 1918-1970 Also issued as an Sc. D. thesis, Massachusetts Institute of Technology. Dept. of Nuclear Engineering, 1967; "MIT-2344-7." 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89541 Fuel cycles in nuclear reactors Shanstrom, R. T.; Benedict Manson; McDaniel C. T. Series numbering from publisher's list; "61"--stamped on cover; "Unclassified. NYO-2131."; Originally issued by the first author as an Sc. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1959 1959-01-01T00:00:00Z https://hdl.handle.net/1721.1/89540 Reliability analysis of complex technical systems using the fault tree modularization technique Modarres, M. (Mohammad); Rasmussen Norman C.; Wolf Lothar Originally presented as the first author's thesis, (Ph. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1980; Includes bibliographical references 1980-01-01T00:00:00Z https://hdl.handle.net/1721.1/89539 Optimization of the axial power shape in pressurized water reactors Malik, Mushtaq Ahmad; Kamal, Altamash; Driscoll, Michael J.; Lanning, David D. Analytical and numerical methods have been applied to find the optimum axial power profile in a PWR with respect to uranium utilization. The preferred shape was found to have a large central region of uniform power density, with a roughly cosinusoidal.profile near the ends of the assembly. Reactivity and fissile enrichment distributions which yield the optimum profile were determined, and a 3-region design was developed which gives essentially the same power profile as the continuously varying optimum composition. State of the art computational methods, LEOPARD and PDQ-7, were used to evaluate the beginning-of-life and burnup history behavior of a series of three-zone assembly designs, all of which had a large central zone followed by a shorter region of higher enrichment, and with a still thinner blanket of depleted uranium fuel pellets at the outer periphery. It was found that if annular fuel pellets were used in the higher enrichment zone, a design ! was created which not only had the best uranium savings (2.8% more energy from the same amount of natural. uranium, compared to a conventional, uniform, unblanketed design), but also had a power shape with a lower peak-to-average power ratio (by 16.5%) than the reference case, and which held its power shape very nearly constant over life. This contrasted with the designs without part length annular fuel, which tended to burn into an end-peaked power distribution, and with blanket-only designs, which had a poorer peak-to-average power ratio than the reference udblanketed case. Statement of responsibility on title-page reads: M.A. Malik, A. Kamal, M.J. Driscoll, and D.D. Lanning; "November 1981."; Originally presented as the first author's M.S. thesis, M.I.T. Dept. of Nuclear Engineering, 1981; Includes bibliographical references (pages 112-114) 1981-01-01T00:00:00Z https://hdl.handle.net/1721.1/89538 Systems studies on the extraction of uranium from seawater Driscoll, Michael J.; Best F. R. This report summarizes the work done at MIT during FY 1981 on the overall system design of a uranium-from-seawater facility. It consists of a sequence of seven major chapters, each of which was originally prepared as a stand-alone internal progress report. These chapters trace the historical progression of the MIT effort, from an early concern with scoping calculations to define the practical boundaries of a design envelope, as constrained by elementary economic and energy balance considerations, through a parallel evaluation of actively-pumped and passive current-driven concepts, and thence to quantification of the features of a second generation system based on a shipboard-mounted, actively-pumped concept designed around the use of thin beds of powdered ion exchange resin supported by cloth fiber cylinders (similar to the baghouse flyash filters used on power station offgas). An assessment of the apparently inherent limitations of even thin settled-b! ed sorber media then led to selection of an expanded bed (in the form of an ion exchange "wool"), which would permit an order of magnitude increase in flow loading, as a desirable advance. Thus the final two chapters evaluate ways in which this approach could be implemented, and the resulting performance levels which could be attained. Overall, U 308 production costs under 200 $/lb appear to be within reach if a high capacity (several thousand ppm U) ion exchange wool can be developed. Includes bibliographical references 1981-01-01T00:00:00Z https://hdl.handle.net/1721.1/89537 Analysis of strategies for improving uranium utilization in pressurized water reactors Sefcik, Joseph A.; Driscoll Michael J.; Lanning David D. Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used ! to verify and supplement these techniques.These methods have been applied to evaluate several suggested improvements: (1) axial blankets of low-enriched or depleted uranium, and of beryllium metal, (2) radial natural uranium blankets, (3) low-leakage radial fuel management, (4) high burnup fuels, (5) optimized H/U atom ratio, (6) annular fuel, and (7) mechanical spectral shift (i.e. variable fuel-to-moderator ratio) concepts such as those involving pin pulling and bundle reconstitution.The potential savings in uranium requirements compared to current practice were found to be as follows: (1) O0-3%, (2) negative, (3) 2-3%; possibly 5%, (4) "15%, (5) 0-2.5%, (6) no inherent advantage, (7) 10%. Total savings should not be assumed to be additive; and thermal/hydraulic or mechanical design restrictions may preclude full realization of some of the potential improvements. Includes bibliographical references (pages 238-241) 1981-01-01T00:00:00Z https://hdl.handle.net/1721.1/89536 Measurements of the material bucklings of lattices of natural uranium rods in D₂O Palmedo, Philip F.; Kaplan Irving 1912-; Thompson Theos Jardin 1918-1970 "NYO-9660."; "AEC Research and development report UC-34 physics (TID-4500, 16th edition)."; Originally issued as the first author's Ph. D. thesis, Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1962 1962-01-01T00:00:00Z https://hdl.handle.net/1721.1/89535 Sensitivity analysis of the reactor safety study Parkinson, W. (William); Rasmussen Norman C.; Hinkle William D. Originally presented as the first author's thesis, (M.S.)--in the M.I.T. Dept. of Nuclear Engineering, 1979; Includes bibliographical references (p. 232-233); Final research project report 1979-01-01T00:00:00Z https://hdl.handle.net/1721.1/89518 The fuel cycle economics of improved uranium utilization in light water reactors Abbaspour Tehrani Fard, Ali; Driscoll Michael J. DOE Contract no. EN-77-S-02-4570; Originally presented as the first author's thesis, (Nucl. E.)--in the M.I.T. Dept. of Nuclear Engineering, 1979; Includes bibliographical references (pages 258-260) 1979-01-01T00:00:00Z https://hdl.handle.net/1721.1/89517 A method for risk analysis of nuclear reactor accidents Maekawa, Mitsuru; Rasmussen Norman C.; Vesely W. E. Originally presented as the first author's thesis, (Ph. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1976; Includes bibliographical references (pages 207-208) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89516 Economic feasibility study of total energy system options for the Massachusetts Institute of Technology Was, Gary S. (Gary Steven), 1953-; Golay M. Includes bibliographical references (leaf 39) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89515 Design and fuel management of PWR cores to optimize the once-through fuel cycle Fujita, Edward Kei; Driscoll Michael J.; Lanning David D. DOE Contract no. EN-77-S-02-4570; Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1978; Includes bibliographical references (pages 238-241) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89514 Spatially dependent velocities for the Gakin II reactor kinetics code Strohmayer, Walter Herbert Includes bibliographical references (leaf 5) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89513 Application of probabilistic consequence analysis to the assessment of potential radiological hazards of fusion reactors Sawdye, Robert William; Kazimi, Mujid S. A methodology has been developed to provide system reliability criteria based on an assessment of the potential radiological hazards associated with a fusion reactor design and on hazard constraints which prevent fusion reactors from being more hazardous than light water reactors. The probabilistic consequence analyses, to determine the results of radioactivity releases, employed the consequence model developed to assess the risks associated with light water reactors for the Reactor Safety Study. The calculational model was modified to handle the isotopes induced in the structural materials of two conceptual Tokamak reactor designs, UWMAK-I and UWMAK-III. Volatile oxidation of the first wall during a lithium fire appears to be a primary means of disrupting induced activity, and the molybdenum alloy, TZM (UWMAK-III), tends to be more susceptible than 316 stainless steel (UWMAK-I) to mobilization by this mechanism. It was determined that the radiological!  hazards associated with induced activity in these reactor designs imply reliability requirements comparable to those estimated for light water reactors. The consequences of estimated maximum possible releases of induced activity, however, are substantially less than the maximum light water reactor accident consequences. "July 1978."; Originally presented as the first author's thesis, (M.S.)--in the M.I.T. Dept. of Nuclear Engineering, 1978; Includes bibliographical references (pages 87-89) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89512 Analysis of design strategies for mitigating the consequences of lithium fire within containment of controlled thermonuclear reactors Dube, Donald A.; Kazimi Mujid S. Originally presented as the first author's thesis, (M.S.)--in the M.I.T. Dept. of Nuclear Engineering, 1978; Includes bibliographical references (pages 117-121) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89511 A Solid breeder tokamak blanket designed for failure mode operation Chen, Franklin Fun Kun; Griffith, P. (Peter); McManamy, Thomas Joseph; Was, Gary S. (Gary Steven), 1953- The objective of this study was to evaluate a new concept for a Tokamak type fusion reactor blanket. The design was based on using a packed bed of lithium aluminate as the breeding material with helium gas cooling. The unique aspect of the design was the assumption that small coolant leaks were inevitable and should not necessitate major maintenance. A modularized design was chosen with cylindrical breeder rods and graphite shim rods. Redundancy was provided by designing the blanket such that if a module failed it could be depressurized and its heat load shared by the neighboring operating modules. The thermal hydraulic analysis evolved analytical and computational methods for determining the temperature profiles of all components and the pumping power requirements. A computer program TRIPORT was developed to evaluate the tritium retention and transport. A one dimensional ANISN code was used to determine the breeding ratio for different configurations.!  The thermal hydraulic, neutronic and mechanical aspects of the Breeder Rod Shim Rod (BRSR) design were combined to determine a design window, that is the allowable range of system parameters. Unfortunately adequate breeding could not be demonstrated so that there was no open window. Basically the low breeding was caused by -he inherently poor breeding potential of LiAlO, combined with the additional structure required for failure mode operation. However, this conclusion is based on a specific design concept (BRSR) and further research in the area may prove more fruitful. Statement of responsibility on title-page reads: Franklin Chen, Peter Griffith, Thomas McManamy, and Gary Was; "May 1977."; "This study is basically an integration and extension of a doctor's thesis by Franklin Chen and a master's thesis by Gary Was."; Includes bibliographical references (leaves 244-248) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89510 Aspects of environmental and safety analysis of fusion reactors Kazimi, Mujid S.; Dube, Donald A.; Green, R. W. Massachusetts Institute of Technology; Lidsky, L. M. (Lawrence Mark); Rasmussen, Norman C.; Sawdye, Robert William; Sefcik, Joseph A. This report summarizes the progress made between October 1976 and September 1977 in studies of some environmental and safety considerations in fusion reactor plants. A methodology to assess the admissible occurrence rate of major accidental releases is outlined. The pathways for tritium releases are defined. Preliminary assessment of the important factors in evaluation of the reactor containment building response to Li-Air fire is presented. Statement of responsibility on title-page reads: M. S. Kazimi, editor, D. A. Dube, R. W. Green, L. M. Lidsky, N. C. Rasmussen, R. W. Sawdye, and J. A. Sefcik; "October 1977."; Includes bibliographical references; Progress report; October 1, 1976 to September 30, 1977 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89509 An improved long range fuel management program Beard, Charles Louis Originally presented as the author's thesis, (M.S.)--in the M.I.T. Dept. of Nuclear Engineering, 1978; Includes bibliographical references (leaf 70) 1978-01-01T00:00:00Z https://hdl.handle.net/1721.1/89508 A modular approach to fault tree and reliability analysis Olmos, Jaime; Wolf Lothar Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977; Includes bibliographical references (p. 311-312) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89507 Spatial homogenization of diffusion theory parameters Worley, Brian Addison; Henry Allan F. Originally presented as the first author's thesis, (Ph. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977; Includes bibliographical references 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89506 Improved multidimensional numerical methods for the steady state and transient thermal-hydraulic analysis of fuel pin bundles and nuclear reactor cores Masterson, Robert Edward; Wolf Lothar Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977; Includes bibliographical references 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89505 Neutronic sensitivity studies on MEKIN, an accident analysis computer code for nuclear reactors Barbehenn, Craig Edwin; Lanning David D. Originally presented as the first author's thesis, (Nucl. E.)--in the M.I.T. Dept. of Nuclear Engineering, 1977; Includes bibliographical references (pages 122-123) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89504 Resonance region neutronics of unit cells in fast and thermal reactors Salehi, Ali Akbar; Driscoll Michael J.; Deutsch Owen Leslie Originally presented as the first author's thesis, (Ph. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977; Includes bibliographical references (p. 226-229) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89503 Evaluation of advanced fast reactor blanket designs Shin, Jae In; Driscoll Michael J. Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977; Includes bibliographical references (p. 321-325) 1977-01-01T00:00:00Z https://hdl.handle.net/1721.1/89502 Mathematical models for predicting the thermal performance of closed-cycle waste dissipation systems Guyer, Eric C.; Golay M. Includes bibliographical references (leaves 42-44) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89501 An engineering and economic evaluation of some mixed-mode waste heat rejection systems Guyer, Eric C.; Golay M. Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977; Includes bibliographical references (p. 336-342) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89500 Multidimensional modeling of the rod drop accident Valente, John Umberto Originally presented as the author's thesis, (Nucl. E.)--in the M.I.T. Dept. of Nuclear Engineering, 1976; Includes bibliographical references (pages 237-239) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89499 Effect of reactor irradiation on Santowax WR : 1. Radiolysis reaction order and fast neutron effect, 2. Radiopyrolysis Mason, Edward A. (Edward Archibald), 1924-; Timmins, Thomas Howard; Morgan, Dean T.; Bley, W. N. (William Norman) Statement of responsibility on title-page reads: E.A. Mason, T.H. Timmins, D.T. Morgan, and W.N. Bley; "Issued: October 1966."; "MIT-334-70 Reactor Technology."; Also issued by T.H. Timmins and supervised by E.A. Mason as an Sc. D. thesis , Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1967; Includes bibliographical references (pages A6.1-A6.7) 1966-01-01T00:00:00Z https://hdl.handle.net/1721.1/89498 Investigation of solution techniques for large sparse band width matrix equations of linear systems Guillebaud, Louis Jean Marie; Golay M. Includes bibliographical references (leaves 107-108) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89497 The Reactor engineering of the MITR-II : construction and startup Allen, G. C.; Clark, Lincoln; Gosnell, James Waterbury; Lanning, David D. The heavy water moderated and cooled research reactor, MITR-I, has been replaced with a light water cooled, heavy water reflected reactor called the MITR-II. The MITR-II is designed to operate at 5 thermal megawatts. The MITR-I was shutdown in May, 1974, dismantling, construction, and preoperational testing continued until the MITR-II went critical on August 14, 1975. Cadmium absorbers were fixed in the upper core of the first fuel loadings to shorten the active core height and provide reactivity control. Solid non-fueled elements were also loaded for additional reactivity control. Swelling of the original cadmium fixed absorbers necessitated a second core configuration. The second core contained additional solid non-fueled elements and no fixed absorbers. The compact core of the MITR-II causes thermal neutron flux and power peaking to occur at the core outer boundaries and incore locations with excess moderator. The active core power density is in the!  range of 100 to 150 watts/cm 3 with peaks up to 300 watts/cm 3 . The power, flow, and temperature distributions of the initial core loadings were determined analytically and experimentally in order to evaluate the safety limit factor and limiting operating conditions. Neutron flux, core temperature, coolant flow, and power distributions were measured by various experimental techniques. The thermal-hydraulic parameters of the initial fuel loadings are evaluated and shown to satisfy the acceptance criteria for operation of the MITR-II. Statement of responsibility on title-page reads: G. C. Allen, Jr., L. Clark, Jr., J. W. Gosnell, and D. D. Lanning; "June, 1976."; Also issued as a Ph. D. thesis by the first author, MIT Dept. of Nuclear Engineering, 1976; Includes bibliographical references (pages 518-521) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89496 The application of alternating-direction implicit methods to the space-dependent kinetics equations Wight, Alan Leonard; Hansen Kent F. Cover title; "MIT-3903-3."; Originally presented as the first author's thesis (Ph. D.), M.I.T. Dept. of Nuclear Engineering, 1969; Includes bibliographical references (leaves 54-55) 1969-01-01T00:00:00Z https://hdl.handle.net/1721.1/89495 The finite element method for neutron diffusion problems in hexagonal geometry Wei, Thomas Ying Chung; Hansen Kent F. Originally presented as the first author's thesis (Ph. D.), M.I.T. Dept. of Nuclear Engineering, 1975; Includes bibliographical references (pages 163-164) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89494 Core design for a small HTGR Ribeiro, Arnaldo Aloisio Telles; Lanning David D. Originally presented as the first author's thesis (Ph. D.), M.I.T. Dept. of Nuclear Engineering, 1976; Includes bibliographical references (leaves 452-458) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89493 The finite element method applied to neutron diffusion problems Deppe, Lothario Olavo; Hansen Kent F. Originally presented as the first author's thesis (Nucl. Eng.)--Massachusetts Institute of Technology Dept. of Nuclear Engineering, 1973; Includes bibliographical references (pages 99-100) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89492 An accident probability analysis and design evaluation of the gas-cooled fast breeder reactor demonstration plant De Laquil, Pascal; Lanning David D.; Rasmussen Norman C. Originally presented as the first author's thesis (Ph. D.), M.I.T. Dept. of Nuclear Engineering, 1976; Includes bibliographical references (pages 510-515) 1976-01-01T00:00:00Z https://hdl.handle.net/1721.1/89491 Finite element synthesis method Yang, Shi-tien; Henry Allan F. Originally presented as the first author's thesis (Ph. D.), M.I.T. Dept. of Nuclear Engineering, 1975; Includes bibliographical references (leaves 151-155) 1975-01-01T00:00:00Z https://hdl.handle.net/1721.1/89490 Assessment of thorium blankets for fast breeder reactors Wood, Paul Joseph; Driscoll Michael J. Also issued as a Sc. D. thesis in the Dept. of Nuclear Engineering, 1973; Includes bibliographical references (pages 538-547) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89489 Transient heat transfer induced pressure fluctuations in the fuel coolant interaction Watson, Charles Edward "31-109-38-2831-2 TR."; Also issued as a M.S. thesis in the Dept. of Nuclear Engineering, MIT, 1973; Includes bibliographical references (leaf 78) 1973-01-01T00:00:00Z https://hdl.handle.net/1721.1/89488 A new approach to solving the multimode kinetics equations Turnage, Joe Clayton; Henry, Allan F. Cover reads: By Joe C. Turner [sic], Allan F. Henry; Also issued as a Ph. D. thesis in the Department of Nuclear Engineering, 1972; Includes bibliographical references (leaves 95-97) 1972-01-01T00:00:00Z https://hdl.handle.net/1721.1/89487 Gamma heating measurements in fast breeder reactor blankets Scheinert, Paul Albert; Driscoll Michael J. Substantially the same as a Nuclear Engineer thesis in the M.I.T. Dept. of Nuclear Engineering, 1974; Includes bibliographical references (pages 247-251) 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89486 Analysis of mixing data relevant to wire wrapped fuel assembly thermal-hydraulic design Ḵẖāṉ, Ahsānullāh; Todreas, Neil E.; Rohsenow, Warren M.; Sonin, A. A. In this report analysis of recent experimental data is presented using the ENERGY code. A comparison of the accuracy of three types of experiments is also presented along with a discussion of uncertainties in utilizing this data for various code calibration purposes. The existence of internal swirl is discussed. The two empirical coefficients in ENERGY are determined from the data within a certain range of accuracy. This range is dictated to a large extent by the accuracy of the experiments and to a smaller extent by the ability of the code to utilize all sets of data in each experiment. The effect of geometry and bundle size on mixing and swirl flow is discussed. A realistic estimate of the degree of accuracy within which we can predict temperature distribution within the bundle and along the duct of a 217-pin wire wrapped fuel assembly of an LMFBR is presented. Gaps in data which need to be filled in to enhance our confidence in predicting coolant te! mperature distributions in a 217-pin LMFBR fuel bundle, are given. A brief description of two experiments that would fill these data gaps is presented. A novel experiment which would be very useful for both fuel and poison assembly mixing studies is described. Conclusions drawn from this study are believed to be quite general in nature. Statement of responsibility on title-page reads: E.U. Khan, N.E. Todreas, W.M. Rohsenow , and A.A. Sonin; "September 1974."; Includes bibliographical references 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89485 Comments on the calculation of thermodynamic and transport properties of helium Eaton, Thomas E. Includes bibliographical references (leaves 28-32) 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89484 Analysis of operating data related to power and flow distribution in a PWR Herbin, Henry Christophe; Lanning, David D.; Todreas, Neil E.; Kirschner, Brian W.; Ladieu, Alan Edward The analysis of the effects of the uncertainties associated with temperature and power measurements in the Connecticut Yankee Reactor leads to the evaluation of the uncertainty associated with the effective flow factor. The effective flow factor is defined as the normalized ratio of the average assembly power to the coolant temperature use in each instrumented fuel assembly. Analysis of operating data indicates that the effective flow factor is a measure of the quality of agreement between the reactor physics and the thermal hydraulic analysis of the core. The methods given are also used for the evaluation of the uncertainties associated with the peaking factors, including the results of a sensitivity analysis developed with the code INCORE. Flow calculations have been performed with the code COBRA III C. The original version of the code COBRA III C has been expanded and a method is given to easily handle any further change in the code. A sensitivity a! nalysis, using the code COBRA III C shows the weak sensitivity of the exit conditions of the coolant on most input parameters and on the inlet flow distribution of the coolant selected for the calculation. This low sensitivity indicates that the information obtained from the assembly exit thermocouple cannot be used for the determination of the cross flow pattern between the fuel assemblies. Statement of responsibility on title-page reads: Henry C. Herbin, David D. Lanning, Neil E. Todreas, Brian W. Kirschner, [and] Alan E. Ladieu; "Issued: May 1974."; Substantially the same as a Nuclear Engineering thesis in the M.I.T. Dept. of Nuclear Engineering, 1974; Includes bibliographical references (leaves 141-143) 1974-01-01T00:00:00Z https://hdl.handle.net/1721.1/89483 Evaluation of high performance LMFBR blanket configurations Brown, Gilbert Jay; Driscoll, Michael J. Substantially the same as a Ph. D. thesis by G.J. Brown in the Dept. of Nuclear Engineering, MIT, 1974; Includes bibliographical references (pages 249-254) 1974-01-01T00:00:00Z