Friday, June 29, 2012

Post # 66: ORNL Light Water Reactor Safety Publications

I've been asked by some to present a more complete listing of Oak Ridge National Laboratory's historical reports on light water reactor safety (especially boiling water reactor safety) research.  Luckily, before I retired from ORNL last year I undertook to compile such a list for my own reference use.  So, with gratitude to Ms. Linda Dockery of ORNL, who spent quite a few hours at the keyboard creating this list,  I present an incomplete bibliography here.  One thing of interest to note.  If my count is correct, the ORNL team published almost 250 publications between 1980 and 1995 alone.
Quite a productive group !


ORNL LIGHT WATER REACTOR SAFETY PUBLICATIONS – 
AN INCOMPLETE BIBLIOGRAPHY COVERING THE PERIOD 1975 – 2010


1975
Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1975, A. P. Malinauskas, R. A. Lorenz, M. F. Osborne, J. L. Collins, and S. R. Manning, ORNL-TM-5021, Oak Ridge National Laboratory, September 1975.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1975, A. P. Malinauskas, R. A. Lorenz, M. F. Osborne, J. L. Collins, and S. R. Manning,ORNL-TM-5143, Oak Ridge National Laboratory, November 1975.
1976
Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–December 1975, R. A. Lorenz, J. L. Collins, and S. R. Manning, ORNL-TM-5290, Oak Ridge National Laboratory, March 1976.

Knudsen Cell-Mass Spectrometer Studies of Cesium-Urania Interactions, J. L. Collins, M. F. Osborne, A. P. Malinauskas, R. A. Lorenz, and S. R. Manning, ORNL/NUREG/TM-24, Oak Ridge National Laboratory, June 1976.

Behavior of Iodine, Methyl Iodide, Cesium Oxide, and Cesium Iodide in Steam and Argon, R. A. Lorenz, M. F. Osborne, J. L. Collins, S. R. Manning, and A. P. Malinauskas, ORNL/NUREG/TM-25, Oak Ridge National Laboratory, July 1976.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period January–March 1976, R. A. Lorenz, J. L. Collins, S. R. Manning, and A. P. Malinauskas, ORNL/NUREG/TM-30, Oak Ridge National Laboratory, July 1976.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1976, R. A. Lorenz, J. L. Collins, S. R. Manning, O. L. Kirkland, and A. P. Malinauskas, ORNL/NUREG/TM-44, Oak Ridge National Laboratory, August 1976.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1976, R. A. Lorenz, J. L. Collins, and O. L. Kirkland, ORNL/NUREG/TM-73, Oak Ridge National Laboratory, December 1976.
1977
Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–December 1976, R. A. Lorenz, J. L. Collins, and O. L. Kirkland, ORNL/NUREG/TM-88, Oak Ridge National Laboratory, March 1977.

“Fission Product Release from Simulated LWR Fuel,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, in Proceedings of the Specialist Meeting on the Behavior of Water Reactor Fuel Elements Under Accident Conditions, Spatind, Norway, Sept. 13–16, 1976 (CSNI Report No. 13), Part Two, Session I, Subsession I.3, March 1977.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1977, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, O. L. Kirkland, and R. L. Towns, ORNL/NUREG/TM-139, Oak Ridge National Laboratory, September 1977.
1978
Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1977, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, O. L. Kirkland, and R. L. Towns, ORNL/NUREG/TM-110, Oak Ridge National Laboratory, January 1978.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–September 1977, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, ORNL/NUREG/TM-186, Oak Ridge National Laboratory, March 1978.

“Fission Product Release from Highly Irradiated Fuel Under SFTA and LOCA Conditions,” J. L. Collins and R. A. Lorenz, presented at the 80th Annual Meeting of the American Ceramic Society, Detroit, MI, May 1978.

“Modeling Fission Product Release from Ruptured LWR Fuel Rods,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, presented at the 1978 Annual Meeting of the American Nuclear Society, San Diego, CA, June 1978.

Fission Product Source Terms for Loss-of-Coolant Accident: Summary Report, R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, NUREG/CR-0091 (ORNL/NUREG/TM-206), Oak Ridge National Laboratory, June 1978.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period January–March 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0116 (ORNL/NUREG/TM-208), Oak Ridge National Laboratory, June 1978.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0370 (ORNL/NUREG/TM-242), Oak Ridge National Laboratory, September 1978.

Fission Product Release from Simulated LWR Fuel, R. A. Lorenz, J. L. Collins, and S. R. Manning, NUREG/CR-0274 (ORNL/NUREG/TM-154), Oak Ridge National Laboratory, October 1978.
“Fission Product Source Terms for the LWR Loss-of-Coolant Accident,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, in Proceedings of the ENS/ANS International Topical Meeting on Nuclear Power Reactor Safety, Brussels, Belgium, Oct. 16–19, 1978, Vol. 3, p. 2293 (October 1978).

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0493 (ORNL/NUREG/TM-280), December 1978.
1979
“Fission-Product Source Terms for the Light Water-Reactor Loss-of-Coolant Accident,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, Nuclear Technology, 46(3), p. 404–410 (1979).

“Fission-Product Release During LWR Loss-of-Coolant Accidents,” A. P. Malinauskas, R. A. Lorenz, and J. L. Collins, Transactions of the American Nuclear Society, 32, p. 651 (1979).

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–December 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0682 (ORNL/NUREG/TM-308), Oak Ridge National Laboratory, April 1979.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period January–March 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, 
NUREG/CR-917 (ORNL/NUREG/TM-332), Oak Ridge National Laboratory, August 1979.
“Fission Product Release from LWR Fuel Defected in Steam in the Temperature Range 500 to 1600°C,” R. A. Lorenz, J. L. Collins, M. F. Osborne, and A. P. Malinauskas, in Proceedings of the Specialists’ Meeting on the Behavior of Defected Zirconium Alloy Clad Ceramic Fuel in Water Cooled Reactors, Chalk River, Canada, Sept. 17–21, 1979, International Atomic Energy Agency, IWGFPT/6, 1979.

“Chemical Behavior of Fission Products Released from Highly Irradiated LWR Fuel Under Accident Conditions,” J. L. Collins, R. A. Lorenz, A. P. Malinauskas, and M. F. Osborne, presented at the Basic Science and Nuclear Divisions of the American Ceramic Society’s Joint Fall Meeting, New Orleans, LA, October 1979.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-1061 (ORNL/NUREG/TM-348), Oak Ridge National Laboratory, October 1979.
1980
“Fission-Product Release from High Gap-Inventory LWR Fuel Under LOCA Conditions,” R. A. Lorenz, J. L. Collins, M. F. Osborne, et al., Transactions of the American Nuclear Society, 34, pp. 462–463 (1980).
Fission Product Release from Highly Irradiated LWR Fuel, R. A. Lorenz, J. L. Collins, A. P. Malinauskas, O. L. Kirkland, and R. L. Towns, NUREG/CR-0722 (ORNL/NUREG/TM-287/R2), Oak Ridge National Laboratory, February 1980.
Fission Product Source Terms for the LWR Loss-of-Coolant Accident, R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, NUREG/CR-1288 (ORNL/NUREG/CR-321), Oak Ridge National Laboratory, August 1980.
Fission Product Release from Highly Irradiated LWR Fuel Heated to 1300–1600°C in Steam, R. A. Lorenz, J. L. Collins, A. P. Malinauskas, M. F. Osborne, and R. L. Towns, NUREG/CR-1386 (ORNL/NUREG/TM-346), Oak Ridge National Laboratory, November 1980.
1981
“The Adsorption-Desorption Characteristics of Iodine on Graphite,” R. A. Lorenz, Transactions of the American Nuclear Society, 39, pp. 444–445 (1981).

“Sorption of Iodine on Low-Chromium Alloy-Steel,” M. F. Osborne, R. B. Briggs, and R. P. Wichner, Transactions of the American Nuclear Society, 38, pp. 463–464 (June 1981).

Fission Product Release from BWR Fuel Under LOCA Conditions, R. A. Lorenz, J. L. Collins, M. F. Osborne, R. L. Towns, and A. P. Malinauskas, NUREG/CR-1773 (ORNL/NUREG/TM-388), Oak Ridge National Laboratory, July 1981.

Station Blackout at Browns Ferry Unit One—Accident Sequence Analysis, D. H. Cook, R. M. Harrington, S. R. Greene, S. A. Hodge, and D. D. Yue, NUREG/CR-2182, Vol. 1 (ORNL/NUREG/TM-455/V1), Oak Ridge National Laboratory, November 1981.
1982
Station Blackout at Browns Ferry Unit One—Iodine and Noble Gas Distribution and Release, R. P. Wichner, C. F. Weber, R. A. Lorenz, W. Davis, Jr., S. A. Hodge, and A. D. Mitchell, NUREG/CR-2182, Vol. 2 (ORNL/NUREG/TM-455/V2), Oak Ridge National Laboratory, 1982.

“Fission-Product Release from LWR Fuel in Steam,” M. F. Osborne, R. A. Lorenz, and R. P. Wichner, Transactions of the American Nuclear Society, 43, pp. 357–359 (1982).

“Investigation of Drywell Flooding at the Browns Ferry Nuclear Plant,” S. R. Greene, W. A. Condon, and H. L. Dodds, Jr., presented at American Nuclear Society Summer Meeting, Los Angeles, CA, June 1982.
“Application of MARCH to BWR Severe Accident Analysis,” S. R. Greene, in Proceedings of the 10th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1982.

“Improvement of MARCH for BWR Applications,” S. R. Greene, in Proceedings of the 10th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1982.

SBLOCA Outside Containment at Browns Ferry Unit One–Accident Sequence Analysis, W. A. Condon, R. M. Harrington, S. R. Greene, and S. A. Hodge, NUREG/CR2672, Volume 1 (ORNL/TM-8119/V1), Oak Ridge National Laboratory, November 1982.

“Severe Accident Sequence Analysis (SASA),” S. R. Greene, presented to R. Minogue, Director, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Oak Ridge, TN, December 10, 1982.
1983
SBLOCA Outside Containment at Browns Ferry Unit One—Volume 2. Iodine, Cesium, and Noble Gas Distribution and Release, R. P. Wichner, C. F. Weber, A. L. Wright, S. A. Hodge, R. A. Lorenz, and J. W. Nehls, NUREG/CR-2672, Volume 2 (ORNL/TM-8119/V2), Oak Ridge National Laboratory, 1983.

“Fission-Product Transport in A BWR Small-Break LOCA,” C. F. Weber, R. P. Wichner, S. A. Hodge,
et al., Transactions of the American Nuclear Society, 45, pp. 479–480 (1983).

“Iodine Volatility Under LWR Accident Conditions,” E. C. Beahm and W. E. Shockley, Transactions of the American Nuclear Society, 45, pp. 482–483 (1983).

Loss of DHR Sequences at Browns Ferry Unit One—Accident Sequence Analysis, D. H. Cook, S. R. Greene, R. M. Harrington, and S. A. Hodge, NUREG/CR-2973 (ORNL/TM-8532), Oak Ridge National Laboratory, May 1983.

“Fission Product Release from Fuel Under LWR Accident Conditions,” M. F. Osborne, R. A. Lorenz, K. S. Norwood, J. L. Collins, and R. P. Wichner, p. 4.1-1 in Proceedings of the International Meeting on Light-Water Reactor Severe Accident Evaluation, Cambridge, Mass., Aug. 28–Sept. 1, 1983,
CONF-830816, Vol. 1 (1983).

The Effect of Small-Capacity, High-Pressure Injection Systems on TQUV Sequences at Browns Ferry Unit One, R. M. Harrington and L. J. Ott, NUREG/CR-3179 (ORNL/TM-8635), Oak Ridge National Laboratory, September 1983.

“The Effect of Low Capacity Injection Systems on Transient Initiated Loss of Vessel Water Injection at Browns Ferry Unit One,” L. J. Ott, presented at the International Meeting on Light Water Reactor Severe Accident Evaluation, Cambridge MA, August 28–September 1, 1983.

“The Effect of Small-Capacity, High-Pressure Injection Systems on BWR Transient Initiated Loss of Injection Accident Sequences,” L. J. Ott, presented at the Eleventh Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 24–28, 1983.

Mechanistic Core-Wide Meltdown and Relocation Modeling for BWR Applications, M. Z. Podowski, R. P. Taleyarkhan, and R. T. Lahey, Jr., NUREG/CR-3525 (ORNL/Sub/81-9080/1), Oak Ridge National Laboratory, December 1983.
1984
Realistic Simulation of Severe Accidents in BWRs—Computer Modeling Requirements, S. R. Greene, NUREG/CR-2940 (ORNL/TM-8517), Oak Ridge National Laboratory, 1984.

“Behavior of Cesium, Iodine, and Tellurium in the Fission-Product Release Program at ORNL,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Transactions of the American Nuclear Society, 47, pp. 320–322 (1984).

Noble Gas, Iodine, and Cesium Transport in a Postulated Loss of Decay Heat Removal Accident at Browns Ferry, R. P. Wichner, C. F. Weber, S. A. Hodge, E. C. Beahm, and A. L. Wright, NUREG/CR-3617 (ORNL/TM-9028), Oak Ridge National Laboratory, 1984.

Data Summary Report for Fission Product Release Test HI-4, M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-3600 (ORNL/TM-9001), Oak Ridge National Laboratory, June 1984.

“Characterization and Chemistry of Fission Products Released from LWR Fuel Under Accident Conditions,” K. S. Norwood, J. L. Collins, M. F. Osborne, R. A. Lorenz, and R. P. Wichner, in Proceedings of the ANS Topical Meeting on Fission Product Behavior and Source Term Research, Snowbird, Utah, July 15–19, 1984.

Noble Gas, Iodine, and Cesium Transport in A Postulated Loss of Decay Heat Removal Accident at Browns Ferry, R. P. Wichner, S. A. Hodge, C. F. Weber, E. C. Beahm, and A. L. Wright, NUREG/CR-3617 (ORNL/TM-9028), Oak Ridge National Laboratory, August 1984.

Design, Construction, and Testing of a 2000°C Furnace and Fission Product Collection System, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, and C. S. Webster, NUREG/CR-3715 (ORNL/TM-9135), Oak Ridge National Laboratory, September 1984.

Pressure Suppression Pool Thermal Mixing, D. H. Cook, NUREG/CR-3471 (ORNL/TM-8906), Oak Ridge National Laboratory, October 1984.

“Chemical Factors Affecting Fission Product Transport in BWR Severe Accidents,” E. C. Beahm, R. P. Wichner, and C. F. Weber, Trans. of 12th Water Reactor Safety Research Information Meeting, U. S. Nuclear Regulatory Commission, Gaithersburg MD, October 23, 1984.

“Effects of Improved Modeling on Best Estimate BWR Severe Accident Analysis,” L. J. Ott and C. R. Hyman, presented at the Twelfth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 23, 1984.

“Behavior of Cs, I, and Te in the Fission Product Release Program at ORNL,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Transactions of 1984 Winter Meeting of American Nuclear Society, 47, pp. 320–22, Washington, DC, Nov. 11–16, 1984.

“Measurement and Characterization of Fission Products Released from LWR Fuel,” M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, and R. V. Strain, in Proceedings of the Fifth International Meeting on Thermal Nuclear Reactor Safety, , Karlsruhe, Federal Republic of Germany, Sept. 9–13, 1984, Vol. 3 (published December 1984 as KfK 3880/1, -/2, -/3) (1984).
1985
“A Chemical-Equilibrium Estimate of the Aerosols Produced in An Overheated Light Water-Reactor Core,” R. P. Wichner and R. D. Spence, Nuclear Technology, 70(3), pp. 376–393 (1985).

BWR–LTAS:  A Boiling Water Reactor Long-Term Accident Simulation Code, R. M. Harrington and L. C. Fuller, NUREG/CR-3764 (ORNL/TM-9163), Oak Ridge National Laboratory, February 1985.

Observed Behavior of Cesium, Iodine, and Tellurium in the ORNL Fission Product Release Program, J. L. Collins, M. F. Osborne, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-3930 (ORNL/TM-9316), Oak Ridge National Laboratory, February 1985.

Highlights Report for Fission Product Release Tests at Simulated LWR Fuel, M. F. Osborne, J. L. Collins, and R. A. Lorenz, ORNL/NRC/LTR-85/1, Oak Ridge National Laboratory, February 1985.

The Modeling of BWR Core Meltdown Accidents—For Application in the MELRPI.MOD2 Computer Code, B. R. Koh, S. H. Kim, R. P. Taleyarkhan, M. Z. Podowski, and R. T. Lahey, Jr., NUREG/CR-3889 (ORNL/Sub/81-9088/2), Oak Ridge National Laboratory, April 1985.

“Tellurium Precursor Effects on Iodine Transport in A BWR Accident,” C. F. Weber, R. P. Wichner, and S. A. Hodge, Transactions of the American Nuclear Society, 49, pp. 257–258 (June 1985).

Data Summary Report for Fission Product Release Test HI-5, M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-4037 (ORNL/TM-9437), Oak Ridge National Laboratory, June 1985.

The Absorption of Gaseous Iodine by Water Droplets, M. F. Albert, NUREG/CR-4081 (ORNL/TM-9488), Oak Ridge National Laboratory, July 1985. 

Data Summary Report for Fission Product Release Test HI-6, M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-4043 (ORNL/TM-9443), Oak Ridge National Laboratory, September 1985.

“Station Blackout Calculations for Browns Ferry,” L. J. Ott, C. R. Hyman, and C. F. Weber, presented at the Thirteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD,
October 22–25, 1985; also in Proc. 13th Water Reactor Safety Research Information Meeting, NUREG/CP-0071, 1985.

“Chemistry and Transport of Iodine in Containment,” E. C. Beahm, W. E. Shockley, and C. F. Weber, presented at the International Symposium on Source Term Evaluation for Accident Conditions, Columbus, OH, October 28–November 1, 1985.

“Fission Product Release and Fuel Behavior in Tests of LWR Fuel Under Accident Conditions,” M. F. Osborne, J. L. Collins, R. A. Lorenz, and R. V. Strain, in Proceedings of the International Symposium on Source Term Evaluation for Accident Conditions, Columbus, OH, Oct. 28–Nov. 1, 1985.
1986
“Fission Product Release from UO2 Under LWR Accident Conditions:  Recent Data Compared with Review Values,” M. F. Osborne, J. L. Collins, and R. A. Lorenz, presented at the 87th Annual Meeting of American Ceramic Society, Cincinnati, Ohio, May 5–9, 1985; also in Advances in Ceramics 17, pp. 191–97 (1986).

Chemistry and Transport of Iodine in Containment, E. C. Beahm, W. E. Shockley, C. F. Weber, S. J. Wisbey, and Y.-M. Wang, NUREG/CR-4697 (ORNL/TM-10135), Oak Ridge National Laboratory, 1986.

“Behavior of Fission Product Tellurium Under Severe Accident Conditions,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Proceedings of the International ANS/ENS Topical Meeting on Thermal Reactor Safety, San Diego, CA, February 2–6, 1986, ANS Order No. 700106 (ISBN Order No.
0-89448-121-5), February 1986.

Design and Final Safety Analysis Report for Vertical Furnace Fission Product Release Apparatus in Hot Cell B, Building 4501, M. F. Osborne, J. L. Collins, P. A. Haas, R. A. Lorenz, J. R. Travis, and C. S. Webster, NUREG/CR-4332 (ORNL/TM-9270), Oak Ridge National Laboratory, March 1986.

Highlights Report for Fission Product Release Test VI-1, M. F. Osborne, J. L. Collins, R. A. Lorenz, and T. Yamashita, ORNL/NRC/LTR-86/7, Oak Ridge National Laboratory, March 1986.

Loss of Control Air at Browns Ferry Unit One—Accident Sequence Analysis, R. M. Harrington and S. A. Hodge, NUREG/CR-4413 (ORNL/TM-9826), Oak Ridge National Laboratory, April 1986.
“Calculational Support for the DF-4 Experiment in the ACRR,” L. J. Ott, presented at the Severe Fuel Damage and Source Term Research Program Review Meeting, Oak Ridge, TN, April 1986.
“Effects of Lateral Separation of Oxidic and Metallic Core Debris on the BWR MK I Containment Drywell Floor,” C. R. Hyman, C. F. Weber, and S. A. Hodge, Committee on Safety of Nuclear Installations (CSNI), Specialists Meeting on Core Debris/Concrete Interactions, EPRI NP-5054-SR, Palo Alto CA, September 3–5, 1986.

“Iodine Behavior in Containment Under LWR Accident Conditions,” S. J. Wisbey, E. C. Beahm, W. E. Shockley, et al., Abstracts of Papers of the American Chemical Society, 192, pp. 92–NUCL (September 7, 1986). 

“The Role of BWR MK I Secondary Containments in Severe Accident Mitigation,” S. R. Greene and S. A. Hodge, in Transactions of the 14th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1986.

“Effects of Lateral Separation of Oxidic and Metallic Core Debris on the BWR MK I Containment Drywell Floor,” C. R. Hyman and C. F. Weber, Trans. of 14th Water Reactor Safety Information Meeting, NUREG/CP-0081, U. S. Nuclear Regulatory Commission, Gaithersburg, MD,
October 27–31, 1986.

“ORNL MELCOR BWR Assessment Efforts,” S. R. Greene, presented at the MELCOR Users Group Meeting, October 31, 1986.

Thermal-Hydraulic and Characteristic Models for Packed Debris Beds, G. E. Mueller and A. Sozer, NUREG/CR-4689 (ORNL/TM-10117), Oak Ridge National Laboratory, December 1986.
Highlights Report for Fission Product Release Test VI-2, M. F. Osborne, J. L. Collins, R. A. Lorenz, and T. Yamashita, ORNL/NRC/LTR-86/18, Oak Ridge National Laboratory, December 1986.

Response of the BWR Shroud Head Under Severe Accident Conditions, E. W. Barnes, M. Keyhani, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, December 31, 1986.
1987
“The Influence of Environment on Release Behavior and Chemical Forms of Fission Products Released Under LWR Accident Conditions,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Proceedings of Workshop on Chemical Reactivity of Oxide Fuel and Fission Product Release, Berkeley, Gloucestershire, England, April 7–9, 1987. 

“The Basic Chemistry Involved in the Internal-Gelation Method of Precipitating Uranium as Determined by pH Measurements,” J. L. Collins, M. H. Lloyd, and R. L. Fellows, Radiochimica Acta, 42, pp. 121–134 (1987).

“Severe Core Damage and Associated In-Vessel Fission-Product Release,” C. Allison, J. Rest, R. A. Lorenz, et al., Progress in Nuclear Energy, 20(2), pp. 89–132 (1987).

“Calculation of Absorbed Doses to Water Pools in Severe Accident Sequences,” C. F. Weber, Trans. Am. Nucl. Soc., 55, pp. 367–368 (1987).

Calculations of Iodine Source Terms in Support of NUREG-0956, E. C. Beahm, C. F. Weber, T. S. Kress, and R. J. Anderman, ORNL/NRC/LTR-86/17, Oak Ridge National Laboratory, 1987.

Effects of Lateral Separation of Oxidic and Metallic Core Debris on the BWR MK I Containment Drywell Floor, C. R. Hyman and C. P. Weber, NUREG/CR-4610 (ORNL/TM-10057), Oak Ridge National Laboratory, January 1987.

“Fission-Product Tellurium Release Behavior Under Severe Light-Water Reactor Accident Conditions,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Nuclear Technology, 77(1), pp. 18–31 (April 1987).
“The Absorption of Gaseous Iodine by Water Droplets,” M. F. Albert, J. S. Watson, and R. P. Wichner, Nuclear Technology, 77(2), pp. 161–174 (May 1987).

“Calculational Support for the DF-4 Experiment in the ACRR,” L. J. Ott, presented at the Severe Fuel Damage, Containment Loads, and Source Term Research Program Review Meeting, Silver Spring, MD, May 4–8, 1987.
“The Release and Transport of Fission Product Cesium in the TMI-2 Accident,” R. L. Lorenz and J. L. Collins, pp. 4-69–4-83 in the Proceedings of the Symposium on Chemical Phenomena Associated with Radioactivity Releases During Severe Nuclear Plant Accidents, held at Anaheim, CA,
Sept. 9–12, 1986, also published as NUREG/CP-0078, June 1987.

ORNL Severe Accident Analysis for 25% Power Operation at the Shoreham Nuclear Power Station, S. R. Greene and S. A. Hodge, Oak Ridge National Laboratory, June 18, 1987.

An Assessment of the Shoreham Nuclear Power Station’s Secondary Containment Severe Accident Mitigation Capability, S. R. Greene, ORNL LTR Report to U.S. Nuclear Regulatory Commission, June 26, 1987.

“Organic Iodide Formation During Severe Accidents in Light Water Nuclear-Reactors,” E. C. Beahm, Y. M. Wang, S. J. Wisbey, et al., Nuclear Technology, 78(1), pp. 34–42 (July 1987).

Highlights Report for Fission Product Release Test VI‑3, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, C. S. Webster, S. R. Daish, H. K. Lee, T. Nakamura, and Y. C. Tong, draft letter report to SFD Partners, July 1987.

“Experimental Studies of Fission-Product Release from Commercial Light-Water Reactor Fuel Under Accident Conditions,” M. F. Osborne, J. L. Collins, and R. A. Lorenz, Nuclear Technology, 78(2), pp. 157–169 (August 1987).

“Modeling of the Transport of Vapor Fission Product Species to Aerosols,” T. S. Kress, E. C. Beahm, and C. F. Weber, presented at the Workshop on Water-Cooled Reactor Aerosol Code Evaluation and Uncertainty Assessment, Brussels, Belgium, September 9–11, 1987.

“The Impact of BWR MK I Primary Containment Failure Dynamics on Secondary Containment Integrity,” S. R. Greene, in Transactions of the 15th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1987.

“Advanced Severe Accident Response Models for BWR Application,” L. J. Ott, presented at the Severe Fuel Damage, Containment Loads, and Source Term Research Program Review Meeting, Silver Spring, MD, October 19–23, 1987.
“Advanced Severe Accident Response Models for BWR Application,” L. J. Ott, presented at the Fifteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 29, 1987.
“Fission Product Iodine and Cesium Release Behavior Under Severe LWR Accident Conditions,” J. L. Collins, M. F. Osborne, R. A. Lorenz, and A. P. Malinauskas, Nucl. Technol. 81(10), pp. 78–94 (November 1987).

“Calculation of Absorbed Doses to Water Pools in Severe Accident Sequences,” C. F. Weber, presented at the Winter Meeting Am. Nucl. Soc., Los Angeles, CA, November 15–19, 1987.
1988
“Considerations for Severe Accident Management Strategies in a Pressurized Water Reactor,” J. J. Carbajo and J. C. Carter, Trans. Am. Nucl. Soc., 57, pp. 159–160 (1988).

“The Xenon Poisoning Problem Previous to the Chernobyl Accident,” J. J. Carbajo, Trans. Am. Nucl. Soc., 56, pp. 404–405 (1988).

Peach Bottom Containment Response Calculations for Unmitigated Short-Term Station Blackout, S. R. Greene, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, February 1, 1988.

“Fission-Product Iodine and Cesium Release Behavior under Light Water-Reactor Accident Conditions,” J. L. Collins, M. F. Osborne, R. A. Lorenz, et al., Nuclear Technology, 81(1), pp. 78–94 (April 1988).

“Proposed ORNL BWR Mark II Containment Response Considerations,” S. R. Greene, presented to NRC Staff, April 21, 1988.

Modeling of Time-Dependent Emergence of Core Debris from A Boiling Water Reactor Vessel Under Severe Accident Conditions, L. J. Ott and S. A. Hodge, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, April 22, 1988.

Small Scale BWR Core Debris Eutectics Formation and Melting Experiment—An Update, G. W. Parker, S. A. Hodge, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), April 22, 1988.

“Separate Effects Melt Experiments,” G. W. Parker, S. A. Hodge and L. J. Ott, presented at the Severe Accident Research Program Review Meeting, Albuquerque, NM, April 26, 1988.
Primary Containment Response Calculations for Unmitigated Short-Term Station Blackout at Peach Bottom (with Consideration of (1) Separated Metal and Oxide Pours and (2) Degassing of Drywell Concrete), S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, May 2, 1988.

“The Role of BWR Secondary Containments in Severe Accident Mitigation:  Issues and Insights from Recent Analyses,” S. R. Greene, in Proceedings of the 4th Workshop on Containment Integrity, Arlington, VA, NUREG/CR-0095, June 1988.

“Containment Venting as A Mitigation Technique for BWR MARK-1 Plant ATWS,” R. M. Harrington, Nuclear Engineering and Design, 108(1–2), pp. 55–69 (June 1988).

“Iodine Behavior in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, S. R. Daish, and
W. E. Shockley, presented at the 2nd CSNI Workshop of Iodine Chemistry in Reactor Safety, Toronto, Canada, June 2–3, 1988.

“Chemistry and Mass Transport of Iodine in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, W. E. Shockley, and S. R. Daish, presented at the Meeting Am. Chem. Soc., Toronto, Canada, June 4–9, 1988.

“Chemistry and Mass-Transport of Iodine in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, et al., Abstracts of Papers of the American Chemical Society, 195, Part 2, pp. 101–NUCL (June 5, 1988).

“The Chemistry and Behavior of Iodine Vapor Species in Nuclear Plant Air-Monitoring Sampling Lines,” A. L. Wright, B. R. Fish, E. C. Beahm, and C. F. Weber, presented at the 20th DOE/NRC Nuclear Air Cleaning Conf., Boston, MA, August 22–25, 1988.

“Trends Status,” T. S. Kress, E. C. Beahm, W. E. Shockley, and C. F. Weber, presented at Severe Accident Research Program Review Meeting, Bethesda, MD, October 17–21, 1988.

“Experiment-Specific Codes for the DF-4 and NRU Experiments with BWR Geometries,” L. J. Ott, presented at the Severe Accident Research Program Review Meeting, Bethesda, MD, October 17–21, 1988.
 “BWRSAR Calculations of Reactor Vessel Debris Pours for Peach Bottom Short-Term Station Blackout,” L. J. Ott and S. A. Hodge, presented at the Sixteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 27, 1988.

“Small Scale BWR Core Debris Eutectics Formation and Melting Experiment,” G. W. Parker, L. J. Ott, and S. A. Hodge, presented at the Sixteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 27, 1988.

“BWR Mark II and III Parametrics Program,” S. R. Greene and S. A. Hodge, ORNL NRC Program Review for E. Beckjord, Director, Office of Research, U. S. Nuclear Regulatory Commission, November 9, 1988.

Primary Containment Response Calculations for Unmitigated Short-Term Station Blackout at Peach Bottom (with Consideration of (1) Separated Metal and Oxide Pours, (2) Melting of Reactor Vessel Bottom Head, and (3) Degassing of Drywell Concrete, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, November 28, 1988.

Boiling Water Reactor Severe Accident Technology at Oak Ridge—Purpose and Goals, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, December 6, 1988.
1989
“Station Blackout Severe Accidents in APWRs,” J. J. Carbajo, presented at the International Topical Meeting on Probability, Reliability, and Safety Assessment, PSA 1989, April 2-7, 1989, Pittsburgh, PA, 1989.

Boiling Water Reactor Severe Accident Response (BWRSAR) Code Description and Assessment, S. A. Hodge and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor, Oak Ridge National Laboratory, February 1, 1989.

Primary Containment Response Calculations for Unmitigated Short-Term Station Blackout at Peach Bottom, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor, Oak Ridge National Laboratory (1989).

“Time-Dependent Release of Fission Products from LWR Fuel Under Severe Accident Conditions,” M. F. Osborne, R. A. Lorenz, J. L. Collins, and T. Nakamura, Proceedings of International ENS/ANS Conference on Thermal Reactor Safety, Avignon, France, Oct. 3–7, 1988 [published October 1988 (Vols. 1–4) and March 1989 (Vols. 5–6)].

Boiling Water Reactor Severe Accident Response (BWRSAR) Post-Test Analyses of the ACRR DF-4 Experiment, L. J. Ott, presented at the Severe Accident Research Program Review Meeting, Idaho Falls, ID, April 11, 1989.
Failure Modes of the BWR Reactor Vessel Bottom Head, S. A. Hodge and L. J. Ott, ORNL/M-1019 (Letter Report to the BWRSAT Program NRC Technical Monitor [Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC]), Oak Ridge National Laboratory, May 10, 1989.

“Iodine Behavior in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, and W. E. Shockley, presented at ICHMT Seminar on Fission Product Transport Processes in Reactor Accidents, Dubrovnik, Yugoslavia, May 22–26, 1989.

“TRENDS:  A Code for Modeling Iodine Behavior in Containment During Severe Accidents,” C. F. Weber, E. C. Beahm, T. S. Kress, S. R. Daish, and W. E. Shockley, in Proc. ICHMT Int. Seminar on Heat and Mass Transfer Aspects of Fission Product Releases, Dubrovnik, Yugoslavia, May 22–26, 1989.

Data Summary Report for Fission Product release Test VI-1, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, C. S. Webster, and T. Yamashita, NUREG/CR-5339 (ORNL/TM-11104), Oak Ridge National Laboratory, June 1989.

“Advanced Severe Accident Response Models for BWR Application,” L. J. Ott, Nuclear Engineering and Design, 115(2–3), pp. 289–303 (July 1989).

Post-Test Analyses of the DF-4 BWR Experiment Using the BWRSAR/DF4 Code, L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/M-1020, Oak Ridge National Laboratory, August 11, 1989.

Data Summary Report for Fission Product release Test VI-2, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, and C. S. Webster, NUREG/CR-5340 (ORNL/TM-11105), Oak Ridge National Laboratory, September 1989.

BWR Severe Accident Experiment Needs, L. J. Ott, presented at the International CORA Workshop 1989, Karlsruhe, FRG, September 25–27, 1989.
BWR Specific Effects in Severe Accident Behavior, L. J. Ott, presented at the International CORA Workshop 1989, Karlsruhe, FRG, September 25–27, 1989.
Experiment-Specific BWR Code Analysis of the ACRR DF-4 Experiment, L. J. Ott, presented at the International CORA Workshop 1989, Karlsruhe, FRG, September 25–27, 1989.
Description of the NRU FLHT-6 Experiment-Specific Code and Preliminary Pre-Test Predictions, L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/M-1021, Oak Ridge National Laboratory, September 22, 1989.

Thermophoretic Transport of Particles that Act as Volumetric Heat Sources in Natural Convection Flow, J. C. Conklin and R. J. Krane, ORNL-6573, Oak Ridge National Laboratory, October 1989.

The DF-4 Fuel Damage Experiment in ACRR With A BWR Control Blade and Channel Box, R. O. Gauntt, R. D. Gasser, and L. J. Ott, NUREG/CR-4671 (SAND-86-1443), Sandia National Laboratories, Albuquerque, NM, November 1989.

“Behavior of Iodine in Containment,” E. C. Beahm, C. F. Weber, M. L. Brown, and W. E. Shockley Presentation, Oak Ridge National Laboratory, Oak Ridge, TN, December 5, 1989.
1990
The Response of BWR Mark II and Mark III Containments to Short-Term Station Blackout Severe Accident Sequences, S. R. Greene et al., ORNL/NRC/LTR-89/13, Oak Ridge National Laboratory, 1990.

“Small-Scale BWR Core Debris Eutectics Formation and Melting Experiment,” G. W. Parker, L. J. Ott, and S. A. Hodge,” Nuclear Engineering and Design, 121 (1990).

“BWRSAR Calculations of Reactor Vessel Debris Pours for Peach Bottom Short-Term Station Blackout,” S. A. Hodge and L. J. Ott, Nuclear Engineering and Design, 121 (1990).

Debris Bed Behavior in the BWR Lower Plenum—The BWRSAR Approach with Recommended Improvements for Installation in MELCOR, S. A. Hodge, ORNL/NRC/LTR-90/10, Oak Ridge National Laboratory, April 26, 1990.

BWR Experiment Analyses and Principal BWR Severe Accident Uncertainties, L. J. Ott, presented at the Severe Accident Research Program Partners Review Meeting, Brookhaven National Laboratory, Upton, NY, May 1, 1990.
Boiling Water Reactor Severe Accident Models for MELCOR, S. A. Hodge, ORNL/NRC/LTR-90/13, Oak Ridge National Laboratory, May 22, 1990.

Data Summary Report for Fission Product Release Test VI‑3, M. F. Osborne, R. A. Lorenz, J. L. Collins, J. R. Travis, C. S. Webster, H. K. Lee, T. Nakamura, and Y.-C. Tong, NUREG/CR-5480 (ORNL/TM-11399), Oak Ridge National Laboratory, June 1990.

“Fission Product Release from Commercial vs Simulated Fuels in LWR Accident Studies,” M. F. Osborne, H. Albrecht, R. A. Lorenz, and J. L. Collins, presented at the 1990 Annual Meeting of the American Nuclear Society, Nashville, TN, June 10–14, 1990; summary published in Trans. 1990 ANS Annl. Mtg., 61, pp. 251–52, American Nuclear Society (June 1990).

“The Role of BWR Secondary Containments in Severe Accident Mitigation: Issues and Insights from Recent Analyses,” S. R. Greene, Nuclear Engineering and Design, 120(1), pp. 75–86 (June 1, 1990).

A Survey of Current Models of BWR Core Plate Failure Used in the Severe-Accident Codes APRIL, BWRSAR, MELCOR, MELPROG, and SCDAP/RELAP, L. J. Ott and W. I. Van Rij, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/14, Oak Ridge National Laboratory, June 29, 1990.

Kinetic Modeling and Parameter Optimization of Iodine Hydrolysis Reactions, C. F. Weber, M.S. Thesis, University of Tennessee, August 1990.

“BWRSAR Calculations of Reactor Vessel Debris Pours for Peach Bottom Sort-Term Station Blackout,” S. A. Hodge and L. J. Ott, Nuclear Engineering and Design, 121(3), pp. 327–339 (August 1990).

“Small-Scale BWR Core Debris Eutectics Formation and Melting Experiment,” G. W. Parker, L. J. Ott, and S. A. Hodge, Nuclear Engineering and Design, 121(3), pp. 341–347 (August 1990).

“Contain Calculations of Debris Conditions Adjacent to the BWR MARK-1 Drywell Shell During the Later Phases of a Severe Accident,” C. R. Hyman, Nuclear Engineering and Design, 121(3), pp.
379–393 (August 1990).

Experiment-Specific BWR Code Analysis of the CORA-17 Experiment, L. J. Ott, presented at the International CORA Workshop 1990, Karlsruhe, Germany, September 28–October 2, 1990.
Post-Test Analyses of the CORA-16 Experiment, L. J. Ott, presented at the International CORA Workshop 1990, Karlsruhe, Germany, September 28–October 2, 1990.
A Survey of Current Models of BWR Lower Head Failure Used in the Severe-Accident Codes APRIL, BWRSAR, MELCOR, MELPROG, and SCDAP/RELAP, L. J. Ott and W. I. Van Rij, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/26, Oak Ridge National Laboratory, September 30, 1990.

Description of the CORA BWR Experiment-Specific Code, L. J. Ott, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/23, Oak Ridge National Laboratory, September 30, 1990.

Post-Test Analyses of CORA BWR Experiments, L. J. Ott, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/24, Oak Ridge National Laboratory, September 30, 1990.

“Results of Recent ORNL Fission Product Release Tests,” R. A. Lorenz, M. F. Osborne, and J. L. Collins, presented at the 18th Water Reactor Safety Meeting at Rockville, MD, October 22–24, 1990.

“Experiment-Specific Analyses in Support of Code Development,” L. J. Ott, presented at the Eighteenth Water Reactor Safety Research Information Meeting, Rockville, MD, October 23, 1990.
1991
“Effect of an In-Containment Refueling Water Storage Tank on Station Blackout Accidents,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 311–313 (1991).

“In-Vessel Hydrogen Generation During Station Blackout Severe Accident Sequences in APWRs,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 313–314 (1991).

“In-Vessel Hydrogen Generation During a Station Blackout Severe Accident,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 327–328 (1991).

“Direct Containment Heating Calculations for the Zion Plant,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 328–329 (1991).

“Characterization of Debris/Concrete Interactions for Advanced Research Reactor and Commercial BWR Severe Accidents,” C. R. Hyman, R. P. Taleyarkhan, and S. R. Greene, in Proceedings for the American Nuclear Society Winter Meeting, 1991.

“Severe LWR Accident Studies of Fission Product Release and Fuel Behavior at ORNL,” M. F. Osborne, R. L. Lorenz, and J. L. Collins, presented at the 15th Nuclear Safety Research Reactor Technical Review Meeting, Washington, DC, 1991.

“Characterization of Debris/Concrete Interactions for Commercial BWR Mark II and Research Reactors Severe Accidents,” C. R. Hyman, R. P. Taleyarkhan, and S. R. Greene, presented at the National Heat Transfer Conference, Minneapolis, MN, 1991.

Calculation of Absorbed Doses to Water Pools in Severe Accident Sequences, C. F. Weber, ORNL/TM-11970, Oak Ridge National Laboratory, 1991.

Data Summary Report for Fission Product Release Test VI-4, M. F. Osborne, R. A. Lorenz, J. L. Collins, J. R. Travis, C. S. Webster, and T. Nakamura, NUREG/CR-5481 (ORNL/TM-11400), Oak Ridge National Laboratory, January 1991.

“Identification and Initial Assessment of BWR In-Vessel Accident Management Strategies,” S. A. Hodge, presented at Accident Management Information Exchange Meeting USNRC—BMU/GRS, April 15, 1991.

In-Vessel Phenomena—CORA, L. J. Ott and W. I. van Rij, presented at the Cooperative Severe Accident Research Program (CSARP) Semiannual Review Meeting, Bethesda, MD, May 7, 1991.
The Response of BWR Mark III Containments to Short-Term Station Blackout Severe Accident Sequences, S. R. Greene, S. A. Hodge, C. R. Hyman, B. W. Patton, and M. L. Tobias, NUREG/CR-5571 (ORNL/TM-11549), Oak Ridge National Laboratory, June 1991.

“Fission Product Behavior,” T. S. Kress, E. C. Beahm, and C. F. Weber, presented at the Annual Meeting American Nuclear Society, Orlando, FL, June 2–6, 1991; also published in Trans. Amer. Nucl. Soc., 63, pp. 260 (1991).

CCCTF Shakedown Tests, Draft, R. N. Morris, J. L. Collins, W. A. Gabbard, J. C. Whitson, and M. F. Osborne, July 1991. 

“Atmospheric Effects on Fission Product Behavior at Severe Accident Conditions,” M. F. Osborne, R. A. Lorenz, and J. L. Collins, in Proceedings of the American Nuclear Society International Topical Meeting on the Safety of Thermal Reactors, Portland, OR, July 21–25, 1991.

Significant BWR Technical Developments from the CORA Program, L. J. Ott, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-91/14, Oak Ridge National Laboratory, July 31, 1991.
Report of Foreign Travel by C. F. Weber, September 7–15, 1991, C. F. Weber, ORNL/CSD/FTR-4049, Oak Ridge National Laboratory, September 1991.

Data Summary Report for Fission Product Release Test VI-5, M. F. Osborne, R. A. Lorenz, J. R. Travis, C. S. Webster, and J. L. Collins, ORNL/CR-5668 (ORNL/TM-11743), Oak Ridge National Laboratory, September 1991.

“Iodine Chemical Forms in LWR Severe Accidents,” C. F. Weber, E. C. Beahm, and T. S. Kress, presented at the 3rd Committee for Safety in Nuclear Installations Workshop on Iodine Chemistry in Reactor Safety, Tokai-mura, Japan, September 11–13, 1991.

“Iodine Chemical Forms in LWR Accidents:  Preliminary Presentation of Results,” E. C. Beahm, C. F. Weber, and T. S. Kress, presented at Committee for Safety in Nuclear Installations Iodine Specialists’ Meeting, Shinhara, Mito-Shi, Japan, September 11–13, 1991; also in Nucl. Technol., 101, pp.
262–269 (1993).

Current Status of the Experimental Data and Phenomenological Modeling in BWR Core Melt Progression, L. J. Ott and W. I. Van Rij, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. Ali Behbahani, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-91/16, Oak Ridge National Laboratory, September 31, 1991.

Assessment of Two BWR Accident Management Strategies, S. A. Hodge and M. Petek, in
CONF-911079-2, October 1991.

BWR Lower Plenum Debris Bed Models for MELCOR, S. A. Hodge and L. J. Ott, presented at the Nineteenth Water Reactor Safety Research Information Meeting, Bethesda, MD, October 28–30, 1991.
“Iodine Chemical Forms in LWR Severe Accidents,” E. C. Beahm, C. F. Weber, T. S. Kress, and G. W. Parker, in Proceedings of the 19th Water Reactor Safety Information Meeting, 2, pp. 325–342, Bethesda, MD, October 28–30, 1991.

BWR Mark II Ex-Vessel Corium Interaction Analyses, S. R. Greene, A. E. Levin, C. R. Hyman, A. Sozer, and R. P. Taleyarkhan, NUREG/CR-5623 (ORNL/TM-11644), Oak Ridge National Laboratory, November 1991.

“ORNL Studies of Fission Product Release Under LWR Accident Conditions,” F. F. Osborne, R. A. Lorenz, and J. L. Collins, presented at the Second Workshop on LWR Severe Accident Research A JERI, Tokoyo, Japan, November 25–27, 1991 (also published in the proceedings of the meeting).
1992
“Spent-Fuel Decay Heat and Source Term Uncertainties,” J. J. Carbajo, Trans. Am. Nucl. Society, 65, pp. 66–67 (1992).

“Effect of the Timing of Vessel Depressurization on a Short-Term Station Blackout in a BWR-4 Performed with the MELCOR Code,” J. J. Carbajo, Trans. Am. Nucl. Society, 66, pp. 327–328 (1992).

The Core Conduction Cooldown Test Facility: Current Status and Issues, R. N. Morris, C. A. Baldwin, J. L. Collins, L. C. Emerson, W. A. Gabbard, C. M. Malone, B. F. Myers, M. F. Osborne, M. L. Peters, J. C. Whitson, and J. L. Wright, ORNL/NRP-91/7, Oak Ridge National Laboratory, January 1992.

Core Conduction Cooldown Test Facility Shakedown Tests, R. N. Morris, J. L. Collins, W. A. Gabbard, and J. C. Whitson, ORNL/NRP-91/25, Oak Ridge National Laboratory, May 1992.

HRB-17 and HRB-18 HEU TRISO UCO Unbonded Irradiated Particle Core Conduction Cooldown Tests, R. N. Morris, C. A. Baldwin, J. L. Collins, C. M. Malone, W. A. Gabbard, J. R. Travis, C. S. Webster, J. C. Whitson, J. L. Wright, and M. J. Kania, ORNL/NRP-92/9, Oak Ridge National Laboratory, June 1992.

Iodine Chemical Forms in LWR Severe Accidents, Final Report, E. C. Beahm, C. F. Weber, T. S. Kress, and G. W. Parker, ORNL/TM-11861 (NUREG/CP-0119), Oak Ridge National Laboratory, 1992.

Iodine Evolution and pH Control, E. C. Beahm, R. A. Lorenz, and C. F. Weber, ORNL/TM-12242, Oak Ridge National Laboratory, 1992.

Models of Iodine Behavior in Reactor Containments, E. C. Beahm, C. F. Weber, T. S. Kress, and G. W. Parker, ORNL/TM-12202, Oak Ridge National Laboratory, 1992.

“MELCOR Modifications for SBWR Applications,” C. R. Hyman, R. L. Sanders, S. R. Greene, and S. A. Hodge, in Proceedings of the Twentieth Water Reactor Safety Information Meeting, Bethesda, MD, 1992.

“Optimal Determination of Rate Coefficients in Multiple Reactor Systems,” C. F. Weber, E. C. Beahm, and J. S. Watson, Comput. Chem., 16(4), pp. 325–333 (1992).

“Iodine Chemical Forms in LWR Severe Accidents,” E. C. Beahm and C. F. Weber, presentation to Advisory Committee Reactor Safeguards (ACRS), Bethesda, MD, January 7–8, 1992.

BWR Core Melt Progression Phenomena – Experimental Analyses, L. J. Ott, presented at the Cooperative Severe Accident Research Program (CSARP) Review Meeting, Bethesda, MD, May 4–8, 1992.
“Iodine Transport in Containment,” E. C. Beahm, C. F. Weber, and T. S. Kress, presented at Cooperative Severe Accident Research Program (CSARP) Meeting, Bethesda, MD, May 4–8, 1992.

An Assessment of MELCOR Modeling Modifications Required for the Simplified Boiling Water Reactor, S. R. Greene, S. A. Hodge, C. R. Hyman, and R. L. Sanders, ORNL/NRC/LTR-92/11, Oak Ridge National Laboratory, May 29, 1992.

BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP, F. P. Griffin, ORNL/NRC/LTR-92/12, Oak Ridge National Laboratory, June 15, 1992.

Post-Test Analysis of the CORA-16 and CORA 17 BWR Experiments, L. J. Ott, ORNL/NRC/LTR-92/17, Oak Ridge National Laboratory, July 10, 1992.

“ORNL Studies of Fission-Product Release Under LWR Severe Accident Conditions,” M. F. Osborne and R. A. Lorenz, Nuclear Safety, 33(3), pp. 344–365 (July–September 1992).

“Iodine Chemical Forms in LWR Severe Accidents,” C. F. Weber, E. C. Beahm, and T. S. Kress, pp. 414–430 in Proc. 3rd CSNI Workshop on Iodine Chemistry in Reactor Safety, Tokai-mura, Japan, September 11–13, 1992, NEA/CSNI/R(91)15, JAERI, 1992.

An Assessment of SCDAP/RELAP5 Modeling Modifications Required for the Simplified Boiling Water Reactor, F. P. Griffin, ORNL/NRC/LTR-92/25, Oak Ridge National Laboratory, September 30, 1992.

Thermal Shock Considerations for External Flooding of A BWR Reactor Vessel, S. A. Hodge and W. J. McAfee, ORNL/NRC/LTR-92-23, Oak Ridge National Laboratory, September 30, 1992.

Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, S. A. Hodge, J. C. Cleveland, T. S. Kress, and M. Petek, NUREG/CR-5869 (ORNL/TM-12080), Oak Ridge National Laboratory, October 1992; also published in CONF-921007-31, October 1992.

ORNL Pre- and Post-Test Analyses of BWR Experiments, L. J. Ott, presented at the International CORA Workshop 1992, Karlsruhe, Germany, October 6, 1992.
“Core Melt Source Reduction System (COMSORS) for Light-Water Reactors (Draft), presented at COMSORS Program Review Meeting, Tenera Bethesda Office, Bethesda, MD, October 20, 1992.

Report of Foreign Travel to KfK, Germany and Participation in the CORA-33 Experiment, L. J. Ott, letter report to the BWRCMP Program NRC Technical Monitor (Dr. Ali Behbahani, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-92/28, November 20, 1992.
BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP, Revision 1, F. P. Griffin and K. A. Smith, ORNL/NRC/LTR-92/12/R1, Oak Ridge National Laboratory, December 31, 1992.

Post-Test Analyses of the CORA-31 Slow Heatup BWR Experiment, L. J. Ott, ORNL/NRC/LTR-92/29, Oak Ridge National Laboratory, December 31, 1992.
1993
“Iodine Evolution and pH Control,” E. C. Beahm, R. A. Lorenz, and C. F. Weber, Trans. Am. Nucl. Soc., 59, pp. 387 (1993).

“Iodine Transport in a Severe Accident at the High Flux Isotope Reactor,” C. F. Weber and E. C. Beahm, Trans. Amer. Nucl. Soc., 68A, p. 275 (1993).

“MELCOR Simulation of a Large-Break LOCA at the High Flux Isotope Reactor,” R. H. Morris, S. E. Fisher, and S. R. Greene, in CONF-930601, Transactions of the American Nuclear Society, 68 (1993).

Severe Accident Source Term Characteristics for Selected Peach Bottom Sequences Predicted by the MELCOR Code, J. J. Carbajo, NUREG/CR-5942 (ORNL/TM-12229), Oak Ridge National Laboratory, 1993.
“Vessel Failure Time for a Low-Pressure Short-Term Station Blackout in a BWR-4,” J. J. Carbajo, Trans. Am. Nucl. Society, 67, pp. 306–307 (1993).

“MELCOR Code Analysis of a Severe Accident LOCA at Peach Bottom Plant,” J. J. Carbajo, Trans. Am. Nucl. Society, 69, pp. 319–321 (1993).

“Containment Failure Time and Mode for a Low-Pressure Short-Term Station Blackout in a BWR-4 with Mark-I Containment,” J. J. Carbajo and S. R. Greene, Trans. Am. Nucl. Society, 69, pp. 325–326 (1993).

MELCOR 1.8.2 Assessment:  Comparison of Fuel Fission Product Release Models to ORNL Vifuel Fission Product Release Experiment, S. R. Greene et al., ORNL/NRC/LTR-94/34, Oak Ridge National Laboratory, 1993.

“Fission-Product Transport Behavior,” T. S. Kress, E. C. Beahm, C. F. Weber, and G. W. Parker, Nuclear Technology, 101(3), pp. 262–269 (March 1993).

“Iodine Transport Models for LWR Accidents,” C. F. Weber and E. C. Beahm, in Cooperative Severe Accident Research Program Meeting (CSARP), Bethesda, MD, May 7, 1993.

“Iodine Transport in a Severe Accident at the High Flux Isotope Reactor,” C. F. Weber and E. C. Beahm, Annual Meeting Am. Nucl. Soc., 1993.

Current Status of BWR Experimental Analyses and the Need for Additional BWR Experiments, L. J. Ott, ORNL/NRC/LTR-93/12, Oak Ridge National Laboratory, June 30, 1993.

Post-Test Analyses of the CORA-33 Dry Core BWR Experiment, L. J. Ott, ORNL/NRC/LTR-93/21, Oak Ridge National Laboratory, August 31, 1993.

Post-Test Analyses of the CORA-28 Preoxidized BWR Experiment, L. J. Ott, ORNL/NRC/LTR-93/21, Oak Ridge National Laboratory, August 31, 1993.

Analysis and Interpretation of the Results of the CORA-33 Dry Core BWR Experiment, L. J. Ott, presented at the International CORA Workshop 1993, Karlsruhe, Germany, September 27, 1993.
“BWR Control Blade/Channel Box Interaction Models for SCDAP/RELAP5,” F. P. Griffin, presented at the 21st Water Reactor Safety Research Information Meeting, Bethesda, MD, October 25‑27, 1993.

Interpretation of the Results of the CORA-33 Dry Core BWR Test, L. J. Ott and S. Hagen, presented at the 21st Water Reactor Safety Research Information Meeting, Bethesda, MD, October 25–27, 1993.
“Containment Failure Time and Mode for a Low-Pressure Short-Term Station Blackout in a BWR-4 with Mark-III Containment,” S. R. Greene, in Proceedings of the American Nuclear Society Winter Meeting, San Francisco, CA, November 1993.

Current Status of ORNL FLHT-6 Experimental Analyses, L. J. Ott and K. A. Smith, ORNL/NRC/LTR-93/30, Oak Ridge National Laboratory, November 30, 1993.

“Iodine Evolution and pH Control,” E. C. Beahm, R. A. Lorenz, and C. F. Weber, in Proceedings of the Winter Meeting of the American Nuclear Society, San Francisco, CA, November 14–18, 1993.
Status of ORNL XR Experimental Analyses, L. J. Ott, ORNL/NRC/LTR-93/34, Oak Ridge National Laboratory, December 30, 1993.

BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP, Revision 2, F. P. Griffin, ORNL/NRC/LTR-92/12/R2, Oak Ridge National Laboratory, December 30, 1993.
1994
“Optimum Depressurization for a Short-Term Station Blackout in a BWR-6 with High Burnup Fuel,” J. J. Carbajo, Trans. Am. Nucl. Society, 71, pp. 324–326 (1994).

“MELCOR Sensitivity Studies for a Low-Pressure, Short-Term Station Blackout at the Peach Bottom Plant,” J. J. Carbajo, Nucl. Eng. and Des., 152, pp. 287–317 (1994).

Data Summary Report for Fission Product Release Test VI-6, M. F. Osborne, R. A. Lorenz, J. R. Travis, C. S. Webster, and J. L. Collins, ORNL/CR-6077 (ORNL/TM-12416), Oak Ridge National Laboratory, March 1994.

“Assessment of 2 BWR Accident Management Strategies,” S. A. Hodge and M. Petek, Nuclear Engineering and Design, 148(2–3), pp. 185–203 (July 1994).
Detailed Analysis of the BWR Dry Core Conditions for the XR2 Experiments, L. J. Ott and F. P. Griffin, presented at the International CORA Workshop 1994, Karlsruhe, Germany,
October 10–12,1994.
Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments, L. J. Ott and F. P. Griffin, ORNL/NRC/LTR-94/38, Oak Ridge National Laboratory, October 31, 1994.
1995
“MELCOR Calculations for a Low-Pressure, Short-Term Station Blackout in a BWR-6,” J. J. Carbajo, Trans. Am. Nucl. Society, 72, pp. 250–252 (1995).

“Recovery Sequences for a Station Blackout Accident at the Grand Gulf Nuclear Station,” J. J. Carbajo, Trans. Am. Nucl. Society, 72, pp. 226–228 (1995).

“Source Terms Released into the Environment for a Station Blackout Severe Accident at the Peach Bottom Atomic Power Station,” J. J. Carbajo, Proceedings of the Topical Meeting on Safety of Operating Reactors, pp. 259–266 (1995).

“Comparison of MELCOR and SCDAP/RELAP5 Results for a Low-Pressure, Short-Term Station Blackout at Browns Ferry,” J. J. Carbajo, Trans. Am. Nucl. Society, 73, pp. 259–269 (1995).

Comparison of MELCOR Modeling Techniques and Effects of Vessel Water Injection on a Low-Pressure, Short-Term Station Blackout at the Grand Gulf Nuclear Station, J. J. Carbajo, ORNL/TM-12771, Oak Ridge National Laboratory, 1995.

Comparison of MELCOR Results, Including Parametric Variations, to SCDAP/RELAP5 Results for a Station Blackout Accident at Browns Ferry, J. J. Carbajo, ORNL/NRC/LTR-95/25, Oak Ridge National Laboratory, 1995.

Description of the XR/BWR Experiment-Specific Code and Preliminary Post-Test Analyses of the XR1 Tests, L. J. Ott, ORNL/NRC/LTR-95/5, Oak Ridge National Laboratory, February 28, 1995.

BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP (Revision 3), F. P. Griffin, ORNL/NRC/LTR-92/12/R3, Oak Ridge National Laboratory, March 31, 1995.

Data Summary Report for Fission Product Release Test VI-7, M. F. Osborne, R. A. Lorenz, J. R. Travis, J. L. Collins, and C. S. Webster, NUREG/CR-6318 (ORNL/TM-12937), Oak Ridge National Laboratory, May 1995.

“Recent SCDAP/RELAP5 Improvements for BWR Severe Accident Simulations,” F. P. Griffin, presented at the 23rd Water Reactor Safety Information Meeting, Bethesda, MD, October 23‑25, 1995.

Initial Development of an Upper Plenum Structure Degradation Model for SCDAP/RELAP5, F. P. Griffin, ORNL/NRC/LTR-95/31, Oak Ridge National Laboratory, October 31, 1995.
1996
In-Vessel Core Degradation Code Validation Matrix, T. J. Haste, L. J. Ott, et al., OCDE/GD(96)14, Committee on the Safety of Nuclear Installations, OECD Nuclear Energy Agency, 1996.

“Comparison of HASCAL and MELCOR Source Terms,” J. J. Carbajo, Trans. Am. Nucl. Society, 75, pp. 271–73 (1996).

“MELCOR Calculated In-Containment Source Terms Using Different Release Models,” J. J. Carbajo, Trans. Am. Nucl. Society, 74, pp. 252–253 (1996).

Assessment of the RN and BH Packages of MELCOR 1.8.3, J. J. Carbajo, ORNL/NRC/LTR-96/01, Oak Ridge National Laboratory, 1996.

BWR Control Blade/Channel Box Model for SCDAP/RELAP5: Damage Progression Theory and User Guide, F. P. Griffin, ORNL/NRC/LTR-96/20, Oak Ridge National Laboratory, July 12, 1996.
1997
“Interpretation of the Results of the CORA-33 Dry Core Boiling Water Reactor Test,” L. J. Ott and S. Hagen, Nuclear Engineering and Design, 167, pp. 387–306 (1997).
“Severe Accident Sequences Analyzed for a Two-Loop PWR,” J. J. Carbajo, Trans. Am. Nucl. Society, 77, pp. 259–260 (1997).

“MELCOR-Calculated In-Containment Source Terms With and Without the BH Package,” J. J. Carbajo, Trans. Am. Nucl. Society, 77, pp. 290–292 (1997).

“Modeling the Chernobyl Accident with the HASCAL Code,” J. J. Carbajo, Trans. Am. Nucl. Society, 76, pp. 273–275 (1997).

“Modeling the TMI-2 Accident with the HASCAL Code,” J. J. Carbajo, Trans. Am. Nucl. Society, 76, pp. 275–277 (1997).

Simulation of the PHEBUS FPT-1 Test with the MELCOR Code, J. J. Carbajo, ORNL/NRC/LTR-97/03, Oak Ridge National Laboratory, 1997.

Validation of SCDAP/RELAP5 Against CORA BWR Test Results, F. P. Griffin, ORNL/NRC/LTR-97/1, Oak Ridge National Laboratory, February 28, 1997.

SCDAP/RELAP5 Mod 3.2 Simulations for the Browns Ferry BWR Design, F. P. Griffin, ORNL/NRC/LTR-97/27, Oak Ridge National Laboratory, December 16, 1997.

“Interpretation of the XR2-1 Experiment and Characteristics of the BWR Lower Plenum Debris Bed,” S. A. Hodge and L. J. Ott, presented at the Second International Conference on Advanced Reactor Safety (ARS’97), Orlando, FL, June 1–4, 1997.

“Applicability of BWR SFD Experiments and Codes for Advanced Core Component Designs,” L. J. Ott, presented at the Technical Session on Thermal Hydraulics of Severe Accidents at the ANS Winter Annual Meeting, Albuquerque, NM, November 1997.

SCDAP/RELAP5 MOD 3.2 Simulation for the Browns Ferry BWR Design, F. P. Griffin, ORNL/NRC/LTR-97/27, Oak Ridge National Laboratory, December 16, 1997.
1998
“Accident Sequences Simulated at the Juragua Nuclear Power Plant,” J. J. Carbajo, pp. 213–220 in Proceedings of the International Topical Meeting on Safety of Operating Reactors, San Francisco, CA, 1998.

“Severe Accident Sequences Simulated at a VVER 440-213,” J. J. Carbajo, Trans. Am. Nucl. Society, 78, pp. 198–199 (1998).

Technical Assistance in Review of Source Term-Related Issues of Advanced Reactors, E. C. Beahm, C. F. Weber, and T. A. Dillow, ORNL/TM-13144, Oak Ridge National Laboratory, 1998.

Iodine Volatility and pH Control in the AP-600 Reactor, C. F. Weber and E. C. Beahm, ORNL/TM-13555, Oak Ridge National Laboratory, 1998.

SCDAP/RELAP5 Modifications for New BWR Control Blade Designs, F. P. Griffin, ORNL/NRC/LTR-98/18, Oak Ridge National Laboratory, September 30, 1998.
1999
MELCOR Small Break LOCA Calculations, J. J. Carbajo, ORNL/NRC/LTR-98/21, Oak Ridge National Laboratory, 1999.

MELCOR DBA LOCA Calculations, J. J. Carbajo, ORNL/NRC/LTR-97/21, Oak Ridge National Laboratory, 1999.

“Severe Accident Sequences Simulated at the Grand Gulf Nuclear Station,” J. J. Carbajo, Trans. Am. Nucl. Society, 80, pp. 174–176 (1999).
Iodine Revolatizitation in a Grand Gulf LOCA, C. F. Weber and E. C. Beahm, ORNL/M-6544, Oak Ridge National Laboratory, 1999.

SCDAP/RELAP5 Lower Core Plate Model, E. W. Coryell and F. P. Griffin, INEEL/EXT-99-01029, Final Design Report, September 1999.
2000
Review of Grand Gulf Methodology and Calculations for pH Control and Iodine Volatility, C. F. Weber, Letter Report to USNRC, ORNL/NRC/LTR-00/07, Oak Ridge National Laboratory, 2000.
2005
Fission Product and Chemical Speciation Test Plan and Cost Estimate, D. D. Lee, J. L. Collins, B. E Lewis, J. L. Binder, S. C. Marschman, C. F. Weber, and R. N. Morris, Oak Ridge National Laboratory (February 28, 2005).

A Review and an Analysis of the ORNL and Chalk River Fission Product Release Tests and Their Relevance to Hypothetical Loss–of–Coolant Spent–Fuel Storage Pool Accidents, J. L. Collins and J. L. Binder, ORNL/TM-2004/20, Oak Ridge National Laboratory, April 2005.
2010
“Spray Sensitivity Study Performed with the MELCOR Code,” J. J. Carbajo and A. Drozd, Trans. Am. Nucl. Society, 102, pp. 324–326 (June 2010).

No comments:

Post a Comment