Quite a productive group !
ORNL LIGHT WATER REACTOR SAFETY PUBLICATIONS –
AN INCOMPLETE BIBLIOGRAPHY COVERING THE PERIOD 1975 – 2010
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).
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).
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.
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.
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.
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.
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.
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).
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).
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.
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.
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).
