-
1
-
-
33845751048
-
The Liquid Salt Pebble Bed Reactor: A New High-Temperature Nuclear Reactor
-
Delft University of Technology Nov
-
S. J. DE ZWAAN, "The Liquid Salt Pebble Bed Reactor: A New High-Temperature Nuclear Reactor," PNR-131-2005-008, Delft University of Technology (Nov. 2005).
-
(2005)
PNR-131-2005-008
-
-
DE ZWAAN, S.J.1
-
2
-
-
51049112577
-
-
Jan
-
J. Nucl. Mater., 360 (Jan. 2007).
-
(2007)
J. Nucl. Mater
, vol.360
-
-
-
3
-
-
0345172386
-
Molten-Salt-Cooled Advanced High-Temperature Reactor for Production of Hydrogen and Electricity
-
C. W. FORSBERG, P. F. PETERSON, and P. S. PICKARD, "Molten-Salt-Cooled Advanced High-Temperature Reactor for Production of Hydrogen and Electricity," Nucl. Technol., 144, 289 (2003).
-
(2003)
Nucl. Technol
, vol.144
, pp. 289
-
-
FORSBERG, C.W.1
PETERSON, P.F.2
PICKARD, P.S.3
-
4
-
-
51049105381
-
-
N. E. TODREAS and P. HEJZLAR, Flexible Conversion Ratio Fast Reactor Systems Evaluation, Nuclear Energy Research Initiative Project 06-40, U.S. Department of Energy Office of Nuclear Energy, available on the Internet at www.ne.doe.gov/neri/2006awards /NERI%2006-040.pdf (2006).
-
N. E. TODREAS and P. HEJZLAR, "Flexible Conversion Ratio Fast Reactor Systems Evaluation," Nuclear Energy Research Initiative Project 06-40, U.S. Department of Energy Office of Nuclear Energy, available on the Internet at www.ne.doe.gov/neri/2006awards /NERI%2006-040.pdf (2006).
-
-
-
-
5
-
-
51049119504
-
Industrial Use of Molten Nitrate /Nitrite Salts
-
Sandia National Laboratory
-
R. W. CARLING and R. W. MAR, "Industrial Use of Molten Nitrate /Nitrite Salts," SAND-81-8020, Sandia National Laboratory (1981).
-
(1981)
SAND-81-8020
-
-
CARLING, R.W.1
MAR, R.W.2
-
6
-
-
51049103079
-
Molten Salt for Heat Transfer
-
May 27
-
H. P. VOZNICK and V. W. UHL, "Molten Salt for Heat Transfer," Chem. Eng., p. 129 (May 27, 1963).
-
(1963)
Chem. Eng
, pp. 129
-
-
VOZNICK, H.P.1
UHL, V.W.2
-
7
-
-
0346471536
-
A New Heat Transfer Medium for High Temperatures
-
W. E. KIRST, W. M. NAGLE, and J. B. CASTNER, "A New Heat Transfer Medium for High Temperatures," Trans. Am. Inst. Chem. Eng., 36, 371 (1940).
-
(1940)
Trans. Am. Inst. Chem. Eng
, vol.36
, pp. 371
-
-
KIRST, W.E.1
NAGLE, W.M.2
CASTNER, J.B.3
-
8
-
-
51049101551
-
-
D. F. WILLIAMS, K. T. CLARNO, and L. M. TOTH, Assessment of Candidate Liquid-Salt Coolants for the Advanced High-Temperature Reactor (AHTR), ORNL/TM-2006/12, Oak Ridge National Laboratory, available on the Internet at http://www.osti.gov/energycitations/servlets/purl/ 885975-9IC4H7/ (2006).
-
D. F. WILLIAMS, K. T. CLARNO, and L. M. TOTH, "Assessment of Candidate Liquid-Salt Coolants for the Advanced High-Temperature Reactor (AHTR)," ORNL/TM-2006/12, Oak Ridge National Laboratory, available on the Internet at http://www.osti.gov/energycitations/servlets/purl/ 885975-9IC4H7/ (2006).
-
-
-
-
9
-
-
51049122694
-
-
D. F. WILLIAMS, Assessment of Candidate Molten Salt Coolants for the NGNP/NHI Heat-Transfer Loop, ORNL/TM-2006/69, Oak Ridge National Laboratory, available on the Internet at http://www.ornl.gov /~-webworks/cppr/y2006/rpt/124838.pdf (2006).
-
D. F. WILLIAMS, "Assessment of Candidate Molten Salt Coolants for the NGNP/NHI Heat-Transfer Loop," ORNL/TM-2006/69, Oak Ridge National Laboratory, available on the Internet at http://www.ornl.gov /~-webworks/cppr/y2006/rpt/124838.pdf (2006).
-
-
-
-
10
-
-
0014735964
-
Molten Salt Reactor Chemistry
-
W. R. GRIMES, "Molten Salt Reactor Chemistry," Nucl. Appl. Technol., 8, 2, 137 (1970).
-
(1970)
Nucl. Appl. Technol
, vol.8
, Issue.2
, pp. 137
-
-
GRIMES, W.R.1
-
12
-
-
0004281313
-
Comparison of Coolants
-
Sec. 9-3, Chap. 6.5, p, H. ETHERINGTON, Ed
-
C. F. BONILLA, "Comparison of Coolants," Nuclear Engineering Handbook, Sec. 9-3, Chap. 6.5, p. 9, H. ETHERINGTON, Ed. (1958).
-
(1958)
Nuclear Engineering Handbook
, pp. 9
-
-
BONILLA, C.F.1
-
13
-
-
51049111059
-
-
SCALE: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluations, ORNL/TM-2005/39, Version 5.1, Vols. I, II, and III, Oak Ridge National Laboratory Nov. 2006
-
"SCALE: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluations," ORNL/TM-2005/39, Version 5.1, Vols. I, II, and III, Oak Ridge National Laboratory (Nov. 2006).
-
-
-
-
14
-
-
51049114138
-
-
D. T. INGERSOLL, Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor LS-VHTR, ORNL/ TM-2005/218, Oak Ridge National Laboratory, available on the Internet at http://www.ornl.gov/~webworks/cppr/y2001/rpt/124473.pdf (2005).
-
D. T. INGERSOLL, "Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor LS-VHTR)," ORNL/ TM-2005/218, Oak Ridge National Laboratory, available on the Internet at http://www.ornl.gov/~webworks/cppr/y2001/rpt/124473.pdf (2005).
-
-
-
-
15
-
-
0003765311
-
-
PU/NE-98-26, Purdue University Sep
-
H. G. JOO, D. BARBER, G. JIANG, and T. J. DOWNAR, "PARCS, A Multi-Dimensional Two-Group Reactor Kinetics Code Based on the Nonlinear Analytic Nodal Method," PU/NE-98-26, Purdue University (Sep. 1998).
-
(1998)
PARCS, A Multi-Dimensional Two-Group Reactor Kinetics Code Based on the Nonlinear Analytic Nodal Method
-
-
JOO, H.G.1
BARBER, D.2
JIANG, G.3
DOWNAR, T.J.4
-
16
-
-
44349107657
-
NESTLE: A Few-Group Neutron Diffusion Equation Solver Utilizing the Nodal Expansion Method for Eigenvalue, Adjoint, Fixed-Source Steady State and Transient Problems
-
Idaho National Engineering Laboratory
-
P. J. TURINSKY et al., "NESTLE: A Few-Group Neutron Diffusion Equation Solver Utilizing the Nodal Expansion Method for Eigenvalue, Adjoint, Fixed-Source Steady State and Transient Problems," EGG-NRE-11406, Idaho National Engineering Laboratory (1994).
-
(1994)
EGG-NRE-11406
-
-
TURINSKY, P.J.1
-
17
-
-
0032755216
-
Preparation and Physical Characteristics of a Lithium-Beryllium-Substituted Fluoroapatite
-
D. LEXA, "Preparation and Physical Characteristics of a Lithium-Beryllium-Substituted Fluoroapatite," Metall. Mater. Trans., 30A, 147 (1999).
-
(1999)
Metall. Mater. Trans
, vol.30 A
, pp. 147
-
-
LEXA, D.1
-
18
-
-
51049085904
-
-
C. F. WEAVER and R. G. ROSS, High-Temperature Fuel Salt-Graphite Compatibility Experiment, Sec. 10.3 in MSR Program Semiannual Progress Report for Period Ending August 31, 1968, ORNL-4344, Oak Ridge National Laboratory (1969).
-
C. F. WEAVER and R. G. ROSS, "High-Temperature Fuel Salt-Graphite Compatibility Experiment," Sec. 10.3 in "MSR Program Semiannual Progress Report for Period Ending August 31, 1968," ORNL-4344, Oak Ridge National Laboratory (1969).
-
-
-
-
20
-
-
84884759948
-
Redox Potential of Novel Electrochemical Buffers Useful for Corrosion Prevention in Molten Fluorides
-
Philadelphia, Pennsylvania, May 12-17, Electrochemical Society
-
G. D. DELCUL et al., "Redox Potential of Novel Electrochemical Buffers Useful for Corrosion Prevention in Molten Fluorides," Proc. 201st Electrochemical Society Mtg., Molten Salts 13, Philadelphia, Pennsylvania, May 12-17, 2002, p. 431, Electrochemical Society (2002).
-
(2002)
Proc. 201st Electrochemical Society Mtg., Molten Salts 13
, pp. 431
-
-
DELCUL, G.D.1
-
21
-
-
33845736907
-
A Review of Possible Choices for Secondary Coolants for Molten Salt Reactors
-
Oak Ridge National Laboratory
-
J. P. SANDERS, "A Review of Possible Choices for Secondary Coolants for Molten Salt Reactors," ORNL CF-71-8-10, Oak Ridge National Laboratory (1971).
-
(1971)
ORNL CF-71-8-10
-
-
SANDERS, J.P.1
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