ANS / Publications / Journals / Nuclear Technology

An Assessment of Large-Eddy Simulation Toward Thermal Fatigue Prediction

A. K. Kuczaj, E. M. J. Komen

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 2-15

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9441

Thermal-Fluid Characterizations of ZnO and SiC Nanofluids for Advanced Nuclear Power Plants

In Cheol Bang, Ji Hyun Kim

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 16-27

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9442

Condensation Correlation for a Vertical Passive Condenser System

Shripad T. Revankar, Seungmin Oh, Wenzhong Zhou, Gavin Henderson

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 28-39

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9443

Summary for Three Different Validation Cases of Coolant Flow in Supercritical Water Test Sections with the CFD Code ANSYS CFX 11.0

Attila Kiss, Attila Aszódi

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 40-53

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9444

Parametric Sensitivity Study on the Reflooding Models of the MARS Code Based on 6 × 6 Rod Bundle Test Results

Ki-Yong Choi, Seok Cho, Hyoung-Kyu Cho, Chul-Hwa Song

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 54-67

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9445

The Enhancements and Testing for the MCNPX 2.6.0 Depletion Capability

Michael L. Fensin, John S. Hendricks, Samim Anghaie

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 68-79

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fuel Cycle and Management / dx.doi.org/10.13182/NT10-2

Development of an Automated Testing System for Verification and Validation of Nuclear Data and Simulation Code

Brian S. Triplett, Samim Anghaie, Morgan C. White

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 80-89

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Radiation Measurements and Instrumentation / dx.doi.org/10.13182/NT10-A9447

Heat Transfer in Intermediate Heat Exchanger Under Low Flow Rate Conditions

Hiroyasu Mochizuki

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 90-99

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9448

An Integral Effect Test on the Reflood Period of a Large-Break LOCA for the APR1400 Using ATLAS

Hyun Sik Park, Ki Yong Choi, Seok Cho, Kyoung Ho Kang, Nam Hyun Choi, Dong Jin Euh, Yeon Sik Kim, Won Pil Baek

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 100-113

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9449

Optimal Geometric Configuration for a Natural Circulation Loop

Jin Ho Song

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 114-122

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9450

Large Break Loss-of-Coolant Accident Analysis of VVER-1000 Reactor Using CATHARE Code

Luben Sabotinov, Abhishek Srivastava

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 123-132

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9451

Development of Advanced Loop-Type Fast Reactor in Japan

Shoji Kotake, Yoshihiko Sakamoto, Takatsugu Mihara, Shigenobu Kubo, Nariaki Uto, Yoshio Kamishima, Kazumi Aoto, Mikio Toda

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 133-147

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fission Reactors / dx.doi.org/10.13182/NT10-7

System Design and Analysis of a 900-MW(thermal) Lead-Cooled Fast Reactor

Sang Ji Kim, Yonghee Kim, Sergi Hong, Chung Ho Cho, Jae-Hyuk Eoh, Jong Bum Kim, Myung Hwan Wi, Kwi Seok Ha, Eui Kwang Kim

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 148-158

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fission Reactors / dx.doi.org/10.13182/NT10-A9453

Technological Feasibility of Two-Loop Cooling System in JSFR

Hidemasa Yamano, Shigenobu Kubo, Ken-Ichi Kurisaka, Yoshio Shimakawa, Hiromi Sago

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 159-169

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT09-6

Minor Actinide-Bearing Oxide Fuel Core Design Study for the JSFR

Masayuki Naganuma, Takashi Ogawa, Shigeo Ohki, Tomoyasu Mizuno, Shoji Kotake

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 170-180

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fuel Cycle and Management / dx.doi.org/10.13182/NT10-A9455

Development of Passive Shutdown System for SFR

Shigeyuki Nakanishi, Takusaburo Hosoya, Shigenobu Kubo, Shoji Kotake, Misao Takamatsu, Takafumi Aoyama, Iwao Ikarimoto, Jungo Kato, Yoshio Shimakawa, Kiyoshi Harada

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 181-188

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Nuclear Plant Operations and Control / dx.doi.org/10.13182/NT10-A9456

European Experiments on 2-D Molten Core Concrete Interaction: HECLA and VULCANO

Christophe Journeau, Jean Michel Bonnet, Eric Boccaccio, Pascal Piluso, Jose Monerris, Michel Breton, Gerald Fritz, Tuomo Sevón, Pekka H. Pankakoski, Stefan Holmström, Jouko Virta

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 189-200

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9457

Transient Conduction Heat Transfer Modeling in Concrete for the Simulation of Long-Term Phase of Molten Core-Concrete Interaction

B. Tourniaire, B. Spindler, M. Guillaumé

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 201-209

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-6

Interaction Between Molten Corium UO_2+X-ZrO₂-FeO_y and VVER Vessel Steel

S. V. Bechta, V. S. Granovsky, V. B. Khabensky, E. V. Krushinov, S. A. Vitol, A. A. Sulatsky, V. V. Gusarov, V. I. Almiashev, D. B. Lopukh, D. Bottomley, M. Fischer, P. Piluso, A. Miassoedov, W. Tromm, E. Altstadt, F. Fichot, O. Kymalainen

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 210-218

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Materials for Nuclear Systems / dx.doi.org/10.13182/NT10-A9459

The DEFOR-S Experimental Study of Debris Formation with Corium Simulant Materials

Pavel Kudinov, Aram Karbojian, Weimin Ma, Truc-Nam Dinh

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 219-230

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics / dx.doi.org/10.13182/NT10-A9460

Development of the Fission Product Release Analysis Code COPA-FPREL

Young Min Kim, Moon Sung Cho

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 231-243

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fuel Cycle and Management / dx.doi.org/10.13182/NT10-A9461

Analysis and Justification of MAAP4.0.7 for PRA Level 1 Mission Success Criteria

Jennifer S. Butler, Darvin Kapitz, Robert P. Martin, Farrokh Seifaee, Ramu K. Sundaram

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 244-260

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Reactor Safety / dx.doi.org/10.13182/NT10-A9462

Corrosion Studies of Candidate Materials for European HPLWR

Sami Penttilä, Aki Toivonen, Liisa Heikinheimo, Radek Novotny

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 261-271

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Materials for Nuclear Systems / dx.doi.org/10.13182/NT10-A9463

Comparison of the High-Temperature Steam Oxidation Kinetics of Advanced Cladding Materials

M. Grosse

Nuclear Technology / Volume 170 / Number 1 / April 2010 / Pages 272-279

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Materials for Nuclear Systems / dx.doi.org/10.13182/NT10-A9464