Published: 1997-09-09

Articles

The Method of Subgroups in the Calculation of Shielded Cross-Sections for WIMS-AECL

H. Chen, P.J. Laughton
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On the Preparation of Library Data for an Implementation of the Method of Subgroups in WIMS-AECL

P.J. Laughton
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Calculation of the Volumetric Heat Generation Rate Radial Profile in a Cylindrical Fuel Rod

V.I. Arimescu, M. Couture, M.E. Klein, A.F. Williams
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An Acceleration Scheme for the Multigroup Sn Equations with Fission and Thermal Upscatter

B.T. Adams, J.E. Morel
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History-Based Calculations Using WTMS-AECL in RFSP

B. Arsenault, James V. Donnelly, D.A. Jenkins
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CATHENA Study of Two phase Water Hammer Inter-Peak Timing

T.G. Beuthe
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Preliminary Investigation of the Solution of Sparse Matrices in CATHENA

T.G. Beuthe, J.B. Hedley
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Refill Effectiveness using High-Boiling Point Emergency Coolant

G.R. McGee
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Darlington NGS A SGECS Condensation Induced Waterhammer Analysis using the TUF Code

C.W. So, P.L. Chang, D.G. Meranda
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A Three-Dimensional Modeling Analysis of CANDU 9 End-Shield Cooling Effects with CFX-4: A Preliminary Report

Y. Zhuang, A.O. Banas, R.Q-N. Zhou
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Analysis of the Fast Neutron Spectrum Inside the Materials Test Bundles in the NRU Loops

T.C. Leung
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Study of Nuclide Field Behaviour in Reactor with Continuously Changing Core Parameters

V.A. Khotylev, J.E. Hoogenboom, D.R. Kingdon, A.A. Harms
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Validation and Benchmarking of the HELIOSRAMONA Model of Dodewaard

B.T. Adams, Lars Moberg, Rudi Stammler, Aldo Ferri, Flavio Guist
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The Placement of Lateral Region in 1D MTR Fuel Lattice Cell Models

S. Day, W.J. Garland
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Photoneutron Source Strength Studies in Pickering Reactors

Mark W. Hersey, Ka Fai So, Charles G. Olive, Fred C. Shanes
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An Analysis of the Effect of Correlation of Response in the System Reliability Analysis in Seismic PSA of Nuclear Power Plants

Yuichi Watanabe, Tetsukuni Oikawa, Ken Muramatsu
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Development of a Computational Framework of Uncertainty Analysis for Level 2 PSA to Consider both Aleatory and Episternic Uncertainties

Ken Muramatsu, Masashi Himi, Katsuhiro Amagasa
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Optimisation of Operating Envelope and Safety Margin Using PRA Methods

Cristian Stoica, C. Keith Scott
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A Spatial Convergence Study for Flows and Hydrogen Distribution in the CANDU 6 Reactor Vault following a Loss of Coolant Accident using the Gothic Containment Analysis Code

Tuan H. Nguyen, W.M. Collins
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Assessment of Refuelling Effects in High Power Channel on Fission Product Releases for Following an End-Fitting Failure

Kang Moon Lee, Ki Man Nho
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Fission Product Inventory and Distribution at Point Lepreau Generating Station

R.A. Gibb, P.J. Reid, T.J. Chapman
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NUCIRC Simulation of a Large Header for the CANDU Type Nuclear Reactor Heat Transport System

A. Kwan, R. Dam, M. Soulard, G.D. Harvel
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Modelling Techniques for Vanadium Detector Compensation

J.C. Handbury, C.W. Newman
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Dynamic Modelling for Shutdown-System-1 Depth Analysis

D.A. Jenkins
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Reactor Physics Simulation of the MNR January 1994 Fuelling Incident

Hassan S. Basha
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Effect of Pressure Tube Creep on the Void Coefficient of a CANDU Fuel Bundle

M.S. Milgram
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System for Neutronic and Thermalhydraulic Diagnostics of Channel Type Reactor Parameters

V.A. Khotylev, N.V. Schukin, A.V. Filatov, S.D. Romanin, A.A. Semenov, Yu.B. Chizhevskiy, V.S. Sidorov
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A Scaling Law Verification of the Delft Simulated Reactor (DESIRE)

B.T. Adams
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Development of a Model for Calculating Sheath Thermocouple Finning Losses for Application in In-Reactor Severe Fuel Damage Tests

L.C. Walters, J.W. DeVaal, N.K. Popov
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Transient Melting and Re-Solidification Model of CANDU Core Debris in Severe Accidents

J.T. Rogers, M. Lamari
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CATHENA Model of the CANDU 9 End Shield Cooling System in the Event of Loss of Forced Circulation

N.K. Popov, L.A. Morris, D.N. Padhi
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Thermalhydraulics Response of a Conceptual CANDU Reactor with a highly Advanced Core subject to various Loss-of-Coolant Accidents

J. Vechiarelli, R.F. Dam, K.F. Hau, F.J. Doria
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Safety Analysis Uncertainty Related to Design and Manufacturing Quality Assurance

E.N. Sokolov
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A Generalized Prediction Method for Single-Phase Pressure Drop in a String of Alinged CANDU-type Bundles

L.K.H. Leung, G. Hotte
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Critical Heat Flux Evaluation of a Conceptual Fuel Bundle Design with the ASSERT Subchannel Code

R.F. Dam, K.F. Hau, J. Vecchiarelli, N. Lee
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Simulation of CANDU Bundle Cross-Sectional Averaged Actual Flow Quality and Void in One-Dimensional Two-Phase Flow Models

G. Hotte, M. Soulard, L. Leung
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Modelling Methodology for Obstructions in the sub-channel Code ASSERT IV

P. White, R.F. Dam, M.R. Soulard, N. Lee
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High Heat-Flux Transfer and Pressure Drop Correlations for Reactor Thermalhydraulic Simulations

David R. Novog, J.S. Chang
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