Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data

Sleepless

Bara bo Odar

Assessment of modernization and integration of BWR components and spatial kinetics in the TRAC-M, Version 3690, code

F. Odar

Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data

F. Odar

Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data

F. Odar

Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data

F Odar

Assessment of modernization and integration of BWR components and spatial kinetics in the TRAC-M, Version 3690, code

F Odar

Software quality assurance procedures for NRC thermal hydraulic codes

F Odar

Software quality assurance procedures for NRC thermal hydraulic codes

F. Odar

Techniken und Methoden der modernen Medizin

J. Odar

The development and assessment of TRACE model for Maanshan Nuclear Power Plant LOCA

Jong-Rong Wang

Jishin ni kakawaru kakuritsuronteki anzen hyōka shuhō no kairyō ni kansuru hōkokusho

EPR, l'impasse nucléaire

Frédéric Marillier

Jishin ni kakawaru kakuritsuronteki anzen hyōka shuhō no kairyō ni kansuru hōkokusho.

Akushidento manejimento chishiki bēsu seibi ni kansuru hōkokusho.

Akushidento manejimento chishiki bēsu seibi ni kansuru hōkokusho.

Roshin sonshō ni taisuru risuku jūyōdo ni motoduku haikan no bunrui to taiōsuru kensa ni kansuru kentō (PWR).

PWR no jiko shīkensu no shikaiseki

Final report on the PMK-2 projects.

The 2007-2012 Outlook for Gravity and Pressurized 1.6 G.p.f. Water Closet Bowls Close Coupled for Use with Flush Tank with 17 Inches from Floor to Rim ... Flush Tank Unless Integral in Japan

Philip M. Parker

The 2007-2012 Outlook for Hand-Portable Foam, Halogenated Agent, and Pressurized Water Fire Extinguishers in India

Philip M. Parker

The 2007-2012 World Outlook for Hand-Portable Foam, Halogenated Agent, and Pressurized Water Fire Extinguishers

Philip M. Parker

Kiki furajiriti hyōka (PWR puranto).

Study of hydride blisters grown on Zr-2.5Nb pressure tube spool piece under simulated condition of in-reactor pressure and temperature

Istorii︠a︡ sozdanii︠a︡ reaktornykh ustanovok VVĖR dli︠a︡ atomnykh ėlektrostant︠s︡iĭ

V. P. Denisov

Draft regulatory guide DG-1107

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research.

Approach for estimating the frequencies of various containment failure modes and bypass events

W. T. Pratt

Behavior of PWR reactor coolant system components, other than steam generator tubes, under severe accident conditions

S. Majumdar

U.S. operating experience with thermally treated Alloy 600 steam generator tubes

K. J. Karwoski

GSI-191

Validation of SCALE sequence CSAS26 for criticality safety analysis of VVER and RBMK fuel designs

GSI-191

D. V. Rao

GSI-191 technical assessment

PWR and BWR pressure vessel fluence calculation benchmark problems and solutions

Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data

F. Odar

Incorporating aging effects into probabilistic risk assessment

TRAC-M/FORTRAN 90 (version 3.0) user manual

Voltage-based alternative repair criteria

Voltage-based alternative repair criteria

Voltage-based alternative repair criteria

Assessment of RELAP5/MOD3.2-NPA3.4 against an inadvertent closure of all three MSIV's in VANDELLOS-II nuclear power plant

C. Llopis

Standard technical specifications

U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.

Analyses of KS test data on the heated rod bundle temperature behavior in RBMK-1500 core model under stop and recovery flow using RELAP5/MOD3.2 and RELAP5/MOD3.2.2 GAMMA

V. A Vinogradov

Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data

F. Odar

Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data

F Odar

Voltage-based alternative repair criteria

Voltage-based alternative repair criteria

Simulation of LOCA 6" and LOCA 2" transients in the RHR of a PWR under low power conditions using RELAP5/MOD3.2

Assessment of RELAP5/MOD3.2 against a main steam isolation valve closure at TRILLO I Nuclear Power Plant

A. de Lucas

Development of a program "BFQ/VERI" to simulate vapour pull through and liquid entrainment under stratified flow condition

RELAP5/MOD3.2 validation using BETHSY test 6.9a

S Bouabdallah

RELAP5/MOD3.2 validation using BETHSY test 6.9a

S. Bouabdallah

Simulation of LOCA 6" and LOCA 2" transients in the RHR of a PWR under low power conditions using RELAP5/MOD3.2

RELAP5/MOD3.2 validation using BETHSY test 6.9a

S. Bouabdallah

RELAP5/MOD3.2 validation using BETHSY test 6.9a

S. Bouabdallah

RELAP5/MOD3.2 post test analysis and accuracy quantification of SPES test SP-SB-04

F. D'Auria

Analysis of the critical flow model in TRAC-BF1

M Gomez

RELAP5/MOD.2 post test analysis and accuracy quantification of Lobi test BL-34

F. D'Auria

Nuclear fuel of pressurized water reactors and fast neutron reactors

Test LOBI-BL06

T. Fiore

RELAP5/MOD3.2 assessment using GERDA small break test, 1605AA

W. Tietsch

RELAP5/MOD3.2 post test analysis and accuracy quantification of Lobi test BL-44

F D'Auria

RELAP5/MOD3.2 post test analysis and accuracy quantification of Lobi test BL-44

F. D'Auria

RELAP5/MOD3.2 post test analysis and accuracy quantification of SPES test SP-SB-03

F. D'Auria

RELAP5/MOD3.2 post test analysis and accuracy quantification of Lobi test BL-44

F. D'Auria

Test LOBI-BL06

T Fiore

Analysis of the critical flow model in TRAC-BF1

M Gómez

RELAP5/MOD3.2 post test analysis and accuracy quantification of SPES test SP-SB-03

F D'Auria

RELAP5/MOD3.2 assessment using GERDA small break test, 1605AA

W Tietsch

Pressurizer surge line Counter Current Flow Limitation during AP600 Mode 5 Cold Shutdown

Sarah E. Colpo