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

Jong-Rong Wang

EPR, l'impasse nucléaire

Frédéric Marillier

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 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

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

K. J. Karwoski

GSI-191 technical assessment

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

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 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

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

W. Tietsch

Final safety evaluation report related to certification of the AP600 standard design docket no. 52-003

Assessment of RELAP5/MOD3.2 for steam condensation experiments in the presence of noncondensibles in a vertical tube of PCCS

Study of transients related to AMSAC actuation, sensitivity analysis

Assessment of RELAP5/MOD3.2 using LOFT large break LOCA test, LP-02-6

Assessment of RELAP5/MOD3.2 with the semiscale natural circulation experiment, S-NC-8B

Action requirements for AFW system failures

Assessment of RELAP5/MOD3.2 with the semiscale natural circulation experiment, S-NC-8B

Low pCo₂ concentrator development

F. H. Schubert

U.S. Department of Energy reactor core protection evaluation methodology for fires at Soviet-designed RBMK and VVER nuclear power plants

Thermal analysis of pressurized water reactors

L. S. Tong

The basic objective of this book is to present the principles underlying the thermal and hydraulic design of pressurized water reactors.

Applications of reliability degradation analysis

W. E. Vesley

U.S. Department of Energy reactor core protection evaluation methodology for fires at Soviet-designed RBMK and VVER nuclear power plants

Steam generator tube failures

U.S. Department of Energy reactor core protection evaluation methodology for fires at Soviet-designed RBMK and VVER nuclear power plants

Two-phase fluid break flow measurements and scaling in the advanced plant experiment (APEX)

David Alan Pimentel

Estimating pressurized water reactor decommissioning costs

M. C. Bierschbach

Effect of aging on the PWR chemical and volume control system

E. J. Grove

Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2

E. W. Coryell

Natural circulation phenomena in nuclear reactor systems

Franklin Cheung

Voltage-based interim plugging criteria for steam generator tubes

U.S. Nuclear Regulatory Commission. IPC Task Group.

XSOR codes users manual

Hong-Nian J. Jow

2D/3D program work summary report

Pressurized-water reactor internals aging degradation study

K. H. Luk

Integrated systems analysis of the PIUS reactor

Assessment of RELAP5/MOD2 against a pressurizer spray valve inadverted fully opening transient and recovery by natural circulation in Jose Cabrera nuclear station

R. Arroyo

Auxiliary feedwater system risk-based inspection guide for the Beaver Valley, units 1 and 2, nuclear power plants

Numerics and implementation of the UK horizontal stratification entrainment off-take model into RELAP5MOD3

W. M Bryce

Voltage-based interim plugging criteria for steam generator tubes

U.S. Nuclear Regulatory Commission. IPC Task Group

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