Cover Image

Uncertainty analysis for Maanshan LBLOCA by TRACE and DAKOTA /

Bibliographic Details
Main Author:	Wang, Jong-Rong (Author)
Format:	Government Document eBook
Language:	English
Published:	Washington, DC : U.S. Nuclear Regulatory Commission, 2014.
Series:	International agreement report ; NUREG/IA-0448.
Subjects:	Nuclear reactor accidents > Analysis. Nuclear reactor accidents > Computer simulation. Pressurized water reactors > Loss of coolant > Analysis.
Online Access:	https://purl.fdlp.gov/GPO/gpo51870 http://purl.fdlp.gov/GPO/gpo51870

Similar Items

Fuel rod behavior and uncertainty analysis by FRAPTRAN/TRACE/DAKOTA code in maanshan LBLOCA /
by: Shih, Chunkuan
Published: (2016)

TRACE/RELAP5 comparative calculations for double-ended LBLOCA and SBO /
by: Legan, S.
Published: (2016)

Assessment of LONF ATWS for Maanshan PWR using TRACE code /
by: Wang, Jong-Rong
Published: (2014)

The development and assessment of TRACE model for Maanshan Nuclear Power Plant LOCA /
by: Wang, Jong-Rong
Published: (2012)

Performing uncertainty analysis of IIST facility SBLOCA by TRACE and DAKOTA /
by: Wang, Jong-Rong
Published: (2013)

BEPU analysis and benchmark with IIST 2% SBLOCA experiment using TRACE/DAKOTA /
by: Shih, Chunkuan
Published: (2015)

Unanalyzed post-LOCA release paths.
Published: (1986)

Developmental assessment of RELAP5/MOD3.1 with separate-effect and integral test experiments : model changes and options /
by: Analytis, G. Th. (Gerassimos Th.)
Published: (1998)

Result of BETHSY test 9.1.b using RELAP5/MOD3 /
by: Petelin, S.
Published: (1998)

Fuel fragmentation, relocation, and dispersal during the loss-of-coolant accident /
by: Raynaud, Patrick A. C.
Published: (2012)

Sensitivity study of the DEG LBLOCA transient on the counter-current flow limitation by using TRACE /
by: Wang, Jong-Rong
Published: (2014)

Audit of small break LOCA guidelines.
Published: (1979)

RELAP5/MOD2 analysis of LOFT experiment L9-4 /
by: Keevill, M. B.
Published: (1992)

TRACE simulation of SBO accident and mitigation strategy in Maanshan PWR /
by: Wang, Jong-Rong
Published: (2013)

Qualification of the three-dimensional thermal hydraulic model of TRACE using plant data /
by: Sánchez-Espinoza, Victor Hugo
Published: (2011)

LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3 /
by: Sánchez, J. I., et al.
Published: (2001)

Development of a Vandellos II NPP model using the TRACE code : application to an actual transient of main coolant pumps trip and start-up /
Published: (2011)

Loss of reactor coolant inventory and associated potential for loss of emergency mitigation functions while in a shutdown condition.
Published: (1998)

Potential for degradation of the emergency core cooling system and the containment spray system after a loss-of-coolant accident because of construction and protective coating deficiencies and foreign material in containment.
Published: (1998)

Fuel behavior under abnormal conditions /
by: Meyer, R. O. (Ralph O.)
Published: (2013)

Potential for loss of shutdown cooling while at low reactor coolant levels.
Published: (1990)

GSI-191 PWR sump screen blockage chemical effects tests : thermodynamic simulations /
Published: (2006)

Depressurization of reactor coolant systems in pressurized-water reactors.
Published: (1987)

Possible sudden loss of RCS inventory during low coolant level operation.
Published: (1988)

Loss of decay heat removal--10 CFR 50.54(f).
Published: (1988)

Inter-system discharge of reactor coolant.
Published: (1990)

Core exit temperature response during an SBLOCA event in the Ascó NPP /
by: Freixa, J., et al.
Published: (2018)

Loss of residual heat removal (RHR) while the reactor coolant system (RCS) is partially filled.
Published: (1987)

Spent fuel pool project phase I : pre-ignition and ignition testing of a single commercial 17x17 pressurized water reactor spent fuel assembly under complete loss of coolant accident conditions /
by: Durbin, S. G.
Published: (2016)

Spent fuel pool project phase II : pre-ignition and ignition testing of a 1x4 commercial 17x17 pressurized water reactor spent fuel assemblies under complete loss of coolant accident conditions /
by: Durbin, S. G.
Published: (2016)

Post-test thermal-hydraulic analysis of PKL tests F1.1 and F1.2 /
by: Freixa, J.
Published: (2014)

Reporting of errors and changes in large-break loss-of-coolant accident evaluation models of fuel vendors and compliance with 10 CFR 50.46(a)(3).
Published: (1997)

Loss of reactor coolant inventory and potential loss of emergency mitigation functions while in a shutdown condition.
Published: (1995)

RELAP5 assessment against PACTEL experimental data /
by: Parzer, Iztok, 1960-, et al.
Published: (1998)

Assessment of current test methods for post-LOCA cladding behavior /
by: Billone, Michael C.
Published: (2012)

Ultrasonic phased array assessment of the interference fit and leak path of the North Anna Unit 2 control rod drive mechanism nozzle 63 with destructive validation /
by: Crawford, S. L.
Published: (2012)

Modeling and small-break loss-of-coolant accident analysis of Comanche Peak Steam Electric Station using RELAP5/MOD2 /
by: Salim, Parvez, 1961-
Published: (1989)

RELAP5/MOD2 assessment, OECD-LOFT small break experiment LP-SB-03 /
by: Güntay, Salih
Published: (1990)

Investigations of the VVER-1000 Coolant Transient Benchmark Phase 1 with the coupled code system RELAP5/PARCS /
Published: (2009)

Aluminum chemistry in a prototypical post-loss-of-coolant-accident, pressurized-water-reactor containment environment /
Published: (2006)