Simulation of the G3.1 Experiment at PKL Facility Using RELAP5/Mod3.3 /

Improvement of RELAP5/MOD3.3 reflood model based on the assessments against FLECHT-SEASET tests /
Published: (2011)

A study of RELAP5/MOD2 and RELAP5/MOD3 predictions of a small break LOCA simulation conducted at the ROSA-IV Large Scale Test Facility /
by: Sloan, Sandra Mernell, 1966-
Published: (1990)

Analysis of a loss of normal feedwater transient at the Ringhals-3 NPP using RELAP5/Mod3.3 /
by: Bánáti, J.
Published: (2010)

Preparation of radiological effluent technical specification for nuclear power plants : a guidance manual for users of standard technical specifications /
Published: (1987)

Preparation of radiological effluent technical specification for nuclear power plants : a guidance manual for users of standard technical specifications /
Published: (1978)

An assessment of the emergency core cooling systems rulemaking hearing : concluding statement - safety phase /
Published: (1974)

Loss of shutdown cooling during disassembly of high pressure safety injection system check valve.
Published: (1991)

RELAP5/MOD3.3 assessment by comparison with PKL III G3.1 experiment (small break in the main steam line) /
by: Zaruba, A., et al.
Published: (2014)

RELAP5/MOD3.3 analysis of event with actuation of safety injection system at the loss of external power /
by: Proŝek, A., et al.
Published: (2015)

Simulation of the experimental series F2.2 at PKL facility using RELAP5/MOD3.3 /
Published: (2012)

The technology for an ultrasafe reactor : hearing before the Subcommittee on Energy Research and Production of the Committee on Science and Technology, U.S. House of Representatives, ninety-ninth Congress, second session, September 17, 1986.
Published: (1987)

Clarification of TMI action plan item II.k.3.31.
Published: (1983)

Simulation of a small break loss of coolant accident conducted at the BETHSY Integral Test Facility /
by: Bott, Charles Patrick, 1968-
Published: (1992)

Reactor coolant and associated systems in nuclear power plants : a safety guide.
Published: (1986)

Validation of SCALE for high temperature gas-cooled reactor analysis /
Published: (2012)

Decay heat management practices during refueling outages.
Published: (1995)

Flow blockage of cooling water to safety system components.
Published: (1988)

Inadequate flow of essential service water to room coolers and heat exchangers for engineered safety-feature systems.
Published: (1990)

Loss of safety injection capability.
Published: (1986)

RELAP5/MOD3 assessment for calculation of safety and relief valve discharge piping hydrodynamic loads /
by: Stubbe, E. J. (Elie J.), et al.
Published: (1994)

Computer codes for small-break loss-of-coolant accidents : a Nordic assessment : final report of the NKA project SAK-3 /
by: Rathmann, Ole
Published: (1985)

ICAP assessment of RELAP5/MOD2, cycle 36.05 against LOFT small break experiment L3-7 /
by: Lee, Euy-Joon, et al.
Published: (1990)

Application of RELAP5/MOD3.1 code to the LOFT test L3-6 /
by: Pylev, S. S. (Stanislav Semenovich), et al.
Published: (1998)

Simulation of the F2.1 experiment at PKL facility using RELAP5/MOD3 /
Published: (2011)

Assessment of RELAP5/MOD3.2 with the LSTF experiment simulating a loss of residual heat removal event during mid-loop operation /
by: Seul, K. W.
Published: (1998)

Regulatory effectiveness of unresolved safety issue (USI) A-45, "shutdown decay heat removal requirements" /
Published: (2003)

The "calculated" loss-of-coolant accident : a review /
by: Ybarrondo, L. J.
Published: (1972)

Accident analysis for nuclear power plants with modular high temperature gas cooled reactors /
Published: (2008)

Customization of XTV graphics output in TRACE v5.0 patches 5, 4, & 3 / O. Zerkak and I. Clifford.
by: Zerkak, O., et al.
Published: (2019)

RELAP5/MOD3.3 analysis of steam generator tube rupture (SGTR) accident for NPP Krško /
by: Benčik, V., et al.
Published: (2014)

Post-test calculation of the PKL-2 test G7.1 using RELAP5/MOD3.3 /
by: Martinez-Quiroga, Victor, et al.
Published: (2019)

Advances in Nuclear Power Process Heat Applications.
by: Author), IAEA (Corporate
Published: (2012)

Analysis of the test OECD-PKL2 G7.1 with the thermal-hydraulic system code TRACE /
by: Belaïd, S., et al.
Published: (2013)

Application of a hydrological uncertainty methodology to nuclear reactor site evaluations /
by: Meyer, P. D., et al.
Published: (2016)

Simulation of the PKL-G7.1 experiment in a Westinghouse nuclear power plant using RELAP5/MOD3.3 /
by: Sánchez-Sáez, F.
Published: (2019)

Nuclear reactor safety heat transfer : presented at the Winter Annual Meeting of the American Society of Mechanical Engineers, Atlanta, Georgia, November 27-December 2, 1977 /
Published: (1977)