• Publications
  • Influence
RELAP5/MOD3 Code Manual. Volume 5, User`s Guidelines
The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant
Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios
This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and
RELAP5 thermal-hydraulic analyses of pressurized thermal shock sequences for the H. B. Robinson Unit 2 pressurized water reactor
Thermal-hydraulic analyses of fourteen hypothetical pressurized thermal shock (PTS) scenarios for the H.B. Robinson, Unit 2 pressurized water reactor were performed using the RELAP5 computer code.
RELAP/MOD3 code manual: User`s guidelines. Volume 5, Revision 1
The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor
Simulation of the chernobyl accident
Abstract An analysis of the April 26, 1986 accident at the Chernobyl-4 nuclear power plant in the Soviet Union is presented. The peak calculated core power during the accident was 550 000 MW t . The
SCDAP/RELAP5 THERMAL-HYDRAULIC EVALUATIONS OF THE POTENTIAL FOR CONTAINMENT BYPASS DURING EXTENDED STATION BLACKOUT SEVERE ACCIDENT SEQUENCES IN A WESTINGHOUSE FOUR-LOOP PWR
................................................................................................................................. iii
RELAP4/MOD6 analysis of forced- and gravity-feed reflood tests
The RELAP4/MOD6 computer code is used for the analysis of the reactor core heat transfer during the reflooding phase of a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor
Break spectrum analyses for small break loss of coolant accidents in a RESAR-3S Plant
A series of thermal-hydraulic analyses were performed to investigate phenomena occurring during small break loss-of-coolant-accident (LOCA) sequences in a RESAR-3S pressurized water reactor. The
RELAP5 Analyses of Overcooling Transients in a Pressurized Water Reactor
In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of
Analysis of loss-of-coolant accidents in the advanced neutron source reactor
The RELAP5 computer code and a model of the Advanced Neutron Source (ANS) were used to simulate system response to hypothetical loss-of-coolant accidents (LOCAs). The computer code was modified to
...
1
2
3
4
5
...