Reactor Safety Information Computational

In modern reactor analysis, the calculations previously outlined are performed on computer codes specially written for the purpose. For most actual operating reactors, these computer codes are proprietary to the reactor vendors (or, in some cases, the utilities that own the reactor) and are not generally available to outside analysts. However, there are computer codes available from Reactor Safety Information Computational Center (RSICC), the computer code distribution center at the Oak Ridge National Laboratory, which can be used for reactor analysis. For the most part, the available computer codes are those that were developed at the various national laboratories (Oak Ridge National Laboratory [ORNL], Los Alamos National Laboratory [LANL], Argonne National Laboratory [ANL]), although some university and privately developed software is also available. 

BELLA a computer code written specifically for the purpose of safety-informed design of lead-cooled fast reactors... 

The Data Processing System displays all operating parameters and the state of the plant. Alarms are displayed separately. Operator action is taken immediately to clear operational faults. Plant defects relating to safety are investigated and cleared if possible in 1 day. Lesser defects are cleared within 3 days. Reasons for all standing alarms are recorded for the information of all operational staff. Major faults may result in the immediate shutdown of the reactor or wait for a planned outage. Safety related plant is covered by alternative conventional instrumentation as backup to the computer-derived data. 

The status of safety-related systems and fiinctions is presented in a similar way, in accordance with the organization of the Emergency Operation Procedures (EOP). The parameters that are of immediate interest in a disturbance situation, are presented in a direct form. This means that the reactor pressure vessel with in- and outflow connections, together with neutron flux, water level, and reactor pressure, as well as control rods fiilly in (or not), are displayed directly. Other safety functions are indicated as normal, disturbed or failed in a similar way as for the plant overview, with detailed information at the reactor operator s desk. In this context, it can be noted that the computer-based reactor scram function via the reactor protection system (RPS) has been supplemented by a scram backup system that is implemented in hard-wired equipment. 

Time-dependent neutron population data for these reactivity measurements are obtained from high sensitivity ion chambers located outside the reactor core. This Information is multiscaled into an on-line computer at a rate of up to 100 cps, and is used to determine reactivity as a flinction of time during the course of a. power transient. This technique allows a complete safety-rod scram-time and worth measurement to be made within one minute, as compared to several hours by other methods. 

The overall objectives of this chapter are to (1) provide a background on heat transfer in reactor systems (2) describe methods of analysis employed in the reactor thermal-hydraulics and safety with basic analysis processes and tools and (3) provide analysis examples, sources of information, and computer codes used for detailed reactor thermal-hydraulics and safety analysis. 

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