Analysis of N Reactor anticipated transients without scram

This report documents the N Reactor thermal-hydraulic system response for the anticipated transients without SCRAM (ATWS). The N Reactor is a graphite-moderated, water-cooled pressure-tube reactor. The primary coolant is pumped through 1003 horizontal pressure-tubes which contain two concentric tubular metallic fuel elements. The calculations were performed using RELAP5/MOD2 computer code. RELAP5/MOD2 is a best-estimate, two-phase, full six-equation, nonequilibrium, and nonhomogeneous thermal-hydraulic computer code designed for pressurized water reactor transients. A point-reactor kinetics model with reactivity feedback was employed to compute the power generated in the reactor core. The transients simulated were 200% double-ended guillotine break in the cold-leg manifold and the station blackout with the simultaneous failure of SCRAM and emergency core cooling system (ECCS) and the loss of ac power. In all the transients, the loss of high pressure injection system (HPI), the complete loss of graphite and shield cooling system (GSCS) and the isolation of steam generator secondary were included. Also, the ATWS transient, due to the 200% guillotine break of the cold-leg manifold assuming the availability of ECCS from three high-lift diesel pumps operating at full capacity and the loss of ac power, was analyzed. These three transients were first evaluated using the void feedback reactivity havingmore » an absolute level of {minus}167 mk corresponding to an infinite lattice MK IV fuel. Then these transients were reevaluated using a more realistic void feedback reactivity having an absolute level of {minus}70 mk. The report includes brief descriptions of the N Reactor, the computer code, the thermal-hydraulic plant model used in the simulations and the discussion of results. 12 refs., 113 figs., 13 tabs.« less