Scram delay

It has been established that absorber rod drop time should be less than 1 s so that the reactivity transient of a few dollar per second resulting from coolant voiding, fuel melting and slumping can be protected by the shutdown system. Analysis of reactivity transient shows that absorber rod speed can be in the range of 1-4 mm/s. 2 mm/s speed has been decided for CSR and 4 mm/s for DSR. Scram delay time on reactivity events should be less than 250 ms. 

An experimenter makes an error loading a rabbit sample. Injection of the sample results in a 100 millisecond period. If the scram setpoint is 1.25 MW and the scram delay time is 0.1 seconds, WHICH ONE of the following is the peak power of the reactor at shutdown (Assume rabbit system is operational for this question.)... 

A reactor is operating at 2 Mw (ioox full power) and the reactor scram is set for 1255 fuli power. What will be the peak power if a nuclear excursion creates a 100 millisecond period and the scram delay time is O.l seconds after i25X power is reached (Assume no temperature or void effects and that the period scram is inoperative)... 

Fig. 6.25. The main coolant flow rate decreases linearly to 50% of the rated flow. Flow rate low level 1 is detected and the scram signal is released at 1 s. Although the trip of the RCP itself would release the scram signal, it is conservatively neglected. The cladding temperature increases until 3.6 s due to the decrease in the flow rate and then decreases due to the decrease in the power. The increase in the hottest cladding temperature is 60°C which is the highest among the abnormal transients. It is sensitive to the coast-down time and the scram delay as shown in Table 6.11.

Flow rate Ratio of density coefficient Scram delay (s) ... 

The results of sensitivity analyses are summarized in Table 6.17. The influence of the coast-down time of the RCPs is not significant because the peak temperature appears at least 5 s after the coast-down has been completed. The influence of the scram delay is not significant because the reactor power is also decreased by the reactivity feedback. It should be noted that the peak temperature does not depend on the capacity or delay of the AFS because the peak temperature appears before the initiation of the AFS. 

Both the high and low level trips cure connected into the IXX Safety Circuit. Any tvo Beckmans in the tripped position will scram the reactor. Interlocks are provided so that bypassing tvo or more Beckmans at the same time will automatically scram the reactor. The BX and BXA relays are time-delayed... 

During initial periods, there were large number of spurious scram due to electro magnetic noise pickup by start up and delayed neutron detector channels and due to cross talk of fine impure test (FIT) pulse. Remedial measures viz., separation of signal cables from power cables, improvements in grounding system and reshaping FIT pulse were carried out and system performance improved. 

In the event of loss of coolant every effort will be made to avoid development of a reactivity state In the dry reactor (after the scram and after any power excursion which may have accompanied the accident) which would decrease the probability of restoring coolant to the reactor. Adequate control elements will be inserted or svsdlable for insertion Into the reactor to ensure subcrltlcallty Immediately following a complete loss of water from the reactor core. A delayed neutron critical state may be permitted after significant niunbers of process tubes have been rendered useless eis coolant channels. 

The power history of a system whose reactivity has been forced over prompt critical can be divided conveniently into two regions of time. The first of these is characterized by a rapid exponential rise and fall in power, commonly called the "spike. Following the spike the system usually adjusts itself to a relatively steady power which is sustained by delayed neutrons from precursors created during the spike and by whatever prompt multiplication exists in the system at this time. This portion of the power excursion, often referred to as the plateau, slowly decreases with time until the system is scrammed or in some manner becomes subcritical. In some accidental excursions the major portion of the fissions must have occurred in the region of the plateau. 

Answer The most important fission product to reactor operation is xenon-135 because of its large absorption cross-section - about 3.2 x 10 barns - for thermal neutrons. The effects of the xenon poison transients influence reactor operation in many ways. For example, although the saturated poison effect at equilibrium operation may cause a reactivity loss of 2 to 2-l/2 k (2000-2500 c-mk) in the usual power ranges, the decay of this amount of poison in the shutdown pile will represent a total swing of 4000 to 5000 c-mk from the initial xenon-free pile at startup. This delayed action effect in xenon formation and decay is the major cause for the scram transient, minimum downtime, and turnaround problems encountered in the operation of a hi power reactor. A detailed discussion on the effects of xenon poisoning on pile reactivity may be found in Chapter IV of this series. 

During the minutes following a reactor scram, reactor power decreases on a negative 80 second period, corresponding to the half-life of the longest-lived delayed neutron precursors, which is approximately ... 

There is a limit to the negative period that can be developed in a reactor by negative reactivity additions. Soon after the insertion of a large amount of negative reactivity such as a scram, the prompt neutron population decreases to a low level. Neutron population is predominantly the result of delayed neutrons which are produced by fission product decay. Within a short time, 2-3 minutes, all of the short lived delayed neutron precursors have decayed away. At this point, and from this point on, the core neutron population is sustained by decay of the longest lived fission product precursor, bromine-87, with a half life of = 55.72 seconds. Since the rate at which core neutron population decreases is determined by radioactive decay of bromine-87, an effective reactor period can be calculated by setting equations (2,9) and (4.7) equal. Neutron population, N will be used to replace activity, A, and power, P, respectively in the two equations. 

Ihc third trip witch will be adjusted to about 1 W reactor power level, and utilised In conjunction with a tlae delay re lay to provide capability of detecting an aboorroally high power level after a reactor scram, and initiate the backup (Ball 3X) safety system action. Xwo-out- af throc... 

Of critical importance in establishing conservative power limits is the initial power distribution within an assembly. Conservative upper bound delay scram times, as prescribed in the technical specifications, are used to maximize the power-flow mismatch. Non-ideal behavior and "hot" spot considerations (e.g., subchannel fuel hot-stripes and eccentricities, assembly azimuthal power tilts, etc.) are fully accounted for in the treatment of uncertainties, which significantly reduces the power from the nominal bounding best estimates case (Reference 14). 

The restart criteria require that transient and accident analyses include the assumption that mitigating systems are not actuated until the most conservative setpoint is reached. The Action Plans discuss setpoints and the time delays allowed for scram. In responding to the staff s inquiry, WSRC has indicated that worst-case setpoints are assumed for the analyses and, in particular, for the scram or the actuation of the Supplementary Shutdown System (SSS). WSRC further assured the staff that it has prepared specific calculations containing tables for all trip setpoints used for its analyses and that WSRC is currently developing a setpoint methodology document. Determination of setpoints is an open item. 

In Superphenix, taking into account uncertainties in an unfavourable way towards detection and time constants of detectors, the first trip parameter involved is the core outlet temperature that leads to the reactor shutdown, 3.8 s after beginning of the accident. The delay between this value and the one indicated in Fig. 2 is due to uncertainties. Later on (after 5 s), the scram threshold on reactivity is then reached, while the scram threshold on power ratio (P/Pn) is not reached. 

While the measurement time eonstant for P/Q, Lin P and p scram parameters is 0.05 s, it is 0.3 s for 0CSAM- In addition to this, trip logic delay time is 0.2 s. Control rod drop time is 0.8 s. Plant design ineorporates two diverse shutdown systems of worths 8000 pcm (Control... 

The pumps start cavitating within 0.05 s and their flows increase to 126% instantaneously. Core flow reaches a minimum value of 30% of nominal at 0.7 s. The power to flow ratio scram parameter crosses its threshold (1.1) at 0.06 s. However, due to the trip logic delay time, maximum reactivity and power of 21 pcm and 106% respectively are reached at 0.25 s before they are rapidly reduced by scram. 

The calculation results are shown in Fig. 6.36. The power decreases to the decay heat level due to the reactivity feedback and reactor scram. Reverse flow occurs in the water rod channel because the buoyancy pressure drop dominates the pressure drop balance. Heat conduction to the water rods increases when the coolant temperature in the fuel channel increases. This implies that the water rods serve as a heat sink . As the coolant expands in the water rods due to heat-up, there is an increase in the flow rate downstream from the water rods, including the fuel channel inlet. Consequently, the fuel channel flow rate is maintained even though the coolant supply from the cold-leg has stopped. This is called the water source effect of the water rods. The heat sink and water source effects mitigate heat-up of the fuel rod cladding, and hence enable the AFS to have a realistic delay time. The hottest cladding temperature begins to decrease before the initiation of the AFS. The increase in the hottest cladding temperature is about 250°C while the criterion is 520°C. 

D. Delayed Neutrons Make the reactor critical at approximately 100 W. When you are satisfied that all transients have died out, scram the reactor. Make a similog plot of the resulting flux as a function of time until the flux as indicated by the lowest range on Linear is within about 50% of chart zero. Then analyze the resultant curve to determine the delayed-neutron parameters, i.e., periods and approximate relative abundances. Only the three longest periods can be resolved. 

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