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NUREG/CR-6042, Rev. 2 [3:4] Appendix

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NUREG/CR-6042, Rev. 2 [3:4] Appendix Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities 3.7 Special Considerations for BWR for reactor vessel pressure control when run Facilities continuously in the recirculation mode, pumping water from the condensate storage Boiling water reactors have unique features tank back to the condensate storage tank and that would cause their behavior under severe periodically diverting a small portion of the accident conditions to differ significantly flow into the reactor vessel as necessary to from that expected for the pressurized water maintain the desired water level. The steam reactor design.1,2 This section addresses taken from the reactor vessel by the turbine several special considerations affecting BWR is passed to the pressure suppression pool as severe accident progression and mitigation. turbine exhaust, which provides a slower In this endeavor, many implications of the rate of pool temperature increase than if the phenomena described in Sections 3.1 through vessel pressure control were obtained by 3.6 (such as zirconium oxidation) will be direct passage of steam from the vessel to demonstrated by example. First, however, it the pool via the SRVs. Plants having both is necessary to review some of the BWR HPCI and RCIC systems can employ the features pertinent to severe accident HPCI turbine exclusively for pressure considerations. control while the RCIC system is used to maintain the reactor vessel water level. The 3.7.1 Pertinent BWR Features HPCI turbine is larger than the RCIC turbine and, therefore, is more effective for pressure An important distinction of the BWR design control. These systems require DC power for is that provisions are made for direct valve and turbine governor control, but have operator control of reactor vessel water level no requirement for control air. and pressure. Reactor vessel pressure control is normally accomplished rather All BWR-5 and -6 plants are equipped with simply by manually induced actuation of the an electric-motor-driven high-pressure core vessel safety/relief valves (SRVs) or by spray (HPCS) system rather than a turbine operation of the reactor core isolation driven HPCI system. The HPCS pump takes cooling (RCIC) system turbine or, for plants suction from the condensate storage tank and so equipped, the isolation condenser or high delivers flow into a sparger mounted within pressure coolant injection system (HPCI) the core shroud. Spray nozzles mounted on turbine. Each of these methods relies to the spargers are directed at the fuel bundles. some extent, however, upon the availability As in the case of HPCI, the pressure of DC power or control air, which may not suppression pool is an alternate source of be available under accident conditions. SRV water for the HPCS. considerations will be described in Section 3.7.2.5. All BWR facilities employ the low-pressure coolant injection (LPCI) mode of the All BWR plant designs except Oyster Creek, residual heat removal (RHR) system as the Nine Mile Point 1, and Millstone 1 dominant operating mode and normal valve incorporate either the RCIC or HPCI steam lineup configuration; the RHR system will turbine-driven reactor vessel injection automatically align to the LPCI mode system; the later BWR-3 and all BWR-4 whenever ECCS initiation signals such as plants have both. These systems can be used low reactor vessel water level or high USNRC Technical Training Center 3.7-1 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities drywell pressure are sensed. LPCI flow is and three times the mass of intended to restore and maintain the reactor zirconium metal (counting both vessel coolant inventory during a LOCA fuel rod cladding and channel after the reactor vessel is depressurized, box walls). either by the leak itself or by opening of the 2. The BWR reactor vessel would SRVs. be isolated under most severe accident conditions, due to All BWR facilities also employ a low closure of the main steam pressure core spray (LPCS) system, which isolation valves (MSIVs). This takes suction on the pressure suppression tends to make the BWR severe pool and sprays water directly onto the upper accident sequence thermal ends of the fuel assemblies through nozzles hydraulic calculation simpler to mounted in sparger rings located within the perform, since natural circulation shroud just above the reactor core. With the pathways through external loops reactor vessel depressurized, the such as hot legs and steam automatically-initiated LPCI and LPCS generators need not be flows, which begin when the reactor vessel considered. pressure-to-suppression pool pressure differential falls below about 2.00 MPa (290 3. Because of the marked reduction psig), are large. As an example, for a 1065 in the average radial power factor MWe BWR-4 facility such as Browns Ferry in the outer regions of the BWR or Peach Bottom, the combined flows would core, degradation events would be more than 3.16 m3/s (50,000 gal/min), occur in the central core region which is sufficient to completely fill an long before similar events would intact reactor vessel in less than four take place in the peripheral minutes. It should be recalled that the regions. amount of vessel injection necessary to remove decay heat (by boiling) is only about 4. SRV actuations would cause 0.013 m 3/s (200 gal/min). important pressure and water level fluctuations within the Eight BWR design features have important reactor vessel. Operator actions implications with respect to differences (mandated by emergency (from PWR behavior) in the expected procedure guidelines) to response of a BWR core under severe depressurize the reactor vessel accident conditions. These are: would lead to early (and total) uncovering of the BWR core. 1. There is much more zirconium metal in a BWR core, which 5. Diversity of core structures under similar conditions would (control blades, channel boxes, increase the amount of energy fuel rods) would lead to released by oxidation and the progressive, downward relocation production of hydrogen. of different materials from the Compared with a PWR of the upper core region to the core same design power, a BWR plate. typically contains about one and one-half times the mass of U0 2 USNRC Technical Training Center 3.7-2 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities 6. With early material accumulation representation of the core. Because of the upon its upper surface, the fate of associated reductions in decay heating core, the BWR core plate determines beyond the central region of the in the whether the initial debris bed predicted severe accident events peripheral would form within the lower central region lead those in the of time. For portion of the core or in the regions by considerable periods vessel lower plenum. example, formation and downward relocations of large amounts of debris are region before 7. Because there are many more calculated for the central to begin steel structures in the BWR lower structural degradation is predicted plenum, BWR debris would have in the outermost core region. a much greater steel content. 3.7.2 Provisions for Reactor Vessel 8. There is a much larger volume of Depressurization water (relative to the core volume) within the The BWR Owners Group Emergency structural 3 lower plenum beneath a BWR Procedure Guidelines (EPGs) require operators act to core. If conditions are favorable, unequivocally that the vessel debris relocating from the core manually depressurize the reactor uncovered region can be completely should the core become partially ) quenched - with sufficient water under conditions (such as station blackout remaining to remove decay heat characterized by loss of injection capability. requirement (by boiling) for several hours The operators would meet this without makeup. by use of the Automatic Depressurization System (ADS). The following discussions of an The importance of each of these items will address why manual actuation "automatic" system is necessary, what is be elucidated in the discussions of Sections by the rapid 3.7.2 through 3.7.7. Item 3, however, expected to be achieved the core deserves special amplification here. Figure depressurization, the status of 3.7-1 illustrates a typical division of a BWR during the subsequent periods of structural is not core into radial zones for code computation degradation (if the accident of keeping purposes. This example is based upon the terminated), and the importance the Browns Ferry Unit 1 core, which comprises the reactor vessel depressurized during accident. 764 fuel assemblies. Since a symmetric core latter stages of the loading is maintained, the drawing shows just one-fourth (191 assemblies) of the core. 3.7.2.1 Why Manual Actuation is Necessary What should be noted from Figure 3.7-1 is that the outer 25.1% of the core (sum of The most direct means of BWR reactor is by use of the volume fractions for zones 9 and 10) is vessel pressure control outside energy characterized by average radial peaking SRVs, which require no safety valve but do factors of just 0.670 and 0.354. This source for operation as a when dramatic falloff is illustrated by Figure 3.7 require both control air and DC power valve. 2, which also indicates the volume-averaged used as a remotely operated relief of central region power factor (1.199) This dependence upon the availability both to associated with a four-radial-zone code control air and DC power pertains USNRC Technical Training Center 3.7-3 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities remote-manual opening of the valves by the the core.
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