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 m3/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 3.7 Special Consideration for BWR Facilities Reactor Safety Course (R-800) representation of the core. Because of the 6. With early material accumulation upon its upper surface, the fate of associated reductions in decay heating beyond the central region of the core, the BWR core plate determines whether the initial debris bed predicted severe accident events in the would form within the lower central region lead those in the peripheral regions by considerable periods of time. For portion of the core or in the example, formation and downward vessel lower plenum. relocations of large amounts of debris are calculated for the central region before 7. Because there are many more structural degradation is predicted to begin steel structures in the BWR lower in the outermost core region. plenum, BWR debris would have a much greater steel content. 3.7.2 Provisions for Reactor Vessel Depressurization 8. There is a much larger volume of water (relative to the core structural volume) within the The BWR Owners Group Emergency lower plenum beneath a BWR Procedure Guidelines (EPGs)3 require unequivocally that the operators act to core. If conditions are favorable, debris relocating from the core manually depressurize the reactor vessel should the core become partially uncovered region can be completely quenched - with sufficient water under conditions (such as station blackout ) remaining to remove decay heat characterized by loss of injection capability. The operators would meet this requirement (by boiling) for several hours without makeup. by use of the Automatic Depressurization System (ADS). The following discussions The importance of each of these items will address why manual actuation of an "automatic" system is necessary, what is be elucidated in the discussions of Sections expected to be achieved by the rapid 3.7.2 through 3.7.7. Item 3, however, deserves special amplification here. Figure depressurization, the status of the core 3.7-1 illustrates a typical division of a BWR during the subsequent periods of structural core into radial zones for code computation degradation (if the accident is not purposes. This example is based upon the terminated), and the importance of keeping Browns Ferry Unit 1 core, which comprises the reactor vessel depressurized during the latter stages of the accident. 764 fuel assemblies. Since a symmetric core 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 volume fractions for zones 9 and 10) is vessel pressure control is by use of the characterized by average radial peaking SRVs, which require no outside energy factors of just 0.670 and 0.354. This source for operation as a safety valve but do dramatic falloff is illustrated by Figure 3.7 require both control air and DC power when 2, which also indicates the volume-averaged used as a remotely operated relief valve. central region power factor (1.199) This dependence upon the availability of associated with a four-radial-zone code control air and DC power pertains both to 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. Current EPGs direct that the control room operators and to the valve operators prevent automatic actuation of the opening logic of the ADS. ADS and instead manually initiate this system when the water level reaches the top The purpose of the ADS is to rapidly of active fuel. This intentional delay of depressurize the reactor vessel so that the ADS for cases when the low pressure pumps low-pressure emergency core cooling are running is a matter of controversy, and systems (ECCS) can inject water to mitigate not all BWR facilities invoke this provision.) the consequences of a small or intermediate loss-of-coolant accident should the high 3.7.2.2 Rapid Depressurization for pressure systems prove inadequate. The Steam Cooling number of ADS-associated SRVs is plant specific; these valves are signaled to open For BWR accident sequences involving automatically if required to provide reactor partial uncovering of the core, the EPGs vessel depressurization in response to low provide that the operators must take action reactor vessel water level caused by to initiate "steam cooling" which, for plants transients or small breaks. ADS initiation is without isolation condensers, is by coincidence of low reactor vessel water accomplished by manually initiating the level and high drywell pressure, provided ADS. The purpose is to delay fuel heatup that at least one of the low-pressure pumps by cooling the uncovered upper regions of is running. Recently, a bypass timer the core by a rapid flow of steam. Because (typically 265 seconds) has been backfitted the source of steam is the remaining to the ADS logic to ensure automatic inventory of water in the reactor vessel, the actuation of ADS on sustained low water steam cooling maneuver can provide only a level even if the high drywell pressure signal temporary respite. is not present. In order to illustrate the effects of steam There is, however, no timer bypass for the cooling, let us first consider a case in which requirement that at least one of the low this maneuver is not used. Figure 3.7-3 pressure ECCS pumps (RHR or Core Spray) shows the calculated reactor vessel water be running. (The actual signal is derived by level for a postulated loss of injection sensing the pump discharge pressure.) This (caused by station blackout and failure of is reasonable, since a great deal of water is RCIC) at Grand Gulf. It should be noted lost from the reactor vessel when the ADS is that after falling below the top of active fuel actuated and therefore it is prudent to (TAF), the calculated curve follows the require that a replacement water source be exponentially decreasing water level available. As explained in the next Section, predicted by Figure 3.2-1 until a downward however, it is desirable to actuate the ADS deviation becomes apparent, near the bottom under certain severe accident situations even of active fuel (BAF). This deviation occurs though there is no operating pump. Without because debris relocating from the upper, the discharge pressure signal, the ADS must uncovered, region of the core is relocating be actuated manually (operator pushbuttons). downward into the water remaining in the lower portion of the core, accelerating its (NOTE: Typically, the ADS timer is initiated boiloff. when the reactor vessel water level is between two and three feet above the top of USNRC Technical Training Center 3.7-4 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities Let us now consider the same accident about time 110 minutes. Without ADS sequence, but with implementation of the actuation, debris relocation begins about 23 steam cooling maneuver. Figure 3.7-4 shows minutes earlier and before core plate dryout, the calculated reactor vessel pressure and which is delayed until time 102.5 minutes. water level when the ADS is actuated at about one-third core height (75 minutes after It is instructive to consider why the first scram). At this time, there has been no local core plate failure occurs earlier for the degradation of the upper core. At Grand case with ADS actuation. Recall that the Gulf, 8 SRVs (of 20 total) are associated core plate is dry when debris relocation with the ADS. The vessel depressurizes begins in this case, so that the hot debris quickly and the accompanying water loss due falls directly on the plate surface. When the to flashing causes the water level to fall into ADS is not actuated, the initial debris the lower plenum, well below the BAF and relocations fall into water overlying the core the core plate. Subsequently, the flashing plate. This initial debris is quenched and ceases, and the remaining water is forms a protective layer over the plate. significant for debris quenching. Later, when water no longer remains over the plate, failure is delayed until the newly The maximum fuel rod temperature in the relocating debris has heated both the plate central region of the core is plotted versus and the previously quenched debris. time in Figure 3.7-5, for both cases. The temperature escalations that occur after time It is important that the ADS be manually 80 minutes for the case without ADS are initiated at the proper time. Too soon means caused by the energy releases associated with that reactor vessel water inventory will be zirconium oxidation. (The dotted lines on lost without the compensatory benefits this figure indicate the time at which the afforded by effective steam cooling. Too temperature increases above 1832°F late means that a steam-rich atmosphere will [1273°K]). For the case with ADS exist during the onset of runaway metal actuation, the temperature decreases water reactions. By procedure, steam immediately after the valves are opened due cooling is to be placed into effect when the to the effects of steam cooling. "Minimum Zero-Injection RPV Water Level" Subsequently, the temperature again is reached. In Revision 4 of the EPGs, this increases, but the time at which the runaway is defined as the lowest vessel level at which zirconium oxidation temperature is reached the average steam generation rate within the has been delayed by about 15 minutes. The covered portion of the core is sufficient to differences in hydrogen generation between prevent the maximum clad temperature in the two cases during the period plotted are the uncovered region of the core from substantial, as can be appreciated by a exceeding 1800'F (1255 K). This level is comparison of the two subplots of Figure plant-specific; the basis for its determination 3.7-6. and procedures for its calculation are described in Appendices to the EPGs. Table 3.7-1 displays the times associated with the major events of the accident 3.7.2.3 Core Region Dry During Core sequence for both cases. When the ADS is Degradation actuated, core plate dryout follows immediately thereafter, and debris begins As explained in the previous section, the relocating from the upper core region at delay in the onset of core damage gained by USNRC Technical Training Center 3.7-5 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities use of the steam cooling maneuver is 3.7.2.4 Threat of Reactor Vessel temporary. Nevertheless, staving off the Repressurization onset of core degradation, which otherwise would begin at about 78 minutes, for an The motivation for keeping the reactor additional 15 minutes can be significant vessel depressurized under severe accident when trying to regain electrical power or conditions is, first, that the capacity for implement other means of restoring reactor quenching of the debris relocating from the vessel injection capability. Even if such core region into the lower plenum is efforts are unsuccessful so that the accident enhanced and, second, that relocation of sequence proceeds into core degradation, the molten debris into the relatively small BWR steam cooling maneuver provides the benefit drywell would then be, should bottom head of assuring that the core region will be penetration failures occur, by gravity steam-starved when runaway metal-water induced flow and not by rapid vessel reaction temperatures are reached. blowdown. For the less probable case that penetration failures do not occur, so that the When considering severe accident vessel bottom head ultimately undergoes progression for BWRs, it is extremely gross failure by creep rupture, the time of important to recognize that when the failure would be delayed by several hours. specified procedures are followed, the core Keeping the reactor vessel depressurized region would be dry during the period of eliminates direct heating concerns and core degradation. As illustrated in the water greatly reduces the initial challenge to the level plot included with Figure 3.7-4, integrity of the primary containment. execution of the steam cooling maneuver causes the water level to fall below the core The chief threats that the reactor vessel may plate. This plot represents the results be pressurized at the time of bottom head calculated when the ADS valves are opened failure arise from two considerations, one with the water level at about one-third core derived from the potential for equipment height, but the final level will fall below the failure and the other derived from the core plate even if this maneuver is initiated possibility of operator error. The question with the water level near the top of the core of equipment failure is chiefly associated (although the achieved fuel temperature with the long-term station blackout accident decrease will be much less). sequence, for which injection capability is maintained until the unit batteries are lost. Figure 3.7-7 shows the water level relative With loss of the batteries, the ability to to the core plate immediately after execution operate the SRVs manually is also lost. of the ADS maneuver. It should be noted Because multiple SRVs are installed and that some water is trapped in the downcomer operation of any one valve is sufficient for region surrounding the jet pumps. This depressurization under severe accident occurs because the initial temperature of the conditions, improved reliability of SRV water in the jet pump region is less than the operation can be attained simply by ensuring temperature of the water in the core region. that the small amount of DC power and Hence, a lower proportion of the water in the control air necessary for opening will be downcomer region is flashed during the available (to at least one valve) when rapid vessel depressurization required. Several BWR utilities have taken steps toward this end such as provision of a USNRC Technical Training Center 3.7-6 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities dedicated small DC diesel generator and discharge from each valve is piped to the backup compressed nitrogen supply bottles. pressure suppression pool, with the line terminating well below the pool surface, so The impact of operator error upon vessel that the steam is subject to condensation in depressurization as typically represented in the pool. The number of SRVs varies from probabilistic risk assessments is direct. It is plant to plant (e.g., 11 at Limerick; 24 at postulated that the operators fail to take the Nine Mile Point 2), as do the rated relief required action to manually depressurize the valve flows. reactor vessel. Although the assumed probability for such failure, say 0.001, may Some operating BWRs are equipped with seem low, typical core melt frequencies are three-stage Target Rock valves, which have much lower - on the order of 10'. exhibited a greater tendency to stick open in Accordingly, it is important to recognize that the past than have other types of valves. such assumptions concerning operator error, Many BWR utilities, however, have replaced while seeming reasonable and conservative, the original three-stage valves with the may lead to the unrealistic conclusions that newer two-stage Target Rock valves (Figure BWR core melt, should it happen, would 3.7-8). Some operating BWRs are equipped always occur in a pressurized vessel and that with Dresser electromatic relief valves. there is no point in providing equipment BWR-5 and BWR-6 plants are equipped with upgrades such as a dedicated DC generator Crosby and Dikkers dual function SRVs since the operators would not use them (Figure 3.7-9). anyway. The differences in SRV operation in the 3.7.2.5 Notes Concerning automatic and remote-manual or ADS modes Operatior can be demonstrated with reference to the two-stage Target Rock design shown in Any serious attempt to study and Figure 3.7-8. During normal reactor comprehend the probable course of an operation, a small piston orifice serves to unmitigated severe accident sequence at a equalize the steam pressure above and below BWR facility must include development of a the main valve piston, and the main valve thorough understanding of the operation of disk remains seated. The reactor vessel the installed SRVs under abnormal pressure (valve inlet pressure) is ported via conditions of reactor vessel and containment the pilot sensing port to tend to push the pressure. The pertinent characteristics of the pilot valve to the right. When the reactor more common SRV designs are described in vessel pressure exceeds the setpoint the following paragraphs. The reader should established by the setpoint adjustment particularly note that control air pressure spring, the pilot valve is moved to the right, sufficient for valve operation under normal the stabilizer disk is seated, and the volume conditions may not be adequate if the reactor above the main valve piston is vented to the vessel is depressurized and the containment valve outlet via the main valve piston vent. pressure is elevated. The sudden pressure differential causes the main valve piston to lift, opening the valve. All SRVs are located between the reactor vessel and the inboard main steam isolation For the remote-manual or ADS modes, the valves (MSIVs) on horizontal runs of the SRV opening is initiated by control air, main steam lines within the drywell. The which is admitted via a DC solenoid- USNRC Technical Training Center 3.7-7 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities operated valve (not shown) to the air inlet at drywell control air system. For severe the right of the setpoint adjustment spring. accident considerations, it is important to The control air moves the valve actuator to recall that remote operation of the SRVs is the right (against drywell pressure), which possible only as long as DC power remains compresses the setpoint adjustment spring available and the pneumatic supply pressure and pulls the pilot valve open, seating the exceeds the containment pressure by some stabilizer disk and venting the space above minimum amount. the main valve piston. Because the control air pressure and the reactor vessel pressure 3.7.3 Recriticality Concerns work in tandem to move the pilot valve to the right, the amount of control air pressure The progression of damage and structural required to open the SRV will depend upon relocation of the various components the reactor vessel-to-drywell pressure (control blades, channel boxes, fuel rods) of differential. Also, because the control air a BWR core during an unmitigated severe acts to move the air actuator against drywell accident sequence will be discussed in detail pressure, the required control air pressure in Section 3.7.4. There it will be shown that will increase with drywell pressure. the first structures to melt and relocate downward are the control blades. Here we (It should be noted that the three-stage pause to consider severe accident sequences Target Rock valves behave differently with that have the potential for early termination, respect to the effect of the reactor vessel-to i.e., accident sequences for which the core drywell pressure differential. A good structure sustains significant damage but description of the operation of this older reactor vessel injection capability is restored valve design is available in Reference 4.) while the major portion of the fuel remains above the core plate. The spring-loaded direct-acting SRV shown in Figure 3.7-9 is opened in the spring mode If significant control blade melting and of operation by direct action of the reactor relocation were to occur during a period of vessel pressure against the disk, which will temporary core uncovering, then criticality pop open when the valve inlet pressure would follow restoration of reactor vessel exceeds the setpoint value. In the power injection capability if the core were rapidly actuated mode, a pneumatic piston within the recovered with cold unborated water using air cylinder moves a mechanical linkage to the high-capacity low-pressure injection compress the spring and open the valve. As systems. Obviously, a neutron poison in the case of the two-stage Target Rock should be introduced into the reactor vessel valve, the control air is provided via DC for reactivity control under these solenoid-operated valves, and the air circumstances, but question arises as to how pressure required for valve opening best to do this. The normal means of adding decreases with reactor vessel pressure and boron to the reactor vessel is by injection increases with drywell pressure. with the standby liquid control system (SLCS). Although -this system is designed to All SRVs associated with the ADS are fitted inject sufficient neutron-absorbing sodium with pneumatic accumulators (located within pentaborate solution into the reactor vessel the drywell) to ensure that these valves can to shut down the reactor from full power be opened and held open for some (plant (independent of any control rod motion) and specific) period following failure of the to maintain the reactor subcritical during USNRC Technical Training Center 3.7-8 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R1-800) 3.7 Special Consideration for BWR Facilities cooldown to ambient conditions, the SLCS is capability. If the SLCS is used to inject not intended to provide a backup for the sodium pentaborate at a relatively slow rate rapid shutdown normally achieved by scram. while the core is rapidly recovered with unborated water using the high-capacity, As indicated in Figure 3.7-10, the basic low-pressure injection systems, then SLCS comprises a heated storage tank, two criticality would occur and the core would 100% capacity positive displacement pumps, remain critical until sufficient boron for and, as the only barrier to injection into the shutdown (at the prevailing temperature) reactor vessel, two explosive squib valves. reached the core region. To avoid the In most of the current BWR facilities, the possibility of temporary criticality, it would sodium pentaborate solution enters the be desirable to inject effective quantities of reactor vessel via a single vertical sparger boron along with the ECCS flow being used located at one side of the lower plenum just to recover the core. A strategy to accomplish below the core plate. However, so as to this using only existing plant equipment but improve the mixing and diffusion of the employing a different chemical form for the injected solution (which has a specific boron poison has been proposed.' gravity of about 1.3) throughout the core region, some BWR facilities have been The only currently available information modified to provide a third displacement concerning the poison concentration required pump and to permit the injected solution to is derived from a recent Pacific Northwest enter the reactor vessel via the core spray Laboratory (PNL) study,5 which indicates line and sparger. that much more boron would have to be injected than is available (as a solution of For the purpose of reducing the time sodium pentaborate) in the SLCS. required for reactor shutdown for the ATWS Furthermore, the dominant loss-of-injection accident sequence, the NRC has issued a accident sequence is station blackout, and Final Rule6 requiring that the SLCS injection without means for mechanical stirring or be at a rate equivalent to 86 gal/min (0.0054 heating of the injection source, the ability to m3/s) of 13 wt.% sodium pentaborate form the poison solution under accident solution, the boron being in its natural state conditions becomes of prime importance. with 19.8 at.% of the boron-10 isotope. Hence the need for the alternate chemical With this increased injection rate, sufficient form. boron for hot shutdown can be pumped into the reactor vessel in about 20 minutes, and The PNL study5 provides the estimate that a for cold shutdown in about 48 minutes. It boron-10 concentration of between 700 and requires approximately an hour to inject the 1000 ppm would be required within the entire contents of the tank. vessel to preclude criticality once control blade melting had occurred. This is much The operators would have no direct means of greater than the concentration (about 225 knowing Whether significant control blade ppm) attainable by injection of the entire relocation had occurred. Thus, there is a contents of the SLCS tank. strong potential for surprise should, for example, a station blackout accident At this point, it should be noted that the sequence suddenly be converted into an conclusions of the PNL study with respect to uncontrolled criticality upon restoration of the boron concentrations required to preclude electrical power and reactor vessel injection criticality are acknowledged by the authors USNRC Technical Training Center 3.7-9 NUREG/CR-6042 Rev. 2 Reactor Safety Course (R-800) 3.7 Special Consideration for BWR Facilities of that study to be very conservative. Stated tank can be gravity-drained through the another way, in the many instances where it standpipe to the main condenser hotwells was necessary to make assumptions during under station blackout conditions. the study, the assumed quantities were selected in a manner that tends to increase Additional information concerning this reactivity (promote criticality). As an example of a candidate accident management example, debris particles are assumed to strategy and the characteristics of the exist in the form of spheres. As discussed in alternate boron poison chemical form is the following paragraphs, the resulting high available in Reference 7. It seems desirable boron concentration requirement makes that the very conservative estimates of the development of a practical coping strategy PNL study should now be replaced by more difficult. realistic estimates, which certainly would be expected to lower the target boron One means to achieve such a high boron concentration from its present value of 700 concentration would be to mix the powder ppm and thereby improve the practicality of directly with the water in the condensate such a strategy. (For example, Reference 8, storage tank during the blackout period and which incorporates an assumption that three then, once electrical power is restored, to fourths of the control blade B C remains in 4 refill the reactor vessel by pumping the the core region, suggests that reflood water solution in a controlled manner using one of boron-10 concentrations as low as 200 ppm the low-pressure injection system pumps. might be sufficient.) In the meantime, many of the BWR facilities have implemented The condensate storage tank is an important accident management measures, on a source of water to the reactor vessel voluntary basis, to provide backup capability injection systems for each BWR unit. As for the SLCS. These backup strategies indicated in Figure 3.7-11 (based upon the invoke such methods as modification of the Browns Ferry arrangement), it is the normal HPCI or RCIC pump suction piping to suction source for the steam turbine-driven permit connection to the SLCS tank, or HPCI and RCIC systems and the alternate poisoning of the condensate storage tank. source for the electric motor-driven RHR and core spray pumps. 3.7.4 Eutectic Formation and Relocation Sequence for BWR Core Structures During normal reactor operation, the condensate storage tank provides makeup This section addresses the progression of flow to the main condenser hotwells via an damage and structural relocation of BWR internal tank standpipe, as indicated on core components that would be expected to Figure 3.7-12. The purpose of the standpipe occur during an unmitigated severe accident is to guarantee a reserve supply of water for sequence, i.e., an accident sequence for the reactor vessel injection systems that take which reactor vessel injection capability is suction from the bottom of the tank. Any not restored. The BWR core is basically an practical strategy for direct poisoning of the assembly of unit .cells, one of which is tank contents must include provision for shown in the center drawing of Figure 3.7 partial draining to reduce the initial water 13. As indicated, each unit cell comprises volume, especially if boron-10 four fuel assemblies, each located in one of concentrations on the order of 700 ppm are the four quadrants of a central control blade. to be established. The condensate storage Additional details concerning the fuel USNRC Technical Training Center 3.7-10 NUREG/CR-6042 Rev. 2