Low pressure coolant injection modification for boiling water reactors

A conventional low pressure coolant injection system for boiling water reactors is provided. With the modification, the cross tie conduits and associated valves remain open between two LPCI divisions. On the occasion of an LOCA, the LPCI pumps are activated and injection valves for each of the LPCI divisions are opened to commence coolant injection in the recirculation loops in simultaneous fashion. However, the rate of flow of water coolant within each injection system is controlled by a hydraulic resistance, which is selected to achieve reactor core cooling within requisite time and in requisite quantities from one injection path. Thus, even though coolant water may flow through a rupture within one recirculation loops, sufficient water will be injected in the other loop to achieve core cooling.

BACKGROUND OF THE INVENTION 
Nuclear power plants traditionally have been designed for achieving long 
term, safe, and reliable performance. To assure safety, the plants 
incorporate systems and procedures representing a studied anticipation of 
emergency conditions. Design approaches will have considered theories or 
premises which may include, for example, design redundancies which are 
challenged by updated rules of performance as operating experience with 
nuclear power progresses. Thus, investigators in this power field 
continuously are called upon to develop improved analytic models of 
operation exhibiting improved boundings of operational factors and to 
further achieve higher levels of safety in view of changing rules of 
safety related performance. Because of the necessarily extensive time 
interval involved in developing or constructing a new nuclear power 
facility, for example such an effort may encompass ten years or more, and 
further in view of the numerous nuclear power facilities now in operation, 
these investigators typically are called upon to meet new rule criteria by 
modification of long-existing facilities. Retrofitting procedures can be 
quite expensive, calling for revised electrical power supplies, major 
valving replacements, and the like. 
The nuclear industry has evolved a variety of reactor types. One such type 
finding substantial field use performs to produce steam for turbine drive 
within the reactor core itself and is referred to as a boiling water 
reactor (BWR). The reactor heated water of the BWR serves not only as 
working fluid, but also as a reaction moderator, and along with other 
parameters, its proper supply and application within the system 
necessarily has been the subject of safety requirements or rule 
generations by government regulatory agencies such as the Nuclear 
Regulatory Commission (NRC). 
Typically, the general structure of a BWR nuclear system will include an 
upstanding reactor vessel which incorporates a lower reactor core 
structure beneath which are control rod drives. Above the core are, in 
order, a steam separator assembly and a steam dryer assembly leading to a 
steam outlet. About the reactor is a shield wall and outwardly of that a 
drywell. A pressure suppression chamber (wetwell), being torroidal in 
shape, is located below and encircling the drywell. 
In more typical BWR installations, water coolant is heated in the reactor 
core to rise within the reactor vessel as a two-phase mixture of water and 
steam. This dual phase mixture then passes upwardly through the steam 
separator assembly and steam dryer structure to enter the steam line 
leading to a turbine. Following turbine drive, the steam is condensed to 
water and returned to the reactor by relatively large condensate and 
feedwater pumps of a feedwater system. The feedwater enters the downcomer 
region of the reactor, where it is mixed with the water returning from the 
steam separator and drying functions. The water in the downcomer region is 
circulated through the reactor core via the vertically oriented 
recirculation pumps which direct flow to the vertical jet pumps located 
between the core shroud and vessel wall (downcomer annulus). In typical 
fashion, two distinct recirculation loops with corresponding recirculation 
pumps are employed for this recirculation function. 
In the event of some form of breakage or excursion generating malfunction, 
referred to as a "loss-of-coolant accident" (LOCA), designers anticipate 
that the relatively higher temperature-higher pressure water within the 
reactor will commence to be lost. A variety of safety systems and 
procedures may then be invoked both for containment and for thermal 
control of this LOCA. For the latter, thermal control, safety designs 
recognize that, while loss of the water moderator terminates the core 
reaction to eliminate a possibility of a nuclear incident, the momentum of 
generated heat or the residual energy within the reactor will remain of 
such magnitude as to require a cooling control to avoid, for example, core 
melt down. In general, the amount of water within the containment system 
is more than adequate for this purpose, for example that contained in the 
suppression pool, or additionally, the condensate storage tank. To apply 
this water coolant for the safety purpose, a variety of safety related 
techniques or "emergency core cooling systems" (ECCS) have been developed 
to accommodate the LOCA. For example, core spray (CS) systems and low 
pressure coolant injection (LPCI) installations have been evolved in a 
variety of configurations. 
The LPCI system incorporates, for example, four pumps which are activated 
by a safety system in the event of a coolant loss. Where the loss of 
coolant is of sufficient extent, and the vessel pressure remains high, for 
example in the event of a small pipe break then, an automatic safety 
system will function to depressurize the reactor vessel permitting the 
relatively lower pressure water supply pumps to operate to introduce water 
to the reactor. Because the recirculation system as earlier described is 
ideally structured for this purpose, generally it is used by the LPCI 
system for water introduction under ECCS conditions. 
Safety designs heretofore have recognized, however, that a recirculation 
loop may be broken under a LOCA condition. Thus, the pumping of water into 
that loop under such a LOCA condition may have no effectiveness. 
Accordingly, the LPCI systems have been equipped with a recirculation loop 
selection feature termed "loop selection logic" to avoid such conditions. 
This safety control detects the broken recirculation loop and initiates a 
procedure injecting water into the redundant, intact recirculation loop by 
actuating appropriate LPCI injection valves. Experience with such LPCI 
loop selection features have shown them to be complex and difficult to 
test and maintain. Under more current rule-based requirements, the design 
must accommodate for such occurrences as valve failure and the like. 
However, to function more effectively under current rules, procedures for 
retrofitting existing facilities to update them are elaborate and quite 
expensive, implementation involving such activities as recabling, pump 
reconnection activities and the like. Thus, an approach has been sought by 
investigators which offers operators the opportunity to eliminate the 
requirement for a loop selection logic regimen and associated costs 
therewith while still improving the reliability of the LPCI system. 
SUMMARY 
The present invention is addressed to an LPCI system and method which 
provides for injection loop modification achieving effective insertion of 
water coolant within the recirculating loops of conventional boiling water 
reactors, but without resorting to complex loop selection logic. The 
procedure recognizes that a break or rupture may have occurred in one of 
the recirculation loops and controls the rate and quantity of simultaneous 
coolant injection into each recirculation loop. Through analysis by 
modeling and the like of the requirements of the LPCI system in terms of 
time for complete coolant injection and in terms of the required quantity 
of injected fluid, flow rates of injection are derived and requisite 
quantities of coolant are determined and identified such that the LPCI 
process is controlled through the simple approach of utilizing flow rate 
controlling hydraulic resistances within coolant injection conduits. Those 
hydraulic resistances may be implemented with a conventional orifice, the 
size and shape of which determines desired flow rates or by the throttling 
of a valve within the injection conduit achieving the equivalent result. 
Under the process, cross tie conduits and associated cross tie valving 
otherwise used for recirculation loop selection for coolant injection are 
not activated, but merely remain in an open condition. Under the new 
method and system, necessary LCPI modifications are achieved without 
resort to the complicated system and instrumentation otherwise required 
for loop selection with a minimum of hardware perturbation, rewiring or 
repiping. 
As another feature, the invention provides a low pressure coolant injection 
system for a nuclear power facility of a variety having a boiling water 
reactor, having a reactor core and normal operating pressure, first and 
second recirculation loops including respective first and second 
recirculation pumps and actuable discharge valves, a suppression pool 
water source, a condensate storage tank, and a safety system responsive to 
a loss-of-coolant accident to generate a safety output. The system 
includes first and second low pressure coolant injection pumps having 
suction inputs and discharge outputs and actuable to pump water. A supply 
conduit arrangement is provided for coupling the suction inputs of the 
first and second low pressure coolant injection pumps in fluid flow 
communication with the suppression pool. First and second coolant 
injection conduits are provided which are coupled with respective 
discharge outputs of the first and second low pressure coolant injection 
pumps and to respective first and second recirculation loops. First and 
second hydraulic resistance components within respective first and second 
coolant injection conduits are provided for restricting the flow of water 
coolant therein to a predetermined fluid rate selected to deliver a 
predetermined quantity of water coolant to each of the first and second 
recirculation loops, the flow rates being selected as effective for 
carrying out the emergency cooling of the reactor core from one coolant 
injection conduit. A control arrangement is provided which is responsive 
to the safety output for actuating the first and second low pressure 
coolant injection pumps. 
As another feature, the invention provides a method for injecting low 
pressure cooling water into the boiling water reactor of a nuclear power 
facility having a source of emergency core cooling water, first and second 
independent recirculation loops normally circulating water through the 
core of the reactor for steam generation and a safety system responsive to 
a loss-of-coolant accident to generate a safety output for effecting the 
supply of at least a predetermined quantity of water coolant to the 
reactor, comprising the steps of: 
providing first and second water flow paths from the source of water 
coolant to respective first and second recirculation loops; 
providing low pressure coolant injection pumps actuable for pumping water 
from the source through the first and second water flow paths; 
providing a valve arrangement actuable from a closed to an open condition 
for effecting flow within the first and second water flow paths; 
actuating the valve arrangement in response to the safety output to permit 
water coolant flow simultaneously in each first and second water flow 
path; 
actuating the low pressure coolant injection pumps in response to the 
safety output; and 
restricting the flow of the water coolant in each first and second water 
flow path to a predetermined fluid flow rate selected to deliver the 
predetermined quantity of water coolant to each respective first and 
second independent recirculation loops, said flow rate being selected as 
effective for carrying out the emergency cooling of the reactor core from 
one water flow path. 
As another feature, the invention provides a low pressure coolant injection 
system for a nuclear power facility of a variety having a boiling water 
reactor with a reactor core, and normal operating pressure, first and 
second recirculation loops including respective first and second 
recirculation pumps and actuable discharge valves, a suppression pool 
water source, a condensate storage tank, and a safety system responsive to 
a loss-of-coolant accident to generate a safety output. The system 
includes first and second low pressure coolant injection pumps having 
suction inputs and discharge outputs and actuable to pump water. A supply 
conduit arrangement is provided for coupling the suction inputs of the 
first and second low pressure coolant injection pumps in fluid flow 
communication with the suppression pool and further includes a cross tie 
conduit arrangement for selectively interconnecting the discharge outputs 
of the first and second low pressure coolant injection pumps. First and 
second coolant injection conduits are provided which are coupled with 
respective discharge outputs of the first and second low pressure coolant 
injection pumps and to respective first and second recirculation loops. 
First and second low pressure coolant injection valves are provided within 
respective first and second coolant injection conduits and are actuable 
between closed and open orientations. Further provided are first and 
second hydraulic resistance devices within respective first and second 
coolant injection conduits for restricting the flow of water coolant 
therein to a predetermined fluid rate selected to deliver a predetermined 
quantity of water coolant to each of the first and second recirculation 
loops, the flow rate being selected as effective for carrying out the 
emergency cooling of the reactor core from one coolant injection conduit. 
A cross tie valve arrangement is provided within the cross tie conduit 
which is actuable between open and closed conditions for selectively 
directing the outputs of the first and second low pressure coolant 
injection pumps to one of the first and second recirculation loops through 
select first and second coolant injection conduits. A control arrangement 
is provided which is responsive to the safety output for actuating the 
first and second low pressure coolant injection pumps, the first and 
second low pressure coolant injection valves and retaining the cross tie 
valve arrangement in the open condition in the presence of the safety 
output. 
Other objects of the invention will, in part, be obvious and will, in part, 
appear hereinafter. 
The invention, accordingly, comprises the system and method possessing the 
construction, combination of elements, arrangement of parts and steps 
which are exemplified in the following description. 
For a fuller understanding of the nature and objects of the invention, 
reference should be had to the following detailed description taken in 
conjunction with the accompanying drawings.

DETAILED DESCRIPTION OF THE INVENTION 
Low pressure coolant injection systems (LPCI) necessarily perform with 
certain of the containment system based water retaining components of a 
nuclear power facility. Referring to FIG. 1, a containment or reactor 
building is represented generally at 10. Schematically represented within 
the figure is an outer wall 12 having a floor 14. Within the structure 10 
is a reactor pedestal 16 which is a component of a biological shield wall 
18 which surrounds a boiling water reactor (BWR) pressure vessel or 
reactor 20. A drywell is shown at 22 surmounted and defined by a steel 
structure or wall 23. A pressure suppression chamber or wetwell is 
represented at 24 which is torroidal in shape, surrounding the drywell 22. 
This suppression chamber 24 is approximately half filled with water to 
define a pressure suppression pool 26 and a vent system connects the 
drywell 22 to the wetwell 24 suppression pool 26. Drywell to wetwell vents 
as represented by main vents 28 and 29 extend from the drywell 22 to the 
suppression chamber 24 and are seen connected to respective vent headers 
30 and 31 contained within the air space of the suppression chamber 24. 
Downcomer pipes are seen at 32 and 33 extending from respective headers 30 
and 31 downwardly to terminate below the water surface of the suppression 
pool 26. 
In the highly unlikely event of a high energy Nuclear Steam Supply System 
(NSSS) piping failure within the drywell 22, reactor water and/or steam 
would be released into the drywell 22 atmosphere to define a loss of 
coolant accident (LOCA). As a result of increasing drywell pressure, a 
mixture of drywell atmosphere, steam, and water would be forced through 
the vent system including main vents 28 and 29 into the pool of water 26 
maintained within the suppression chamber 24. The steam vapor would 
condense into the suppression pool 26, thereby limiting internal 
containment pressure. The non-condensable drywell atmosphere would be 
transferred to the suppression chamber and contained therein. 
The secondary containment or reactor building 10 further may include such 
features as reactor building rooms as at 34 and 35 as well as a refueling 
floor 36. Intermediate the rooms 34 and 35 are such components as spent 
fuel storage pools as shown at 37 and 38. Not shown, but remotely located 
would be a condensate storage tank (CST). 
The suppression pool 26 provides, as noted above, a means to condense any 
steam released in the drywell area during a hypothetical LOCA; provides a 
heat sink for the reactor core isolation cooling system during hot 
stand-by operation until the decay heat can be piped directly to residual 
heat removal (RHR) heat exchangers; provides a heat sink for venting the 
nuclear system safety/relief valve; and provides a source of water for the 
emergency core cooling systems (ECCS). The suppression pool also serves as 
a heat sink under normal operating conditions. Blow-down through the main 
stream safety/relief valves during anticipated reactor transients is 
routed to the suppression pool or the steam discharges through a quencher 
at the end of the discharge piping and is condensed. 
Referring to FIG. 2, a highly schematized representation of the 
recirculation components as they relate to reactor vessel 20 is provided. 
Vessel 20 is seen to include a reactor core 44 which is comprised of a 
matrix positioned array of fuel assemblies extending between a core plate 
46 and a top guide 48. Core 44 is controlled by control rods which are 
located within an assemblage of control rod guides 50. The control rod 
drive hydraulic lines and motors are accessed through the bottom of the 
vessel 20 as represented generally at 52. Surmounting the core 44 and a 
portion of the control rod guide assemblage 50 is a cylindrical shroud 54 
which functions to direct the circulation of water coolant within the 
vessel 20. In this regard, water is forced to flow downwardly along the 
annulus between shroud 54 and vessel 20 (downcomer region), whereupon it 
is directed upwardly through the core 44 in somewhat continuous fashion. 
FIG. 2 additionally schematically portrays the reactor water recirculation 
system with which the LPCI modification of the invention performs. The 
function of this reactor water recirculation system is to circulate the 
required coolant and moderator through the reactor core. Consisting of two 
loops or divisions represented generally at 56 and 57, external to the 
reactor vessel, the system provides a pump for each loop within the 
interior of drywell 22 as represented, respectively, at 58 and 60. Each of 
the pumps 58 and 60 is structured in conjunction with a directly coupled 
water-cooled motor along with a variety of valves, here shown, for the 
purpose of simplicity, as recirculation discharge valves depicted, 
respectively, at 62 and 64, and respective suction or shut-off valves 66 
and 68. In general, the redundant recirculation loops 54 and 56 provide 
for circulation flow through the core 44, taking suction from the downward 
flow in the annulus or downcomer region between the core shroud 54 and the 
wall of vessel 20. About one-third of the core flow is taken from the 
vessel 20 through the recirculation loops 56 and 57. Within these loops, 
it is pumped to a higher pressure, distributed through a manifold through 
which a number of pipes (not shown) are connected, and returned to the 
vessel 20, whereupon it is directed to a sequence of jet pumps, two of 
which are schematically depicted at 70 and 72. As the flow is directed to 
the initial stages of the jet pumps, momentum exchange induces the 
surrounding water in the downcomer region to be drawn into the jet pump 
throat where these two flows mix and flow is directed into the lower 
plenum of the reactor vessel 20. Flow then is redirected upwardly through 
core 44 for heat exchange. Feedwater is added to the system through 
spargers located above the downcomer annulus and joins the downward flow 
of water. 
The low pressure coolant injection function (LPCI) is a part of the 
residual heat removal system (RHR) and, upon the occasion of an LOCA, 
functions to inject water to the reactor core through the loops 56 and 57, 
a coolant input approach which is selected because these loops inherently 
will place such emergency coolant at the right location within vessel 20. 
Under such a loss of coolant condition, the normal operating pressure, for 
example 1,000 psi within the reactor vessel 20 will be released, in part, 
because of the excursion at hand, and through a depressurization system. 
At this occurs, four LPCI pumps of the emergency core cooling system are 
activated. These pumps, while dedicated to the LPCI, additionally may be 
used for cooling water in the suppression pool 26, as well as other RHR 
system functions (e.g. containment spray cooling, suppression pool spray 
cooling and shut down cooling). However, upon the occasion of a signal 
calling for LPCI action, they are dedicated to the injection of water 
within loops 56 and 57, as represented by respective input arrows 74 and 
75. Thus, coolant may be injected simultaneously at positions 74 and 75 
from the suppression pool 26 and condensate storage tank (not shown) 
representing water sources already available with the facility 10. In the 
past, however, accommodation necessarily has been made to a rule or 
condition wherein it must be assumed that one loop 56 or 57 is ruptured. 
Thus, a selection type valve logic has been imposed to select that loop 
which remains viable for LPCI utilization. As noted above, this 
requirement has imposed the need for highly complex controls and water 
diversion schemes. 
Turning to FIG. 3, a schematic diagram is provided representing a split 
loop injection modification (SLIM) for the low pressure coolant injection 
(LPCI) system extant in certain current nuclear power installations. In 
the figure, a reactor vessel 80 is seen contained within a drywell defined 
by the wall or structure 82. Additionally shown within the drywell region 
are recirculation loops identified by the earlier numeration 56 and 57. 
These loops are seen to be present redundantly, as before, incorporating 
respective pumps 58 and 60, suction valves 66 and 68, and discharge valves 
62 and 64. Below and encircling the drywell boundary 82 is a torroidal 
shaped suppression pool sectionally revealed in general at 84, and shown 
connected to an emergency core cooling system (ECCS) header 86 which is 
located within the shield building. Extending from the drywell wall 82 to 
the suppression pool 84 are downcomers 88 which direct liquids and/or 
steam released from a break thereinto. 
Now looking to the low pressure coolant injection (LPCI) components, the 
suction inputs of two pumps 90a and 91a are seen connected to the ECCS 
header 86 and, the suppression pool 84 via lines 92a and 94a. A motor 
operated suction side valve is connected to the input of each as shown, 
respectively at 96a and 98a. Valves 96a and 98a are normally open. In 
addition to being directed to the ECCS header 86, it may be noted that 
line 94a also is coupled via line 100a to the condensate storage tank 
represented at block 102. Tank 102 generally will be mounted somewhat 
remotely within the power plant facility and, for the instant LPCI system, 
represents an alternate source of water. The outputs of pumps 90a, 91a, 
are directed via respective lines 104a and 106a through a heat exchanger 
by-pass valve 108a to line 110a during such occasion as valve 108a is 
open. 
Flow from conduit 110a is directed to conduit 116a, whereupon its flow is 
controlled by an orifice 118a or equivalent hydraulic resistance, thence 
through an open LPCI valve 120a and to conduit 122a. With the opening of 
the valve 124a and the closure, inter alia, of valves 128a and 132a, the 
cooling flow is injected via conduit 136a to the recirculation loop 56 
downstream of discharge valve 62. On such occasion, discharge valve 62 is 
closed. The position of injection from line 136a and closure of discharge 
valve 62 follows a postulation that rupture will occur on the 
recirculation line. 
Under non-accident conditions where suppression pool cooling is required, 
valves 90a and 91a serve to supply a heat exchange loop function with the 
closure valve 108a and the by-passing of fluids from the suppression pool 
84 and header 86 via conduit 112a to a heat exchanger 114a. Upon cooling, 
the liquid flows therefrom to conduit 110a. Under conditions requiring 
pool cooling, the fluid will flow through line or conduit 110a to conduit 
116a, thence through an open LPCI injection valve 120a and to conduit 
126a. For these normal conditions, a next serially disposed LPCI injection 
valve 124a will be closed to effect a diversion of the cooled fluid 
through conduit 126a, cooling valve 128a, line 130a, serially coupled 
cooling valve 132a, and line 134a into the suppression pool as at 84. 
Conditions may be experienced where a high pressure condition exists above 
the water within the suppression pool as at 84. This high pressure 
condition will be steam induced and, accordingly, the pumps 90a and 91a 
may be employed to perform a quenching function. In this regard, it may be 
seen that conduit 140a extends from conduit 110a to a first suppression 
pool injection valve 142a, the output of which at line 144a is coupled to 
an next serially connected suppression pool spray injection valve 146a and 
thence via conduit 148a to a spray sparger 150a. 
In the event that a high pressure condition occurs within the containment 
interiorly of wall 82, then a similar form of quenching or spraying may be 
employed. Such quenching will function, under accident conditions, to 
protect the pressure boundaries of the overall containment scheme. 
Accordingly, conduit 110a is seen to extend to a containment spray valve 
154a, the output of which at line 156a is directed to a serially connected 
next containment spray valve 158a and thence via parallel lines 160a and 
162a to paired spray spargers 164a and 166a. 
The LPCI components thus far described in connection with recirculation 
loop 56 perform essentially symmetrically with respect to the 
corresponding components of recirculation loop 57. Such common components 
have been identified with the suffix "a" in the description presented 
immediately above for loop 56. Accordingly, the components common 
therewith for recirculation loop 57 are identified in the figure with the 
same numeration and a suffix, "b". In each case, the dual or redundant 
systems perform in conjunction with ECCS header 86, suppression pool 84, 
and condensate storage tank (CST) 102, and the description given above 
with respect to recirculation loop 56, applies to the corresponding 
components of recirculation loop 52. One additional feature of the LPCI 
system in the figure resides in a cross-tie line represented by conduit 
170a, cross-tie valve 172a, conduit 174, cross-tie valve 172b, and line 
170b. This cross-tie line has been used in earlier systems for various 
purposes including diverting the pumped flow from one side of the 
redundant system to the other. Without such control or, without the 
features of the instant system and method, water under pressure being 
injected under the system would follow the path of least resistance and 
exit from the ruptured recirculation loop while being ineffective in the 
intact loop. The complexities associated with earlier control systems are 
eliminated with the present invention. Under the arrangement of the 
instant invention, cross-tie valves 172a and 172b are left open 
continuously. Correspondingly, the size of orifice combination 118a and 
118b are selected to limit or restrict the flow of water coolant to a 
fluid rate delivering a predetermined quantity of water coolant to each of 
the independent recirculation loops 56 and 57 simultaneously. This rate is 
effective for carrying out the emergency cooling of the core reactor 80 
within time constraints required. The flow rate established through use of 
hydraulic resistance at each orifice 118a-118b is based upon a knowledge 
of first: (a) the quantity of fluid required to carry out necessary LPCI 
cooling of the core of reactor 80, and (b) the time available for carrying 
out emergency cooling. Under analysis, a sufficient quantity of this water 
will be present for this task because the coolant passing into the 
ruptured one of recirculation lines 56 or 57 is discharged to the drywell 
where the coolant evenutally flows into the main vent through the drywell 
downcomer back into the suppression pool. Thus, a correction technique is 
provided with elegant simplicity and which is readily available for the 
purpose of retrofitting existing BWR installations. Only a minimum amount 
of hardware perturbation is involved for retrofitting, rewiring procedures 
and the like as well as extensive repiping procedures not being required. 
In effect, the modified LPCI technique becomes one somewhat, passive in 
nature. The improved technique is referred to as a "split loop injection 
modification" (SLIM). 
Referring to FIG. 4, an LPCI SLIM logic diagram is presented. In the 
discussion below, common components of the system at hand previously 
identified with the suffix "a" or "b" are identified by the associated 
numeration only. In the figure, block and bracket 190 indicates sensing 
parameters wherein a safety output condition and corresponding initiation 
signal are generated. In this regard, as represented at logic diagram 
position 192, a low water level may be detected within reactor 80. This 
condition may obtain or, as represented at position 194, a high drywell 
pressure may be detected. These conditions represent the detection of a 
loss of coolant accident (LOCA). Upon their detection, then as represented 
by respective lines 196 and 198, the control logic proceeds to position 
200 wherein a standby diesel engine is started to activate an electrical 
power source for the pump motors as well as those motors or electrical 
drive components functioning to control the valves employed in this 
system. This region of the logic diagram is shown identified at block and 
bracket 202 as making power available to valve and pump motors and for the 
purpose of meeting the criteria of "permissives", the latter conditions 
being conditions set upon the depressurization or development of requisite 
low pressures, P1 and P2, with respect to reactor 80. 
Following the commencement of diesel engine derived power, then, as 
represented at lines 204 and 206, leading to respective positions 208 and 
210, a decision or determination is made as to whether power is available 
at the bus or buses serving to supply electrical power at identified LPCI 
pumps 90, 91 and valves as at 120, 124, and 62, 64. Accordingly, as 
represented at line 212 and position 214, a decision or determination is 
made as to the presence of power at the LPCI injection pumps 90, 91. 
Simultaneously, as represented at line 216 and position 218, a 
determination is made that power is available at LPCI injection valves 120 
and 124. Additionally, as represented by line 220 and position 222, a 
decision or determination is made as to whether power is now available at 
the discharge valves 62 and 64 of respective recirculation loops 56 and 
57. 
An additional condition is imposed with respect to the requirements 
represented at position 218 for powering LPCI injection valves 120 and 
124. As represented at lines 224, 226 and position 228 the pressure 
exhibited at reactor 80 must fall lower than a predetermined lower 
pressure P1. That pressure may, for example, be in the range of 300 psi to 
400 psi. In this regard, injection valves 120 and 124 are permitted to 
operate only on the occasion of a certain pressure differential across 
them. 
Another permissive condition within the logic disclosed at block and 
bracket 202 is represented at position 230 wherein a decision or 
determination is made as to whether the reactor pressure is less than a 
predesignated value P2. This value, P2, is less than pressure value, P1, 
being, for example about 200 psi. Where that condition is met, then the 
logic continues as represented at line 232. 
The logic diagram then continues to the region represented by block and 
bracket 234 providing for pump and valve actuation. It may be further 
noted that the logic diagram may be categorized vertically. For example, 
block and bracket 236 represent a region concerning the starting of the 
LPCI pumps 90, 91, while block and bracket 238 represent the vertical 
region of the diagram concerned with the opening of injection valves 120, 
124, and block and bracket 240 are concerned with the closing of the 
discharge valves 62, 64, of respective recirculation loops 56 and 57. Note 
accordingly, that upon the decision or determination of the presence of 
power at pumps 90 and 91 as represented at position 214, then as shown by 
line 242 and position 246, the LPCI pumps 90, 91 are ready for rated flow. 
This means that the pump operation is up to rated speed and is now 
delivering at rated capacity in terms of flow. The conditions of control 
require that a full flow condition be satisfied before the assumption is 
made that the LPCI system is available for the LOCA condition. 
Now looking to the presence of power at the LPCI injection valves 120 and 
124, and the meeting of the pressure permissive condition as represented 
at position 228, with the conjoint occurrence of these conditions as 
represented by ANDing lines 248 and 250, the logic flow continues as 
represented at line 252 and position 254 to identify the condition that 
the LPCI injection valves 120 and 124 are open. Similarly, it is also 
necessary that the discharge valves 62, 64 be closed as represented at 
position 256 within vertical region 240. This discharge valve closure for 
recirculation loops 58 and 60 is made with the postulation that a break in 
the recirculation loops will be on the suction side of those recirculation 
pumps. Thus, the injection of the LPCI system is directed at the discharge 
side of pumps 58, 60 as well as corresponding discharge valves 62, 64. 
Finally, it is necessary that the conditions represented by diagram 
positions 246, 254, and 256 occur conjointly. This requirement is 
represented by ANDing lines 258, 260, and 262. 
Since certain changes may be made in the above system and method without 
departing from the scope of the invention herein involved, it is intended 
that all matter contained in the above description or shown in the 
accompanying drawings shall be interpreted as illustrative and not in a 
limiting sense.