ML20114F830
| ML20114F830 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/16/1982 |
| From: | Horton W, Kirstein B, Lobner P SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20114F832 | List: |
| References | |
| CON-NRC-03-82-096, CON-NRC-3-82-96, FOIA-84-880 SAI-1382-147LJ-R01, SAI-1382-147LJ-R1, SAI01382-147LJ, SAI1382-147LJ, SAI1382-147LJ-R1, NUDOCS 8207220398 | |
| Download: ML20114F830 (135) | |
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SAIO1382-147LJ, Rev.1 A Preliminary Assessment of Core Melt Probability In Cold Shutdown Following a Postulated LOCA at the Sequoyah Nuclear Plant Peter Lobner William Horton Bruce Kirstein July 16, 1982 Prepared for U.S. Nuclear Regulatory Comission Pursuant to Final Work Assignment 5 Issued Under NRC Contract No. NRC-03-82-096 32.rA7tto 39 8g .
TABLE OF CONTENTS Section Page 1 INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . . . 1 2 TRANSITION TO COLD SHUTDOWN FROM POWER OPERATION. . . . . . . 3 3 POTENTIAL PWR SAFETY CONCERNS RELATED TO A LOCA OCCURING DURING COLD SHUTDOWN . . . . . . . . . . . . . . . . 7 3.1 Description of Potential Mode 5 LOCAs. . . . . . . . . . 7 3.2 Role of the Operator in Mitigating a LOCA Occuring During Cold Shutdown. . . . . . . . . . . . . . 9 3.3 Decay Heat and Coolant Inventory Makeup Req u i reme n t s . . . . . . . . . . . . . . . . . . . . . . 10 3.4 RCS and RHR Overpressure Protection Requirements . . . . 11 3.5 Criticality Control Requirements During LOCA Response . . . . . . . . . . . . . . . . . . ._. . . . . 13 3.6 Containment Isolation Requirements During Mode 5 . . . . 14 4 DESCRIPTION OF SYSTEMS DURING COLD SHUTDOWN . . . . . . . . . 31 4.1 Front-Line Systems . . . . . . . . . . . . . . . . . . . 31 4.2 Support Systems. . . . . . . . . . . . . . . . . . . . . 35 4.3 Instrumentation Related to LOCA Detection in Mode 5 . . . . . . . . . . . . . . . . . . . . . . . . . 36 4.4 Technical Specification Requirements in Mode 5 . . . . . 36 5 EVENT TREES FOR RESPONSE TO MODE 5 LOCAs. . . . . . . . . . . 55 5.1 Event Tree General Assumptions . . . . . . . . . . . . . 55 5.2 Large LOCA L Event Tree . . . . . . . . . . . . . . . . 58 5.3 Large LOCA L Event Tree . . . . . . . . . . . . . . . . 59 5.4 Medium LOCA vent Tree . . . . . . . . . . . . . . . . . 59 5.5 Small LOCA Event Tree. . . . . . . . . . . . . . . . . . 60 6 FAULT TREES FOR RESPONSE TO MODE 5 LOCAs. . . . . . . . . . . 69 7 QUANTIFICATION OF THE EVENT TREES . . . . . . . . . . . . . . 71 7.1 Data for Quantification. . . . . . . . . . . . . . . . . 71 7.2 Assumptions Regarding Quantification . . . . . . . . . . 71 7.3 Description of Calculations. . . . . . . . . . . . . . . 72 7.4 Quantification of Core Melt Probability. . . . . . . . . 73 7.5 Analysis cf Resul ts. . . . . . . . . . . . . . . . . . . 77 7.6 Treatment of Common Mode Failure . . . . . . . . . . . . 77 7.7 Uncertainty in Analysis. . . . . . . . . . . . . . . . . 73 i
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TABLE OF CONTENTS (CONTINUED)
Section Page 8
SUMMARY
, CONCLUSIONS,'AND RECOMMENDATIONS. . . . . . . . . . . 93 8.1 Sumary and Conclusions . . . . . . . . . . . . . . . . . 93 8.2 Recommendations . . . . . . . . . . . . . . . . . . . . . 94 9 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . 99 APPENDIX A - FAULT TREES APPENDIX B - ESTIMATION OF LOCA LEAK RATES e
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- 1. INTRODUCTION This report presents the results of an analysis to conservatively estimate the probability of a core melt accident in cold shutdown (Mode 5) initiated by a postulated loss of coolant accident (LOCA) at the Sequoyah Nuclear Plant. The terms " cold shutdown" and " Mode 5" are synonymous and, at Sequoyah, refer to the plant state when: (a) average reactor coolant tempera-ture is less than or eq'ual to 2000F, (b) the reactor is shutdown with Keff less than 0.99, (c) percent rated core thermal power, excluding decay heat, is zero, and (d) the reactor vessel head bolts are fully tensioned (Ref.1).
Event trees were used to define the Sequoyah plant response to small, medium and large LOCAs. Simplified fault trees further defined the equipment failures and operator errors that were associated with each event in the event trees. Using data mostly from WASH-1400 (Ref. 2), the fault and event trees were quantified by means of hand calculations.
A total of 20 cases were analyzed with varying assumptions regarding the LOCA initiating event (safe shutdown earthquake or operator error), time of LOCA initiation following reactor shutdown, LOCA size, availability of offsite power during a 100-hour period following the LOCA, and maintenance status. With no significant maintenance in progress at the time of LOCA initiation, the probability of core melt in cold shutdown for these 20 cases was estimated to be in the range from 3.96 x 10-5 to 1.14 x 10-7 per reactor-year. If maintenance affecting one electrical or cooling water support system train is in progress, the probability of core melt for the 20 cases increases and is estimated to be the range from 7.53 x 10-5 to 8.46 x 10-6 per reactor-year. In contrast, an overall core melt probability of 6 x 10-5 per reactor-year was reported in WASH-1400 (Ref. 2). It could be noted, however, that the estimates of core melt probability presented in this report are based on a number of assumptions and simplifications which, taken as a group, should yield very conservative results. The estimates of core melt probability in Mode 5 should therefore be considered as upper limits. A more detailed analy-1 w
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sis would most likely indicate that the probability of core melt in Mode 5 is significantly lower.
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- 2. TRANSITION TO COLD SHUTDOWN FROM POWER OPERATION Cold shutdown (Mode 5) is usually a transitional plant operating state between power operations (Mode 1) and refueling (Mode 6). A represen-tative plot of reactor coolant average temperature versus time during a shut-down and cooldown is shown in Figure 2-1 (from Ref. 3).
Following reactor shutdown, initial cooldown and depressurization of the reactor coolant system (RCS) is accomplished by using the steam generators to transfer heat to the main condenser (e.g., by dumping steam) and pressur-izer spray and heaters to control RCS pressure. Four hours after reactor shutdown and initiation of cooldown, the Sequoyah RCS nominally will be at 3500F and 425 psig tref. 4). At this point the residual heat removal (RHR) system can be placed in operation to continue the cooldown.
At Sequoyah, there is an administative cooldown limit of 500F per hour using the RHR system, although a maximum cooldown rate of 1000F per hour
-is permitted by the Technical Specifications (Ref.1). A design basis for the RHR system is to. cool the RCS from 3500F to 1400 in 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (e.g., 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor shutdown). Considering the administrative cooldown limit and the need for a period of time to accomplish the transition to RHR cooling, it is likely that the Sequoyah plant could be in Mode 5 as soon as eight to ten hours after reactor shutdown.
' As a point of comparison to other pressurized water reactor (PWR) plants, Figure 2-2 (from Ref. 3) and Figure 2-3 (from Ref. 5) respectively
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illustate the average RCS temperature during cooldown of a B&W and a C-E '
plant. From these figures, it is estimated that six hours after react'or shutdown is the earliest time at which a PWR plant will enter Mode 5.
It should be noted that cold shutdown may also be a steady-state plant condition during maintenance, testing, inspection or repair activities that cannot be conducted during other plant cperating modes. Such activities include reactor coolant pressure boundary repairs (e.g., steam generator repair or replacement, pipe crack repairs) which may require the extended maintenance of a cold shutdown condition.
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C Initiate two hour nold 0 Initiate Decay Heat System Operation I f # e i e t i i I l 2 4 6 8 10 12 14 16 18 20 22 24 Time, nours Figure 2-1. Reactor Coolant Temperature vs. Time Plot for Typical Shutdown From Full Power and Cooldown (from B-SAR-205).
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Figure 2-2. B & W Reactor Coolant Temperature Followin With Two RHR Looos in Operation (top curve)g and With Reactor One RHRShutdown Loop in Operation (bottom curve). (From B-SAR-205) 5
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o" g' 350 1
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13 0 O 20 40 60 80 100 120 140 160 TIME AFTER SHUTDOWN, HOURS Figure 2-3. C-E Reactor Coolant Temperature Following Reactor Shutdown With Two RHR Loops in Operation (top curve) and With One RHR Loop in Operation (bottom curve). (From San Onofre 2 & 3 FSAR) 6 l
- 3. POTENTIAL PWR SAFETY CONCERNS RELATED TO A LOCA OCCURRING DURING COLD SHUTDOWN In comparison to power operation, cold shui '3wn is a relatively I benign plant opreating mode. The reactor core is already shutdown, decay heat is less than one percent of full power and the stored energy in the RCS has i been largely dissipated during cooldown and depressurization. There are,
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however, a variety of safety concerns that may be associated with a LOCA that is initiated while the plant is in cold shutdown. Included are the following:
l e Dependence on the operator for initiation of protective actions, '
l e Core decay heat removal and coolant inventory control requirements following LOCA, o RCS and RHR overpressure protection requirements during response to some LOCAs with high-head coolant injection pumps,
-o Criticality control requirements if extended makeup is provided to the RCS from an inadequately borated water source, t
e Relaxed containment integrity requirements in cold shutdown.
This section describes potential LOCA sources during cold shutdown and provides background information related to each of the potential safety concerns listed above.
' 3 .1 - DESCRIPTION OF POTENTIAL MODE 5 LOCAs Potential Mode 5 LOCA sources are summarized in Table 3-1. The random pipe break LOCAs listed in this table are also potential LOCA sources in other modes of plarit operation. Most of the LOCAs due to system mis-l-
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alignment by an operator are unique to periods when: (a) the RHR system is in operation, or (b) personnel are performing testing or maintenance inside containment. A LOCA due to stuck-open RHR shutdown cooling suction line safety valve can only occur when the RHR system is aligned for shutdown core cooling. A LOCA could also be induced by an unmitigated pressure transient l and failure of overpressure protection systems. In Mode 5, upper limits for l
operating pressure are established by the RHR system design pressure (600 psig) and by RCS temperature and pressure limits which will be discussed in j detail in Section 3.4. When RCS cooldown or heatup is not in progress, the l RHR system pressure limit is more restrictive than the RCS temperature-pres-sure limits down to an average coolant temperature of approximately 1250F.
! For the purpose of analysis, Table 3-2 defines three Mode 5 LOCA categories: large, medium and small. The flow rate selected as the dividing point between medium and large LOCAs is related to the capacity of the high-head coolant injection pumps at RHR system design pressure. This flow rate would be approximately equal to the respective pump runout flow rates listed in Table 4-1 (e.g., 550 to 650 gpm). If coolant loss from the RCS at 600 psig l would be equal to or greater than the capacity of the respective high-head pump, the RCS cannot be pressurized above RHR system design pressure and further degradation of RCS pressure boundary integrity due to overpressuriza-tion would not be expected. If, on the other hand, the coolant loss from the RCS at 600 psig is significantly less than the capacity the respective high-
, head pump, the RCS could be pressurized to a point where a balance is achieved between pump capacity (decreases as RCS pressure increases) and coolant loss via the LOCA (increases as RCS pressure increases). The potential may exist for further RCS pressure boundary degradation due to overpressurization. This subject is discussed in more detail in Section 3.4.
The flow rate selected as the dividing point between medium and small LOCAs is related to the decay heat rate expected following extended operation of the reactor at full power. See Section 3.3 for details regarding Sequoyah decay heat generation rates.
To gain some insight into the break sizes that fall into each LOCA category, calculations were performed to estimate the leak rate from various size pipe breaks when the RCS was at 600 psig. The code listing and de-tailed results of the calculations are included in Appendix B. A summary of these results are listed in Table 3-3 and are illustrated graphically in 8
Figure 3-1. For each size pipe break, two different leak rates are listed in Tabl e 3-3. The first value is for a guillotine break that occurs one foot down a smaller diameter pipe that is connected to RCS loop piping. For this postulated LOCA geometry, kinetic or " entrance" losses have a greater impact on leak rate than frictional losses. Choke flow conditions were assumed not to exist thus, the calculated flow rates are the maximum possi-ble flows. The second value in Table 3-3 is for a guillotine break that occurs 100 feet down a straight, smaller diameter pipe that is connected to RCS loop piping. As can be seen in Table 3-3, frictional losses due to flow in the long smaller diameter pipe can have a significant ef fect on a LOCA leak rate. The effect is greatest in the smallest diameter piping where the leak rate is a factor of 3 to 4 less than the leak rate from a one foot stub pipe.
For a guillotine break one foot from the interface with the RCS loop piping, the three Mode 5 LOCA categories would include the following piping:
s Large LOCA - pipe diameter > 1 inch e Medium LOCA -
0.5 inch < pipe diameter i 1 inch e Small LOCA -
pipe diameter 1 0.5 inch Most of the piping listed in Table 3-1 would therefore be considered as potential large LOCA ' sources in Mode 5.
3.2 ROLE OF THE OPERATOR IN MITIGATING A LOCA OCCURRING DURING COLD SHUTDOWN Unlike power operation, there are few protective actions that will occur automatically following a LOCA in cold shutdown. It must also be noted, however, that few protective actions are required to maintain the plant in a safe condition following a LOCA in cold shutdown, and considerable time is usually available to accomplish necessary manual or remote-manual protective actions.
Required operator actions are included in the event trees and fault trees presented in Section 5 and Appendix A, respectively. Time lines are included in Section 5, and results of calculations are summarized in this section to provide insight into the time available for necessary operator 9
l actions following a LOCA in cold shutdown. It is assumed that the operator l has the capability to detect basic symptoms of a LOCA i n Mode 5 (e.g., l decreasing or zero pressurizer level and RCS pressure) and to initiate a timely response. i 3.3~ DECAY HEAT AND COOLANT INVENTORY MAKEUP REQUIREMENTS As discussed in Section 2, six hours following reactor shutdown is the earliest time that a PWR would be expected to reach a cold shutdown condition. Following extended full power operation, core decay heat gener-ation rate six hours after reactor shutdown would be approximately one percent of rated core thermal power. For the Sequoyah nuclear plant, this equates to about 34 MWt. At 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (30 days) after reactor shutdown, decay heat generation rate would be approximately 5 MWt. Table 3-4 provides a summary of the estimated decay _ heat generation rate following extended full power oper-ation of the Sequoyah nuclear plant. Both fission product and heavy element
- decay contribute to the total decay heat rate listed in Table 3-4 and plotted in Figure 3-2.
L Following plant cooldown and establishment of a relatively constant RCS temperature in cold shutdown, all sensible heat from the reactor coolant and RCS and core internal structures has been transferred to the ul ti mate heat sink. . Decay heat production continues as described above. The inte-grated heat transferred by an RHR system to the ultimate heat sink following shutdown of a 3800 MWt PWR is shown in Figure 3-3 (from Ref. 3). This figure illustrates that, for shutdowns as short as a few days, the majority of the core decay heat is produced after cold shutdown conditions have been es-tablished.
. Having an estimate of decay heat generation rate, calculations were performed to estimate the time it would take to boil down to the core mid-plane following a large LOCA which initially drops coolant level to the elevation of the RCS hot leg nozzles. Approximate water volumes in the j
reactor vessel are shown in Figure 3-4. To drop coolant level to the core 6
mid-plane, about 64,650 pounds of water must boil off, requiring 64.7 x 10 BTU assuming an initial water temperature of 2000F. Assuming that all decay heat is transferred into the reactor coolant, Table 3-5 lists times to boil down to the core mid-plane as a function of time following reactor shutdown.
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Under the stated assumptions, the times in Table 3-5 represent minimum times for uncovering the core to the mid-plane. Actual times may be longer.
The calculated times in Table 3-5 are not particularly sensitive to initial coolant temperature. If coolant temperature was initially 1400F, the time to boil down to the core mid-plane would only be increased by about 15 percent. -
> This study assumes that the onset of core melt will occur after coolant level has dropped to the core mid-plane. The times listed in Table 3-5-will be used to as an estimate of the time available for the operator to initiate protective actions to prevent core melt.
4 Also included in Table 3-5 is an estimate of the makeup rate neces-sary to match coolant boil off. If at least this makeup rate can be provided following a LOCA, reactor coolant level can be stabilized or increased. A single centrifugal charging or safety injection pump can provide adequate makeup flow at pressures in excess of RCS design pressure in Mode 5. In I addition, a single RHR pump can provide the required makeup flow when RCS pressure remains below RHR pump shutoff head (about 195 psig, see Table 4-1).
This study does not take credit for the reciprocating charging pump because its makeup capacity (98 gpm, see Table 4-1) is less than the makeup require-ments listed in Table 3-5 during the first two days following a shutdown.
3.4 RCS AND RHR OVERPRESSURE PROTECTION REQUIREMENTS A variety of transients initiated in cold shutdown can cause an RCS pressure excursion. Table 3-6 lists the types of transients that are usually considered in the design of the overpressure protection features of the RHR system. Rates of RCS pressure rise for these transients are listed in Table l
3-6 and are illustrated graphically in Figure 3-5 (from Ref. 3). These values i apply to a 3800 MWt B&W standard plant, but serve to illustrate that high-head coolant injection pumps are capable of rapidly increasing RCS pressure during
> cold shutdown.
As discussed previously, RHR system design pressure (600 psig) and the RCS temperature-pressure limits establish the upper limit for RCS pressure during cold shutdown. The Sequoyah RCS cooldown temperature-pressure limita-tions are shown in Figure 3-6 (from Ref.1). It'is likely that coolant
.- injection in response to a LOCA in Mode 5 wi.11 cause a cooldown of the RCS.
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e--, , . , . - , . - - - - - , - - - , , - - - , . . - - - . - . _ , - - - - -
Allowable combinations of RCS pressure and temperature for specific cooldown rates are below and to the right of the limit lines in Figure 3-6. As is
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evident in this figure, pressure limits become more restrictive as the rate of RCS cooldown increases. A temperature-pressure limit curve for RCS~ heatup is shown in Figure 3-7 (from Ref.1). These graphs define limits to assure prevention of nonductile failure.
RHR system and RCS overpressure protection is provided by a single relief valve on the shutdown cooling suction line from the RCS. This valve is
! capable of relieving the combined capacity of all charging pumps at the relief valve setpoint pressure of approximately 600 psig. Only one centrifugal charging pump is normally operating during cold shutdown. Further overpres-sure protection for the RHR system is provided by the redundant shutdown cooling suction line isolation valves which close automatically if RCS pres-sure increases above 600 psig. Once an isolation valve is closed, the RHR suction safety valve is isolated from the RCS, which now is protected only by.
power-operated relief valves (PORVs, 2350 psig setpoint) and pressurizer safety valves (2485 psig setpoint). These setpoints are far in excess of the limits imposed by the temperature-pressure curves in Figures 3-6 and 3-7. The operator could, however, remotely open a PORY to provide RCS overpressure protection in this case.
Medium and small LOCAs during cold shutdown have been defined in Table 3-2. For these types of LOCAs, a centrifugal charging pump or a safety I
injection pump can provide makeup at a rate that exceeds the leak rate at an RCS pressure of 600 psig (e.g., 550 to 650 gpm). The high-head pumps can eventually restore RCS coolant inventory, and pressurize the RCS to a point i
. where makeup and leak rates become equal. Maximum RCS pressure will be re- 1
)
lated to: (a) performance of the RHR overpressure protection features and operator actions related to overpressure protection, (b) the respective pump characteristic curve (see Section 4), and (c) the LOCA pipe break characteris-tics. With successive failures of overpressure protection features, it is I possible that RCS pressure could be driven above the applicable temperature-pressure limit curve during response to a medium or small LOCA with a high-head coolant injection pump.
For this study, it will be assumed that an unmitigatible breach will result following overpressurization of the RCS and/or RHR system during re-sponse to a medium or small LOCA with a high-head coolant injection pump. A core melt is then assumed to be inevitable.
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3.5 CRITICALITY CONTROL REQUIREMENTS DURING LOCA RESPONSE Prior to the LOCA, the reactor coolant is adequately borated to maintain Keff less than 0.99 (shutdown margin greater than 1.0 percent delta k/k), as required by the Technical Specifications (Ref.1). Nominal Mode 5 RCS boron concentration is in the range from 945 ppm (all rods inserted) to 1,016 ppm (BOL, one rod stuck out). Boron reactivity worth in cold shutdown is i approximately 1.0 percent delta k/k per 70 ppm boron.
! It should be evident that extended RCS makeup from an inadequately
- and an inadvertent criticality following a Mode 5 LOCA. In this study, it ~ is assumed that reactor coolant makeup is provided from the refueling water storage tank (2,000 ppm boron), or is provided from some alternative water
. source such as the primary water storage tank (0 ppm boron) only after proper boration via the chemical and volume control system (CVCS) boric acid blender.
It _is estimated that the continuous makeup rate from the latter water source is limited to 150. gallons per minute because of the need to provide adequate boration. Coordinated manual actions may be necessary to maintain an adequate supply of concentrated boric acid to meet long-term boration requirements.
! Other borated. water sources such as the cold leg accumulators and the upper head injection accumulator (all approximately 2000 ppm boron) could be aligned to provide limited makeup to the RCS. These water sources are not modeled in the event trees in Section 5 because-they have only a very limited capability to prevent a core melt.
As discussed previously, it is likely that coolant injection in response to a LOCA in Mode 5 will cause a cooldown of the RCS. In the reactor coolant temperature range of concern (e.g., less than 2000F). The moderator temperature coefficient of reactivity is numerically much smaller than during
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power operation. With a coolant boron concentration of 1000 ppm, the modera-tor temperature coefficient would be approximately -0.25 x 10-4 ap/0F (see Figure 3-8). If Keff were initially 0.99, an RCS cooldown from 2000F to 1000F would add 0.25 x 10-2 ap reactivity, resulting in a final Keff of 0.9925. A F significant cooldown of the RCS will therefore only affect the margin of subcriticality of the core, and should not result in the core becoming criti-l cal.
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l 3.6 . CONTAINMENT ISOLATION REQUIREMENTS DURING MODE 5 There are no Technical Specification requirements that containment !
isolation be maintained at Sequoyah during cold shutdown. A preliminary survey of several PWRs, conducted as part of this study, suggests that it is common practice for containment equipment hatc;hes to be open and both airlock doors open simultaneously (e.g., the normal interlock on the doors is bypas-sed) during cold shutdown to facilitate the movement of personnel and equip-ment into and out of containment. The initial opening of an airlock or equipment hatch -does not occur immediately upon reaching cold shutdown, but rather at some later time associated wi6h planned activities inside contain-ment. A number of plants leave the equipment hatch open whenever that are in Mode 5 ~ to simplify access requirements, and in some cases, to help maintain habitable containment temperatures. Other plants, including Sequoyah, open the equipment hatch only when it is necessary to move large equipment that cannot. be brought through a personnel airlock. All plants in this preliminary survey maintained at least one airlock open in Mode 5.
Equipment hatches are not necessarily designed for rapid closure.
Some plants, including Sequoyah, have equipment hatches that open inward and are moved away on tracks with the aid of a dedicated winching system. It may be possible to reclose such hatches in 30 to 60 minutes. At least one plant has an equipment hatch that opens outward and is removed to a nearby laydown area with the aid of a crane. At this plant, it was estimated that four to eight hours would be required to reclose the hatch. When an equipment hatch is reclosed in an' emergency, there would be no means to verify the leak tightness of the hatch seal (e.g., a containment leak test could not be conducted). The value assumed for containment leak rate following restoration of containment integrity may therefore be uncertain.
Temporary service lines (e.g., welding cables) may occasionally be run through a personnel airlock. In spite of this type of obstruction, it appears reasonably certain that airlocks can be reclosed much more quickly than equipment hatches. Quick disconnect fittings may be used on some service fines, thereby providing a capability to rapidly reclose at least one airlock
- door.
t It is conservative to assume that the probability of having a non-j isolated containment in Mode 5 is one (e.g., containment isolation require-
- ments will be relaxed). After a Mode 5 LOCA, the time available to restore i
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c containment isolation before the onset of core melt will be a function of the time after reactor shutdown and the actual core power history. Minimum times for protective actions by the operator are discussed in Section 3.3. At Sequoyah, an open airlock could very likely be reclosed when required. As stated prevsiouly, the Sequoyah equipment hatch is only opened when required to move large equipment. The equipment hatch design appears to provide a relatively rapid reclosure capability, however, the time required to reclose the hatch (e.g., 30 to 60 minutes) is comparable to the minimum time available for operator protective actions during the interval from eight to 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> after reactor shutdown (e.g., 39 to 62 minutes). Significantly greater time is available if the opertor is successful in establishing an effective short-term coolant injection capability (see Section 5.2).
This study did not include containment response in the LOCA event tr'ees in Section 5; therefore, no assumptions were made in the analysis re-garding the initial containment isolation status or the capability to restore containment isolation.
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Table 3-1. Potential LOCA Sources in Mode 5.
b) Piping connected to RCS*
e RHRshutdowncoolingsuctionline(14")
e Pressurizer surge line (14")
e ECCS cold leg injection lines (4x10")
e Upper head injection accumulator lines (4x8")
e RHR and/or safety injection hot leg injection lines (4x6")
e Pressurizer safety valve lines (3x6")
e Pressurizer power-operated relief valve (PORV) header (6")
e Pressurizer spray lines (2x4")
e Individual PORV lines (2x3")
e CVCS normal charging line (3")
e CVCS alternate charging line (3")
e CVCS high pressure letdown line (3")
e RCS loop flow / temperature sensor return line to RCP saction (4x2")
.e RCS loop flow / temperature sensor supply line (8x2")
,' e RCS loop low point drains to RC drain tank (4x2")
l e Pressurizer auxiliary spray supply line (2")
e Charging. pump safety injection lines (4xl-1/2")
e CVCS excess letdown line (l")
e Miscellaneous vent, drain, sample lines and sensor fittings (3/4")
- 2. LOCA due to system misalignment by operator a Inadvertent alignment of RHR pump to RHR containment spray line (8")
b Inadvertent alignment of RHR pump to RWST return line (8")
c Inadvertent opening of PORY (3")
d Inadvertent, alignment of RCS to RC drain tank (2" to 3/4")
e Inadvertent opening of miscellaneous vent, drain or sample lines or sensor fittings (3/4")
a Loss of RHR heat removal capability-b CVCS makeup valves stuck fully open c Low pressure letdown valve fails shut when RCS is solid, with continued operation of charging pumps d) Pressurizer heaters inadvertently energized when RCS is solid e) Inadvertent safety injection pump startup
- 4. LOCA due to unmitigated pressure transient and pressurization of RCS well above the pressure / temperature limit curve, causing RHR' pipe breach or reactor vessel failure.
a) Inadvertent alignment of upper head injection accumulator to RCS b) Inadvertent pressurization of RCS using charging or safety injection pumps Notes:
- The RCS and connected piping are illustrated in Figure 4-4.
0 16 m sa * . , .
i l
Table 3-2. Definition of Mode 5 LOCA Size Categories.
1 LOCA Category Definition Large LOCA Rate of coolant loss from the LOCA exceeds the capacity of a single centrifugal charging pump or safety injec-tion pump when the design presswe.ta)RCS .
is assumed Primary to be at safety concern is RHR main-system taining adequate core coolant inventory. The high-head pumps cannot repressurize the RCS significantly above RHR system design pressure.
Medium LOCA Rate of coolant loss from the LOCA is less than the capacity of a single centrifugal charging pump or safety injection pump, when t!)e at RHR system design pressure (a)RCS
, but is assumed the rate of co9 1- to be ant Mss is sufficient to remove all core decay heattb).
Opernion of the RHR system in the RHR mode is not re-quired. Overpressurization of the RCS and the RHR system by the high-head pumps is a potential safety conern.
Small LOCA Coolant loss at a rate that is insufficient to remove all core decay heat (b, c). Heat removal from the RCS, in addition to that provided by the LOCA itself, is re-quired. Overpressurization of the RCS and the RHR by the high-head pumps is a potential safety concern.
Notes: (a) About 550 to 650 gpm at approximately 600 psig RCS pressure.
(b) About 160 gpm boil-off rate at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after shut-down (assuming RCS boiling at atmospheric pressure with no makeup), decreasing as a function to time.
(c) There may be a time following reactor shutdown beyond which no LOCA size fits this definition (e.g., decay heat generation rate has decreased to such a level that an additional core heat removal capability, beyond the LOCA itself, is no longer required)..
17 ;
Table 3-3. Estimated LOCA Leak Rate at 600 psig RCS Pressure.
I Leak Rate in'gpm, as a Function of Actual Pipe Length of Pipe Between RCS Loop Inside Diameter Piping and the Break Location (inches)'
1 foot 100 feet-0.5 146 33 0.75 345 91 1 628 186 1.5 1,442 499 2 2,579 995 3 5,797 2,582 4 10,250 5,024 6 22,990 12,680 8 40,950 24,280 10 63.380 39,870 14 100,300 75,990 '
18
Table 3-4. Estimated Sequoyah Nuclear Plant Decay Heat Rate.
Decay Heat Rate (106 8TU/hr)* Oecay Heat Power (P)
Time After As a Fraction of Shutdown . Fission Heavy Design Power (Po)**
(hours) Product Element Total P/Po 1 162.91 20.73 183.64 1.58x10-2 2 129.10 17.91 147.01 1.26x10-2 3 115.20 17.25 132.45 1.14x10-2 4 106.08 16.97 123.05 1.06x10-2 5 98.82 16.75 115.57
-3 9.93x10 6 92.71 16.53 109.24 9.38x10'3 7 87.51 16.33 103.84 8.92x10'3 8 83.G6 16.14 99.20 8.52x10'3 9 79.24 15.94 95.18 8.18x10'3 10 75.96 15.74 91.70 7.68x10-3 11 73.13 15.56 88.69 7.62x10'3 12 70.66 15.36 86.02 7.39x10'3 20 - -
74.80 "
- 6.43x10'3 24 56.23 13.26 69.49 5.97x10~3 48 46.99 9.87 56.86 4.88x10~3 72 41.61 7.36 48.97 4.21x10'3 96 37.89 5.47 43.36 3.73x10'3 120' 35.24 4.08 39.32 3.38x10'3 144 33.25 3.04 36.29 3.12x10-3 168 31.71 2.26 33.97 2.92x10-3 192 30.46 1.68 32.14 2.76x10'3 216 29.40 1.26 30.66 2.63x10'3 240 28.48 0.93 29.41 2.53x10'3 400 23.95 0.13 24.08 2.07x10'3 500 21.84 0.03 21.87 1.88x10'3
-3 600 20.06 0.01 20.07 1.72x10 700 18.57 0.0 18.57 1.60x10'3 720 18.30 0.0 18.30 1.57x10'3 Notes:
- Adapted from 8-SAR-205, Tables 9.4-2 and 9.4-3 assuming 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of reactor operation at the Sequoyah Nuclear Plant design power level of 3411 Mwt.
" Sequoyah design power P , is 11,641.7 x 1008TU/hr (Sequoyah FSAR Table 4.491 ).
- From Sequoyah FSAR. Table 5.5-8.
I 19
Table 3-5. Estimates of Time to Boil Down to Core Mid-plane Following LOCA and Makeup Rate to Match Boil-off.
Estimated Estimated Time to Boil Makeup Rate Time After Decay eqt Down to Required to Shutdown Rate al Core Mid-plane (b) Match Boil-off(c)
(hours) (106 BTU /hr) (minutes) (gpm) 6 109.24 36 197 7 103.84 37 187 8 99.20 39 179 9 95.18 41 171 10 91.70 42 165 11 88.69 44 160 12 86.02 45 155 20 74.80 52 135 24 69.49 56 125 35 62.50(d) 62 113 48 56.86 68 102 72 48.97 79 88 96 43.36 90 78 120 39.32 99 71 144 36.29 107 65 168 33.97 114 61 192 32.14 121 58 400 24.08 161 43 720 18.30 212 33 Notes: (a) From Table 3-4.
(b) Assumes reactor coolant is initially at 2000F, and boils at atmospheric pressure.
(c) Assumes reactor coolant is at 2120F, boiling at atmospheric pressure, and makeup water is initially at 700F.
(d) Estimated from Figure 3-2.
20
Table 3-6. Pressure Rise and Relief Valve Capacity *.
RATE OF PRESS. REQUIRED RELIEF INCIDENT DESCRIPTION RISE, PSI / MIN CAPACITY, GPM**
Loss of DHR System Cooling 12 225 Makeup Control Valve Fatis Full Open 36 524 All Pressurizer Heaters Energized 5 1435 High Pressure Injection (HPI) actuation (all three HPI pumps operate) 162 2000 EOUILIBRIUM PRESSURE, PSIG Core Flood Tank Outlet Valve Opened At Pressurizer At DHR Pump Suction Initial Pressurizer Pressure at Midpoint of Band 432 455 Initial Pressurizer Pressure at High Point of Band 474 497 Notes:
- From B.SAR-205
- At a setpoint of 455 psig, the minimum required capacity is 2000 gpm at 10% accumulation. This relief valve will prevent the DHR system design pressure from being exceeded by more than 10% during the worst incident concurrent with the DHR pump operating at any developed head up to and including shutoff head. Each of the dual DHR suction lines contains a relief valve sized for the full relief flow rate in the event that the DHR system is being oDerated on one letdown line only.
i 21
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- 4. DESCRIPTION OF SYSTEMS DURING COLD SHUTDOWN This section provides a brief description of the following systems at the Sequoyah nuclear plant that are related to LOCA initiation and response in co'ld shutdown:
e Reactor coolant system (RCS), o Residual heat removal (RHR) system, o Chemical and volume control system (CVCS), o High pressure safety injection (HPSI) system, o Class 1E AC electric power system, o Component cooling water (CCW) system, o Essential raw cooling water (ERCW) system. . The first four systems will be called " front-line" systems because of their direct role in LOCA response. The last three systems will be called " support systems" because they are required for the operation of the front-line sys-tems. Although not modeled in the event trees in Section 5, the cold leg accumulators and the upper head injection (UHI) accumulator are also briefly described in this section. ' 4.1 FRONT-LINE SYSTEMS A summary of design data for pumps capable of providing reactor coolant makeup is presented in Table 4-1. Characteristic curves for the centrifugal charging, residual heat removal and safety injection pumps are shown in Figures 4-1 to 4-3, respectively. Design and operating data for potential reactor coolant makeup water sources is presented in Table 4-2. 31 4.1.1 Reactor Coolant System The Sequoyah nuclear plant has a four-loop RCS as shown in Figure 4-
- 4. All interfacing piping larger than one inch in diameter appears in this ,
l diagram. Actual line sizes are listed in Table 3-1. In cold shutdown, the RCS can be maintained with a steam bubble in the pressurizer or it can be placed in a water solid condition. In either case, the CVCS controls RCS pressure, and makeup is provided by one centri-fugal charging pump. The general practice at Sequoyah is to maintain a steam bubble in the pressuriter when possible. 4.1.2 Residual Heat Removal System The RHR system at Sequoyah is shown in Figure 4-5 in its shutdown cooling alignment. This system has two loops which share a common shutdown cooling suction line. Two independent return paths to the RCS exist via motor-operated valves 63-93 and 63-94 During cold shutdown, letdown from the RCS to the CVCS is accomplished via a low pressure letdown flow path on the outlet side of the RHR heat exchangers. During shutdown cooling, the RHR system is protected against over-pressurization by relief valve 74-505 on the shutdown cooling suction line. This relief valve is sized to relieve the combined flow of all charging pumps at the relief valve setpoint of approximately 600 psig. During cold shutdown, only one centrifugal charging pump is in operation. Further overpressure protection is provided by the series shutdown cooling suction isolation valves 74-1 and 74-2 which close automatically if pressure exceeds 600 psig. In addition to its shutdown cooling function, the RHR system also can perform coolant injection and recirculation functions as part of the emergency core cooling system (ECCS). The RHR pumps can be aligned to provide low pressure coolant injection from the refueling water storage tank (RWST) by isolating the normal shutdown cooling suction path (e.g., by closing valve 74-1 or 74-2) and opening the RWST isolation valve 63-1. If the shutdown cooling suction path is not isolated and if RCS pressure is higher than the static head from the RWST, the RWST suction check valve 63-505 will seat, preventing the RHR pumps from taking a suction on the RWST (Ref. 6). The RHR pumps can also be aligned to provide low pressure recir-culation from the containment active sump to the RCS. Valves 63-1, 63-72 and 63-73 are interlocked to prevent the simultaneous alignment of an RHR pump to 32 the RWST and to the containment sump. The switchover from injection to re-1 circulation during cold shutdown would be accomplished manually when RWST level reaches a low-level alarm setpoint. Due to Sequoyah containment geometry, it is necessary to add a considerable amount of water from the RWST to ensure that the active con-tainment sump (bottom at elevation 667.0 feet, curb top at approximately 680.5 feet) is flooded for all LOCAs and that an RHR recirculation flow path can be ; established. There is also a pit sump (bottom at elevation 658.3 feet) be-neath the reactor vessel. Some break locations such 'as a break at a reactor vessel hot or cold leg nozzle may result in flooding of the pit sump before , any water reaches the containment floor at the 679.78 foot elevation and spills over an approximately eight-inch curb into the active sump. Rough calculations indicate that about 100,000 to 120,000 gallons of water will flood the pit sump. It was estimated by NRC inspectors that 105,000 gallons would fill the Sequoyah active sump and flood the contain-
- ment floor at the 679.78 foot elevation to a depth of about 18 inches (10 inches above the sump curb) if little or no water leaked to the pit sump.
It is therefore estimated that a minimum of 155,000 to 175,000 gallons of water must be dumped into containment to ensure that the active sump will be flooded and that an RHR recirculation flow path is available. A large LOCA will rapidly dump approximately 68,000 gallons (9091 ~ ft )3 to containment if the RCS was being maintained with a steam bubble in the pressurizer. Slightly more would be dumped if the RCS was in a water solid condition. Note that it may be possible to establish an unorthodox flow path between the RWST and the suction of an RHR pump by: (a) aligning a containment spray pump to the RWST by opening valve 72-21 or 72-22, (b) bypassing valve interlock circuitry, (c) simultaneously aligning the selected containment i spray pump to the containment sump suction header by opening valve 72-20 or 72-23, and (d) closing the RHR pump suction isolation valves 74-3 and 74-21. The RWST is then connected to the sump suction header via the containment spray system piping and it should then be possible to draw a suction on the RWST with the RHR pumps. ? r 33 j , , . , , . - - - . - - - - - . - .,,,,e,,.,,.w.cr,n,..,,ee-~w . , , - - - - . -nnw.._,..,,,-,--,,,.nn--..w,..,- - . - - . - ~ . , , . -,.,, -e,--,---, / 4.1.3 Chemical and Volume Control System The Sequoyah CVCS is illustrated in Figure 4-6. During cold shut-down, the high pressure letdown path used during power operation remains in operation and a low pressure letdown path from the RHR system is established. The maximum letdown flow rate is approximately 120 to 150 gpm. The normal charging return path to the RCS is also used during cold shutdown. Only a single centrifugal charging pump is operating. The control mode for the CVCS depends on whether the RCS is main-tained with a steam bubble in the pressurizer or is placed in a water solid condition. With a bubble in the pressurizer, the low pressure letdown valve establishes a fixed backpressure, and the ,ressurizer level controller auto-matica11y positions a makeup control valve to modulate makeup flow rate as necessary to control pressurizer level. In this mode of operation, makeup flow rate would automatically increase as pressurizer level oecreased during a LOCA. Small LOCAs may Le adequately controlled and automatic response would be limited only by the ability to provide extended makeup to the volume control tank (VCT) which would commence when VCT level drops to a low-level setpoint. It is dstimated that a long-term makeup rate of 150 gpm can be established from the primary water storage tank (PWST) and properly borated via the boric acid blender. Automatic makeup from this source is limited by the capacity of the b.oric acid storage tanks. If VCT level cannot be ade-quately maintained from the normal makeup source and level drops to a low-low level setpoint, the charging pump suctions will be automatica11'y shifted to
- the RWST by opening valve 62-135 or 62-136.
If the RCS is maintained in a water solid condition, the low pres-l sure letdown valve is again set to establish a fixed back pressure, and the j makeup control valve is set for a fixed makeup flow rate. During a LOCA, the l makeup flow rate would remain constant, and the RCS would rapidly depres-surize. The low pressure letdown valve would automatically close in an at-tempt to maintain RCS pressure. Makeup to the VCT would be provided auto-I matica11y as described before. The centrifugal charging pumps are part of the ECCS and can be aligned to a high pressure coolant injection flow path by opening valve 63-39 l or 63-40 and valve 63-25 and 63-26. This flow path feeds a single header which branches to inject into all four RCS cold legs. 1 I l 34 4.1.4 High Pressure Safety Injection System The HPSI system at Sequoyah is illustrated in Figure 4-7. The system has two loops which share a common suction line from the RWST. During cold shutdown, the circuit breakers for the HPSI pumps are racked-out to render the pumps inoperable. 4.1.5 - Cold Leg Accumulators Four c.old leg accumulators are provided, one for each of the four RCS cold legs. A typical cold leg accumulator is 'shown in Figure 4-8. During cold shutdown, the motor-starter for the motor-operated accumulator isolation valve is racked-out to render the~ valve incapable of remote operation. l 4.1.6 Upper Head Injection Accumulator - The UHI injection system injects directly into the reactor vessel f head, as shown in Figure 4-9. During cold shutdown, the redundant hydraulic isolation valves in the injection lines are closed, and are locked by means of motor-operated gags. When actuated, the UHI accumulator isolation valves remain open until the water accumulator reaches a low-level setpoint of approximately 3366 gallons (450 ft 3). The isolation valves close automatically to prevent in-jecting pressurized nitrogen into the RCS. 4.2 SUPPORT SYSTEMS A summary of support system requirements of major components is ! presented in Table 4-3. l 4.2.1 Class 1E AC Electric Power System l -- The Sequoyah Class 1E AC electric power system consists of two independent load groups or divisions as shown in Figure 4-10. Offsite power is the normal power source during cold shutdown. Following a loss of offsite power, each load group can be supplied from its own standby diesel generator.
- . All 6900 GC Class 1E loads are listed a Figure 4-10. Only the 480 VAC loads l directly related to LOCA mitigation in cold shutdown are included in that figure.
'35 4.2.2 Component Cooling Water and Essential Raw Cooling Water Systems Together, the CCW and ERCW systems establish the heat transfer paths from components requiring cooling to the ultimate heat sink. The CCW and ERCW systems at Sequoyah are shown in Figures 4-11 and 4-12, respectively. As illustrated, these systems serve both Units 1 and 2 at the Sequoyah site. A simplified model of a CCW and ERCW loop is shown in Figure 4-13. Although not an accurate model of the Sequoyah systems, this is likely to be a conservative model because it lacks much of the component redundancy and all of the built-in cross-connections actually found at Sequoyah. 4.3 INSTRUMENTATION RELATED ".0 LOCA DETECTION IN MODE 5 Table 4-4 lists instrumentation that may be of use in detecting a LOCA occurring in Mode 5. 4.4 TECHNICAL SPECIFICATION REQUIREMENTS IN MODE 5 Table 4-5 lists the limiting conditions for operation and surveil-lance requirements in NUREG-0789 (Ref.1) that are applicable during Mode 5. e 36 / s' -- .- ._ -. -. .-= Table 4-1. Design Data for Pumps Capable of Providing Reactor Coolant Makeup. 1' Centrifugal Reciprocating Residual Safety Boric Acid Charging Charging Heat Removal Injection Transfer , Pump Pump Pump Pump Pump a Number (per unit) 2 1 2 2 2 Type Horizontal Variable Speed. Vertical Horizontal Horizontal Centrifugal Positive Dis- Centrifugal Centrifugal Centrifugal Placement Design Flow (gym) 150 98 3000 425 75 Design Head (ft.) 5800 5800 375 2500 235 Runout Flow (gpm)* 550 - 5500 650 iicad at Runout Flow (ft.)* 1400 - 250 1500 Shutoff Heat (ft) 5800 - 450 3500 w (psid)** 2514 - 195 1517 Normal Status in Mode 5 One pump op- Pump not used. Operable, with Inoperable. 'allable for erating, one but operable. 2. I or 0 Circuit ,'4rmal bora-pump not used Circuit breaker operating breaker for tion or emer-but operable. not racked out, pump racked gency boration Circuit break- but control out, as necessary, er for latter switch tagged. pumping to pump not racked charging pump out, but con- suction, trol switch i tagged. 1 4 1 From RESAR-35, Westinghouse Reference Safety Analysis Report. Docket STN-50545
- Conversion is 1 foot water at 60*F = 0.4335 psi
Table 4-2. Design Data for Reactor Coolant Makeup Water Sources, upper Head Refueling hter Primary Water Boric Acid Cold Leg Injection , Storage Tank Storage Tank Storage Tank Accumulators Accumulator Number (per unit) 1 1 1.+1'(shared) 4 1 Type Atmospheric Atmospheric Atmospheric Pressurized Pressurized Nominal Wter Volume (gal)* 370,000 to 175,000'(est) 6.542 7,857 to 8,071 13,500 to 13,850 375,000 (1,805 to 1,851 ft 3) 4 Minimum hter Volume 35.443 --- 2,175- --- --- in Mode 5 (gal)** 4
- Operating Pressure (psig) --- --- --- 385 to 447 1,185 to 1,285 Boron Concentration (ppm) 2,000*** O 20,000 to 1,900 to 2,100 1,900 to 2,100 22.500 Nomal Status in Mode 5 Available as a Available as a Available for Inoperable. Single Inoperable. Re-a- Id motor isolation dundant hydraulic water source for water source for boration of CVCS RHR, charging, the chemical and makeup water from valve in each dis- valves in each safety injection volume control primary water charge line to RCS injection line shut and containment system (CVCS). storage tank, cold leg shut and and locked with spray pumps. Boration may be circuit breaker for motor-operated gag.
required. valve motor racked out.
- Modes I to 3 (all) and Mode 4 (RWST and BAST) as per NUREG-0789
- NUREG-0789, Limiting Condition for Operation 3.1.2.5
- 2,000 to 2,100 ppe in Modes 1 to 4 l
0 Table 4-3. Summary of Major Support System Requirements. AC Component Electric Cooling Water Power Essential Raw Load Heat Cooling Water Component Group Header. Exchanger Loop RHR Pump 1A-A 1A 1A A RHR Heat Exchanger IA - 1A A Centrifugal Charging Pumo IA-A 1A 1A A SI Pump 1A-A 1A 1A A CCW Punp 1A-A 1A 1A A CCW Pump C-S (Alt) 1A(alt) 1B C CCW Heat Exchanger A - 1A - IB or 2A ERCW Pump J-A 1A - - 1A or 2A ERCW Pump K-A (Unit 2 Pump) 2A - - lA or 2A ERCW Pump Q-A 1A - - 2A or lA ERCW Pump R-A (Unit 2 Pump) 2A - - 2A or lA Diesel GeneratorlA-A 1A - - 1A or 2B RHR Pump 18-B 1B IB C RHR Heat Exchanger IB - IG C Centrifugal Charging Pump 18-B 1B IB C Reciprocating Charging Pump 1B - - SI Pump 18-B IB 1B C CCW Pump 1B-B _1B 1A A CCW Pump C-S IB(norm) IB C CCW Heat Exchanger C - IB - 1A or 2B ERCW Pump L-B 18 - - IB or 2B ERCW Pump M-B (Unit 2 Pump) 2B - - IB or 2B ERCW Putp N-B 18 - - 2B or IB ERCW Pump P-B (Unit 2 Pump) 28 - - 2B or 1B
- Diesel Generator 1B-B IB - - 18 or 2A I l
! 39
- t. 1
Table 4-4. Sumary of Plant Variables That Could Provide Indication of a LOCA. Variable Response During LOCA Remarks Pressurizer levtl Decreasing level, or instrument I cold calibrated and 3 hot pegged low. calibrated channels. Low level alarm. RCS pressure, wide Decreasing pressure, rapid decrease 2 channels. No alam when pressure reage if RCS was initially water solid, decreases (already below low pres-sure alarm setpoint wnen in Mode 5). RHR Icop A/B flow rate Flow erratic or drops to zero if Low flow alarm on *miniflow* pump pumps cavitate or become airbound. recirculation line. CVCS letdown flow rate Flow decreases or drops to zero as No alarm low pressure letdown valve attempts to control RCS pmssure. CVCS makeup flow rate Increasing flow to control pressur- No alarm izer level if steam bubble was maintained in pressurizer. CYCS Volume Control Decmasing level when letdown flow Low level alarm automatically Tank (VCT) level decreases, and makeup flow constant initiates makeup. Iow-low level or increasing. alarm shifts charging pump suction to RWST. VCT makeup flow rate Primary water and boric acid flow Indivicual instruments monitor indicated when VCT level drops to primary water and boric acid makeup low level setpoint. flow ratas. No alarm. Remote position Unexpected valve position To determine if a flow diversion indication for indicated, path has been established due to selected valves operator error. Multiple valves served by comunon audible alarm with no reflash capability. RWST level Decreasing if charging pumps auto. Low level alarm matica11y realigned on low-low VCT level. Increasing if operator error establishes a return flow path from RHR system to the RWST. Containment pocket (or Alarm if pocket sump flooded. Part of RCS leak detection system. pit) sump level Single level sensor alarms in Incore Instrument Room. Not required in Mode $*. Reactor building floor Alarm if containment floor flooded. Sensor about 6 inches above floor water level level. RHR pump room sump Alarm if flooding occurs in RHR Extended operation of an airbound level pump room. RHR pump following a LOCA may lead to pump mechanical seal failure and pump room flooding. CVCS pump room sump Alarm if flooding occurs in CVCS level pump room. Containment temperature Increasing Available on plant computer. Containment atmosonere Increasing Part associated with vent isola-particulate activity tion operable in Mode 5. Containment atmosphere Increasing Part associated with vent isolation gaseous radioactivity operable in Mode 5. Containment purge Increasing Not required in Mode 5**. exhaust radioactivity
- $ee '4UREG-0789. LCO 3.4.6.1 *
- See fluRCG-07S3 LCO 1.3.3.1 40
- - + Table 4-5. Sequoyah Technical Specifications Applicable in Mode 5. Limiting Conditions Surveillance System For coerstion (LCO) Recuirements Remarks Peactivity 3.1.1.2 4.1.1.2 Shutdown margin h 3.1.2.1 3.1.2.3 4.1.2.1 4.1.2.3 Boration system flow paths Charging pumps 3.1.2.5 4.1.2.5 Berated mater sources 3.1.3.3 4.1.3.3 Rod position indication Instrtseen- 3.3.1 4.3.1 Source range instrumentation tation 3.3.3.1 4.3.3.1 Radiation monitoring 3.3.3.3 4.3.3.3 Seismic instrumentation 3.3.3.4 4.3.3.4 Meteorological instrumen-tation 3.3.3.6 4.3.3.6 Ch,torine detection system 3.3.3.8 4.3.3.8 Fire protection instrumenta-tion 3.3.3.9 4.3.3.9 Rad. liquid effluent monitors 3.3.3.10 4.3.3.10 Rad. gas effluent monitors Reactor 3.4.1.4 4.4.1.4 Nut" Der of RHR locos "* 3.4.2 4.4.2 Number of safety valves t 3.4.8 4.4.8 Cociant soecific activity 3.4.9.1 4.4.9.1 RCS temperature-pressure limits 3.4.9.2 .4.4.9.2 Pressurizer temperature and delta T limits 3.4.10 4.4.10 Structural integrity Plant 3.7.2 4.7.2 Steam generator minimum Systems temperature-pressure limits 3.7.7 4.7.7 Control room emerg. HVAC 3.7.9 4.7.9 Some snubbers 3.7.10 4.7.10 Seated sources 3.7.11 4.7.11 Fire suppression system 3.7.12 4.7.12 Fire barrier penetrations Electric 3.8.1.2 4.8.1.2 AC powe 3.8.2.2 4.8.2.2 AC distribution 3.8.2.4 4.8.2.4 DC distribution 3.8.3.2 4.8.3.2 MOV thermal overload protection Soecial 3.10.5 4.10.5 Rod position indication Test Exception Radioactive 3.11.1 4.11.1 Liquid effluents Effluents 3.11.2 4.11.2 Gaseous effluents 3.11.3 4.11.3 Solfds 3.11.4 4.11.4 Total dose Radiological 3.12.1 4.12.1 Nnitoring pmaram Environneetal 3.12.2 4.12.2 Land use census Moni toring
- 3. 2.3 4.12.3 Interlaboratory comoarison orcoram 41
k 6000 - 90 HEAD-CAPACITY . 5000 80 - 4000 - 70 3000 - 60 g g LEGEND: CERTIFIED CC/St PERFORMANCE
- W 2000 - E 50 5 E o 1000
- BASIS FOR ECCS PERFORMANCE
- 5 40 - h i ANALYSIS e 'O- 30 HPSH 20 - X 10 I I I I l l o 0 100 200 300 400 500 600 700 . U.S. GALLONS PER MINUTE Figure 4-1. NPSil and llead Capacity Curves for Charging Pumps (From Sequoyah FSAR). 1 500 100 90 400 80 TOTAL HEAD ~. - 70 _ e E300 -s 60 _ z ~ Ifi ' LEGEND: e , S 5 50 _ y z CERTIFIED RHR PERFORHANCE a E \ O O 200
- 40 -
$
- BASIS FOR ECCS PERFORMANCE ANALYS13 30 - .
100 - 20 - HP3tl
- 10 -
0- ! ! ! l l l 0 0 1000 2000 3000 11000' 5000 6000 7000 j GALLONS PER MINUTE , , Figure 4-2. NPSil and llead Capacity Curves for RilR Pumps (from Sequoyah FSAR). 4000. 100 3500 - 90 _ N. e HEAD-CAPACITY l:- 3000 - 80 - ~
- N .
2500 70 - r 2000 r 60 - LEGEND: Y u-U m CERTIFIED HHSI PERFORMANCE
- 5 1500 - 5 50 ,
o z + BASIS FOR ECCS PERFORMANCE , d 2 - ANALYSIS 1000 -
- 40 500 30 NPSH 0 20
' 10 l l l l l l 0 i 0 100 200 300 400 500 600 700 i ~ it.S. GALLONS PER HINUTE AT 120-3000F Figure 4-3. NPSif and llead Capacity Curves for Safety Injection Pumps (from Sequoyah FSAR). 4 SAravvvAtwa$ b 'b aL 348t # o rnassunsaan . matsarsama ,y, g' Posse R-OPE R ATO O delit p W A&WE5 AteD ,7 g agjggg ggg y , SLOCN WALVt3 g ymissunstEn f SPAAV fitOM CVLS o SG 3 , m SG 3 C Mn 41 n ...P..A. LOOP 3 / k l E YArea /ak LOOP 3 ,,,,,,,,- 4
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..o .c.o..n =>. g i -c< 1 .sm. euur. -c< ,o c ,o .. oo. r. l N-N---l,ii 8 ,o co.. ., , cu ' c 1 . i 1 . i -. co, o.. , s. , ..Ca.C.U i -c< ' ... .>.n o, ,' '"a .. i -N .= ,, N--lEi,, e "Q g =- ! ,,,,_ 1 ' " "' W <l - .. y o)I("-'," i .N.. a- ....c . . . .o .. 2 ,. i . .p , D;, ,r c o.ono <- I * .-<- .. cm.o.. .. . g.;.,e,g, g g . .o. . .: . ....o.. .CCU . . t o. 2 . o,J b .- i . , , ... , a] [ o] l - s J . o, . a .n - -N " "' W'lOI ="""" 1 " " - 1g,' 1 D N-c< & ' ..e ~' 2. "-"' l ..N. ""' N- '0!.-o . . , .. . .C..... . i n. .... , == +^ ! ....m... .CCUMU ID 3 Figure 4-7. Sequoyah High Pressure Safety Injection System, f a 700: FROM NITROGEN L J jh SYSTEM r7 m COLD LEG ACCUMULATOR 2(H - FROM TO REACTOR COOLANT k J kJ k 2 2 77 77 ACCUMU T DRAIN TANK 77 g( gE ER FROM SI COLD LEG INJECTION HEADER FROM RHR d' M [00[I(n OLD LEG LOOP A (B) I OP (4) COLD LEG FROM St FROM COLD LEG COLD LEG ACCUMULATOR 3 (4) INJECTION HEADER Figure 4-8. Typical Cold Leg Accumulator. 49 a / Z Z mR OH T D A E CL AE Z ESS E Z fR V [ T N E d
- M [
EI g ( - I! 'l '! l I s;o N( ! I I ' N kI 1 I .I ! I i l i ;I l m e t - , n , s , I er S y n o I ., o - g i u .r i t c o j e a n - 6 ,:: V ?<E r, I d l a le . ?< 3, ,, r e p _ p S n o E U T "c . E 9 - U E - r 4 c c a g e - r y u g g i g i F n o r a ' i %u m s c c a V r< -M E T S =S Y mo _ ALYt#4AffPtEOfAPROM ALTER 9Aff PEEDER ALYtmeATE PttOgn PROM ALTammATEFitOER 4900 v AC SOARO 10' 880M UNsY I 8000 v AC BOA AO 1&* PnOM Unp? 2 9300 v AC BOARO 2 A-A 0000 v AC BOAAO 294 OstSEL 01534 L GENERATOR GENaRATOR WORMAL PEEDER PROh8 1A*A h0AMAL PEEDE A PROkt Ib5 0000 VAC DOARO IC* se00 VAC BOAnD IS' ,)= )= :)= )= ]= ]= ) =0 ] ~O , , , , , $*eUTDop*4 88eUTOOW85 90AADVA-A BOARO16-8 >= )= 1)= S mP> ")= >= ")= 1 mP> N00 VAC ESP 1A LOAOS 8000 vAC 1B LOAOS . Res. ,U . R,e. PU P , P,A...A. . - 1A-. . C =.C,.A .. -S. . . C.E=. . PU C,eA. 1 . PU P,8-4 .. CS.C.U, E P P .A-.PU A-A SA-A.0-. . CS ERCWPU- ,.-4PSL-8. -. . . A . . .E AUX. PERO,,guA.TER EUR 5 .EAfm -A PUhEP 1A.A . P.UI PE.E,O.W. E .U ATE R PURAP 18-8 t .E ATERS -r 1A-1A 1A-A 1A2.A 101-8 10-8 192-8 0000/400 VAC Tm ahSPOResE RS LA.WL.A=W W ("Y"' % 8"V"' " &% M % NO ' ) ) esc 40 ) ) NC 880 VAC SMUTOOust 400 WAC SMUTOpuuse 00AA01 A2-A SOARO182-8 g mp3 ) mp3 ,) 4AD WAC ESP 1 A2-A LOAOS** e58 WAC ESP 1N-4 LOAOS** . CCW PunsP C-8 (ALT. PE EDER) . CCW PUh0P C-S INOmanAL PIEDERI . asOv eOARO 1A2-A . RAOWBOARO132-8 . vtTAL BATTERV CHARGERS II. lW . VITAL BATTI AV CHAAGElt$ le,IV IALT. PitOERI = ) )O e)') O ese WAC $MUTOOsuse 880 VAC SMUTOOpuse SOAAD141-4 DOAAD 101-8 m., m., ) 400 wAC ESP 1 Al-A LOADS ** ese VAC ESP 181-8 LOAOS** . CCW Puesp 1 A-A . CCW PUMP 18-8 . MOV SOAA01Al-A . RECIPROCATYNG CMARG4ase PUMP . WITAL SATTERY CHAAGERS I.ges . asOW SOAA0181-4 . WITAL S4TTERY CHARGERS L 118 IALT. P E EDE RI IsOffSa
- SWO VAC 30ARO 14. it.1C Asse 10 CAN ASCIfvt POWE# PROIS T4G Usse? 1 tsalm TUnenst GEseERATOR On ONE OF TUUO STAATUP TRA8s5POnnsERS
- OfeLY SELECTIO est VAC LOAOG ARE Lafff0 9 ASSREV1ATIOfuS USED IN TMf5 FIGURE ARE AS POLLOWS:
E SP e Bes GipsEEREO SAFETT PEATURE Room e ag.OUAL DeEAT RGAAOVAL
- COssTAsmaseNT MAY E ftCW e SWENTIAL RAN COOLHeG Watta CCW
- COasPOneNT COOLueG WATER hsOV e neotOR-OPERATEO v&LVE 15 0
- seOsenaALLY OPfee 8sC
- 8eOmanALLT CLOMO Figure 4-10. Sequoyah Class 1E AC Electric Power System.
- 51 -e - - - -e. - p w v.-y V i V .V V. . i E V. .VV, , !!!! i!! i!! i!!il;!!:g::a l.:ai!! als !!!i!!il mis esa itesa i: gigg h :h h kh hh b = h n i 5 I # s1 5
- i_I:ini
!' !i =l1sul! lp!it!i :2;.! . . iI 5
- 15 i
sgi.l - t e.ie a q ser epi q .. g i t t t _. M""'4 ,, s. "o !6t >64 T 'd '" o dx dy dv " I. i. , } 3 . .. ! ! : -J t . ,j i , ,: e g j j i' ', - ]
- ,,- =! 'v' !l,j;'l g g-I,a 2 3
'j' ' "" - o X{ { '[ X { j t -T #f ([ #f T S .7 .. ; T 6 T6 - I L l5 2 * ,i . ?
- i s e n
=l"gjj in as , % -c< 1 ;9; a, l, ,, , , >0-i I 8 .ii.? . as i =.i. ling.: g!gl.iis! 52 4 .S.I.o.,_F. sG M -- p- .C _w r- _ , , n - - . . . . l l c [=;= .c. 1 -N-D<: .m C - --- ,g ._ _
g _ , g-
-. 7_.
4<
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Q u.s, a P'8 E.CW l1.6 m.
- 8 0- e 3 m.
, ___ b I
0 g
l LM g g *
'g
- M1- M>G-- O -!____,'C<}
'b O [$$=.l . $'.
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- a
- Figure 4-12. Sequoyah Essential Raw Cooling Water System.
..)
COMPONENT 8 RHR HEAT EXCHANGER l COOLING WATER HEAT 8 CENT. CHARGING PUMP I
EXCHANGER e SAFETV INJECTION PUMP 8 CONT. SPRAY PUMP CCW PUMP
^
, O N FROM l _ ' TO DISCHARGE
- m ULTIMATE ERCW CANAL A HEAT PUMPS SINK D N V ,
- P NOTE
~ ~
IT IS ASSUMED THAT ELECTRIC POWER, COMPONENT COOLING WATER (CCW)
AND ESSENTIAL R AW COOLING WATE R (ERCW) SUPPORT SYSTEMS ARE ALIGNED SUCH THAT A-TRAIN SAFETY-RELATED DIESEL PUMPS ANO THE RHR HEAT EXCHANGER HEAT ARE SERVED BY THE A-TRAINS OF CCW EXCHANGER AND ERCW, AND POWERED FROM ELEC-TRICAL DIVISION A. A SIMILAR ASSUMPTION IS MADE FOR 8-TRAIN EQUIPMENT.
Figure 4-13. Simplified Model of Sequoyah CCW and ERCW Cooling Loop.
- 5. EVENT TREES FOR RESPONSE TO MODE 5 LOCA's
, Event trees were developed to describe the Sequoyah plant response to postulated LOCAs. The following four LOCA event trees are presented in this section:
e large LOCA L2 (2 RHR pumps initially operating) e Large LOCA L1 (1 RHR pump initially operating) e Medium LOCA e Small LOCA In these event trees, a successful event sequence (e.g., no core melt) is one in which on adequate long-term core cooling capability can be
, established following a successful short-term response. Long-term core cooling can be provided either by establishing an RHR recirculation flow path between the containment active sump and the RCS, or by providing continuous makeup to the RCS at a rate that at least equals the RCS coolant boiloff rate.
In the event trees, multi-mode systems such as the RHR system and the CVCS are modeled as multiple events including: (a) an event to represent the portion of the system that is common to all operating modes, and (b) additional events as necessary to represent the components associated with each unique operating mode of the system.
Within each event tree, individual events are arranged under the functional groupings listed in Table 5-1. Containment response to the LOCA is not included in the event trees.
5.1 EVENT-TREE GENERAL ASSUMPTIONS The following general assumptions were made in the development of the cold shutdown LOCA event trees.
e An operator may commit an error which initiates a LOCA. During i
55
~ "
L--
.L:-.
'~
' 7 .^ " ' ' L '. ~ .L
response to a given LOCA, operator errors of omission are considered (e.g., operator fails to take a necessary action) as well as some operator errors of commission (e.g., operator creates a diversion path in a response system).
e Information required for the operator to make timely decisions regarding LOCA response is available from control room instrumenta-tion. In NRC Inspection and Enforcement Report No. 50-327/81-07 (Ref. 6) it was indicated that the Sequoyah plant operators reacted rapidly to the basic symptoms of a Mode 5 LOCA (e.g., decreasing or zero RCS pressure and pressurizer level). It was not necessary for a diagnosis of the cause of the LOCA to be made before responding
. appropriately to the symptoms of the LOCA.
e If the LOCA is initiated by operator error (e.g., valve misalign-ment), no subsequent operator action is taken to terminate the LOCA .
(e.g., by restoring proper valve lineup). In the NRC I&E report cited above (Ref. 6), it was indicated that the plant operators did ,
not diagnose the cause of the, loss of reactor coolant and the incor-rect valve lineup was corrected only after "the auxiliary operator returned to ,the contro) room and reported that he had opened the RHR spray valve." Therefore, no credit is taken for operator diagnosis and termination of ti.e cause of a LOCA.
e Only water that is adequately borated will be used for reactor coolant makeup. Boron dilution events caused by makeup with inade-quately borated or unborated water were not considered.
e Only one centrifugal charging, RHR or HPSI pump is necessary for adequate coolant makeup following a LOCA.
e Operator is assumed to utilize injection systems in the following l
order: charging pump, RHR pump, HPSI pump.
l e The refueling water storage tank (RWST) is a finite water supply with an available inventory of about 350,000 gallons upon entering 56
Mods 5. Replenishment of the RWST from alternate water sources is not considered.
e Charging, RHR and safety injection pumps can supply water from the RWST to the RCS at their respective runout flow rates following a large LOCA.
e The primary water storage tank (PWST) and the boric acid storage tank (BAST) together constitute an essentially ' infinite water source assuming alternate water sources can be aligned to replenish the PWST, and batches of boric acid can be produced manually to replenish the BAST.
e Nominal continuous makeup rate of properly borated water to the CVCS volume control tank (VCT) from the PWST and BAST is 150 gpm.
e No credit is taken for the limited amount of water in the cold leg and UHI accumulators.
e The LOCA diverts flow from one of the coolant injection paths available to the system being used for coolant makeup..
e If the RHR suction safety valve and isolation valves fail to operate and the RHR system is pressurized above its design pressure, system will fail.
e If the reactor coolant system is pressurized significantly above the RCS minimum pressure and temperature curve limits,an unmitigatible reactor vessel failure may occur.
e Containment recirculation sump failure is not considered as a potential contributor to failure of the recirculation mode of the RHR system.
57 e
5.2 LARGE LOCA L2 EVENT TREE i The large LOCA L2 event tree is shown in Figure 5-1. Two RHR pumps are assumed to be operating prior to the LOCA. RHR pump suction will be i rapidly lost during the LOCA, causing these pumps to cavitate and become airbound. -It is assumed that an airbound RHR pump will eventually fail if not i secured by the operator. Additional information on pump failure when airbound is presented in Reference 7.
A representative time line for plant response following a large LOCA initiated 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after shutdown is shown in Figure 5-2. If no operator
- action is taken, reactor coolant will boil down to the core mid-plane in I approximately 52 minutes (see Section 3.3). Coolant injection systems can provide makeup from the RWST for a variety of times, based on the runout flow
~
rate of the respective pump. An RHR pump can empty the RWST in about one hour. An SI pump and a centrifugal charging pump take about 9 and 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, respectively, to empty the RWST. When the RWST is empty, the RHR system must be aligned for containment sump recirculation if core melt is to be prevented.
The normal CVCS makeup source (e.g., the PWST and BAST) is also available, but at a maximum continuous makeup rate of 150 gpm. If it is
> assumed that the charging pumps are aligned to the ECCS injection path via the boron injection tank (see Section 4.1), one of four coolant injection paths
! may be affected by the LOCA, resulting in only 113 gpm reaching the reactor
, vessel. This makeup rate can match the coolant boil-off rate at approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> af ter reactor shutdown. If the RWST can serve as a coolant makeup
' source until 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> after reactor shutdown, the following two protection options are avail'able to the operator: (a) align the RHR system for contain-
- ment sump recirculation, or (b) provide long-term, properly borated makeup to the RCS from the PWST and BAST. A time line illustrating the time constraints
, associated with successful LOCA mitigation using the normal CVCS water source
- is shown in Figure 5-3.
- Use of the normal CVCS return path was not considered because it represents a less effective makeup capability than the ECCS injection path.
At best, the operator could be assured that coolant from only one of two makeup paths was reaching the reactor vessel following a LOCA (e.g., the LOCA directly affects one injection path). In this case, the actual coolant makeup rate would be 75 gpm.
i-58
,-*-+,e+.- ,,- -,&,.,+e..,_ . , _ . ,,--,,.,_._y', , . , ."
, _ , . - - , , . , - , , , , . , ,h , - --,,ay,.m,p_%, -
.,,..-_,,,.-,,,-,_,_.e. _ . _ _
Although not modeled in the event tree, a potential role for the cold leg _and UHI accumulators can be seen in Figure 5-3. Each accumulator contains sufficient water volume to reflood the reactor vessel from the core mid-plane to the hot leg nozzles. Two cold leg accumulators could be dumped simultaneously to ensure that water from at least one reached the core. The i UHI accumulator could be depressurized and dumped directly into the reactor vessel. Altogether, these accumulator could provide approximately three hours of core cooling during the period from 20 to 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> after reactor shutdown (assuming the operator knew the optimum time to dump each accumulator). By I
adding thisthree hours of core cooling to the time lines in Figure 5-3, a
- large LOCA occurring three hours earlier could be successfully mitigated using the normal CVCS water source. Alternatively, core melt could be delayed for three hours. In the latter case, core melt would not be prevented, but additional time is available to restore containment isolation and to implement the site emergency plan.
5.3 LARGE LOCA L1 EVENT TREE
! The large LOCA Li event tree is shown in Figure 5-4. One RHR pump is assumed to be operating prior to the LOCA. If the operating RHR pump
- becomes airbound and fails, the idle RHR pump is available and can be aligned l
for coolant injection or containment sump recirculation. Other aspects of this event tree are comparable to the large LOCA L2 event tree described previously.
introduces additional events related to RHR and RCS overpressurization when a high-head pump (e.g., centrifugal charging or HPSI pump) is used for coolant i nj ecti on. The high-head pumps are capable of providing makeup at a rate greater than the LOCA leak rate at approximately 600 psig RCS pressure. Ulti-mately, these pumps could restore RCS coolant inventory and pressurize the RHR and/or the RCS above their respective pressure limits if the operator fails to control pressure and the RHR overpressure protection features fail. RHR pipe rupture or reactor vessel failure may result from this overpressurization.
4 59
-,- ,,----- , .n , - . - - . - , .
5.5 SMALL LOCA EVENT TREE The small LOCA event tree is shown in Figure 5-6. This event tree includes the RHR and RCS overpressure protection events found in the medium LOCA event tree and introduces additional events related to core heat removal.
At 600 psig RCS pressure, the small LOCA does not carry away all core heat.
Without additional core heat removal, the RCS will gradually heat up and pressurize above RHR design pressure until the leak rate increases enough to establish an equilibrium RCS temperature and pressure.
60
/
Table 5-1. Grouping of Events in the Cold Shutdown LOCA Event Trees.
Event Tree Functional Grouping of Events Large LOCA, L1 and L e Equipment Protection (RHR Pumps) 2 e Short-term Coolant Injection e Long-term Core Cooling Medium LOCA, M e Short-term Coolant Injection e RHR Overpressure Protection e RCS Overpressure Protection e Long-tenn Core Cooling Small LOCA, S e Nonnal Makeup e Short-tenn Coolant Injection e RHR Overpressure Protection e Core Heat Removal (also perfonns RCS overpressure protection function) e Long-term Core Cooling I
l 61
_ . _ , _ _ _ , . y ,_,,-, , - - r
s e
t
/
/
r
.e.=:; ..e 4,vgA~,,, I
.or 4 g',..= , ..
~.
m.='
. . tam,,,, ., .=.
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, . , .,, . , . pung,. w m news.
% o S 8 9 0 8 6 m g
I N
g ,
a Leg Oa.e>asmes 3 L20se RE LT
- 4 LDg Os e > m esas a
n' L gDee IAELT i
- Y a p ,
7 Lg ca t > 3s soas S L2EO" "'I LI i
- SE8 **
g ,
'w Lgene Os e > m mes 11 LgCtese ast Lt
, it L Cap Ost. e + m eses 2
13 L E3 LI 2
- to LICD Ca e a Mas i n l 1 L ICDet estLT j is 9 Ca 17 L200 "E LI
= Lgne On I '
se L,see inte i
33 L g age tarkt 29 Lp taE Lt II L4 g Oa.e 7 30 mas 33 Lg an amLY
, 38 LgAC Oa.e > 3B eems a
5 LgACM asELT
, N L gACF Os.e s m osas 27 k gACFee ap0 L7 35 kgAS 48E LT Figure 5-1. Large LOCA L Event Tree.
2 62
/
/
/
- ~~wgy e y.r m m. r - , -r ig . .. . .. . . . p .
,,,5 ,. .
a i OPERATOR LOCA RESPONSE ACTIONS
.. 1. NO INITIAL OPERATOR RESPONSE ONSET OF
-! 2. OPER ATOR ALIGNS CHARGING PUMP CORE MELT TO INACT FROM RWST (~550 spel
- 3. OPERATOR ALIGNS RHR PUMP TO
INKCT FROM RWST F5600 spm) l
- 4. OPERATOR RESTORES AND ALIGNS Si PUMP TO INECT FROM RWST { ~ 650 spel l'
'5 5. OPERATOR ALIGNS RHR FOR LOW l PRESSURE RECIRCULATION
- 8. OPERATOR FAILSTO ESTABLISH 5 RECIRCULATSON FLOW PATH WilEN g 2 OK RWST IS EMPTIED 2l !
ONSET OF l CORE MELT l.
l
- ' ' = a LARGE
' ' ' ' , LOCA
"~
REACTOR TRANSITION l 1,~
SHUTDOWN TO RHR EARLIEST l INITIATED COOLING TIME IN 5 MODE 5 l - OK cn 3b l ONSET OF CORE MELT I
1 I
l : oK 4 ',
ONSET OF .
CORE MELT a a e R e a R R R R R R R R R R R R { g 3 l l l l 3 g l l l 3 l l l l g 0 5 to 15 20 25 30 35 TIME AFTER REACTOR SHUTDOWN (HOURSI Figure 5-2. Time Line for Plant Response Following Large LOCA Initiated at 20 llours After Reactor Shutdown.
5 1 OK-LARGE LOCA l WITH INITIAL lI RESPONSE BY . m
- OK CHARGING '
PUMP INJECTING I FROM RWST l O
ONSET OF Q*I CORE MELT
. 5 r OK OPERATOR LOCA RESPONSE ACTIONS LARGE LOCA l
--l WITH INITI AL C 5. OPERATOR ALIGNS RHR FOR LOW RESPONSE BY a lI
' ~ OK PRESSURE RECIRCULATION RHR PUMP '
- 6. OPERATOR FAILS TO ESTABLISH I TING l FRM RWST g RECIRCULATION FLOW PATH WHEN RWST IS EMPTIED ONSET OF CORE MELT
- 7. OPER ATOR AllGNS CVCS TO PROVIDE EXTENDED MAKEUP FROM ALTER-NATE WATER SOURCE (~ 150 gpm) .
5 A 8. OPERATOR FAILS TO PROVIDE OK 3
MAKEUP FROM AN ALTERNATE WATER SOURCE AFTER RWST LARGE LOCA l SS EMPTIED WITH INITIAL g7 RESPONSE BY , ,
- OK St PUIW , ,
INKCTING l FROM RWST 0*8 ONSET OF CORE MELT .
1 a I E I e a R I e I a i e i i e a i I i n 20 25 30 35 - 40 TIME AFTER REACTOR SHUTDOWN (HOURS)
Figure 5-3. Time Line Illustrating Time Constraints Associated With Successful LOCA Mitigation With a 150 gpm Alternate Water Source.
- f
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8- - ..
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1 L, Os it op 3 Roses a L,0 ca. > 3. mas l ,
3 L,GM RE LT
. t,o on.. > = een
' asELT 5 L,Du e L,c onitos amas I L,CG 04.4 D N uRS l ,
' tee LT O L, CGM 9 L,CE DE 1109 3 #Mae 10 L,CEG Os_ s >m eses
[ ,
' 11 t*E LT L,CEGN is 6,ca s on. , a nas i
v3 L,Caen asEtt
- 1. L,CD OE e > 3se#9 15 L,CDet esE LT 16 L8 3 Os 1107 3 mmet
' 17 L,80 esELT
, se L,st os sior a name l
' 19 L,SEG asELT
- I at L,ett asELT 21 L30o asELT 22 L,A OE 11OF t meteH i , as 4,40 os. e > p was a
de.
M L, AGM asELT j ,
B L,40 CE.t > N eset
' asELt 3B L, AOM g
rr 4,AC os is osi sans
- a 4,Ac0 om e > as ==s N L,ACGM asELT 3s L,AC on it or i nuni m 6,Aca0 ca. e > m nas II L,ACEGM AsE LT 33 L, ACE 8 Da e ,- a sosis i
M L, ACIF4 WE LF E L,ACD OE t .,33 eens 1
N L,ACDee tee LT IF L 9AS Os at 081 #sen.
' M L, Ate asELT
, . %A.E .. s,se.,
I '
ao 6,AsEG isEtt i .t 4,Aat, .sEL,
.: 6,Aso esEtt Figure 5-4. Large LOCA L) Event Tree.
65
i l
l ev esa.coe6ame.=merie= es.s eve ===eauss e.ovecm @p'*]e @ l'."
"=* = ==. ,= .=, =,. r:::, .=, 2_~ ~==r~5~ jL=6 ~- ~- -~.- . . .
= == -. == = - =: ,: . .=,
M , , . , , . , , . . i . ..
1 M M l , 2 MG OK t 24 Mas 3 esGM MELT 4 tot OK I
l , 5 amG OK. t % 24 MAS 8 MeGM RSELT
, 7 teu OK l , e asWG OK.t > 24 Mas 9 MuGM MELT 10 MuN MELT 11 asuK MELT 12 MO OK. *
- 24 MA5 13 MOM MELT
, 14 esO4 OK,I e 24 MAS 15 asOIM MELT
, 18 MOW OK.t>24 MAS I 17 asouM MELT I
, N M M j , 21 tsCO OK. t > 34 MAS 22 RACGM MELT y 23 asCE OK l , M esCEG OK, e > N MR$
25 atCIGH WELT 30 nsCEI OK l ,
27 asCElG OK.t>29MR$
28 asCEIGH MELT 29 heCEW OK 30 MCEuG OK, t > 23 Mag 31 MCEUGM asELT 32 heCluN MELT 33 RACEuK MELT 34 tsCEF OK. t > 33 Mas 35 esCEFM teELT
, 30 MCO OK. >35 MAS 37 nebDM MELT
, 28 MS OK 39 MSG MELT
, 40 assE OK
' 41 utEG MELT
, 42 asete OK 43 AsSE14 MELT 44 heefu OK I de testuG asE LT 48 heetues MELT
. of RsDE uK MELT 48 McEF MELT 49 As0C WELT Figure 5-5. Medium LOCA Event Tree.
66
.om.m .seeve eso.v tse em -e m.cte. enneweassessue. ,seven coes .e.v a .owa l . ecee
=== - evc. , ~ ~
ea ' gw agv
- ag I
- c.e see
.... . c .c.e. o. .. .uso.s. _
-.v.e c.s. m.
.o t e a C S E 8 8 8 0 4 F 4 G 1 S OK
, 2 SO OK.es3SMaS l , 3 SOr OK.i 3Swas 4 SOJw MELT
, T SM4 OK S SMiG MELT 9 SM80 OK 10 SMcQ MELT
, 11 SHu OR I '
12 SMUG MELT I
13 SMUN MELT to SMuK WELT 15 SHO MELT
, is SMC OK 17 SMCG MELT IS SMCr OK I 19 SMCJ13 MELT 28 SMCJW MELT 21 SMCE OK 22 SMCEG MELT
, 23 SMCEI OK 24 SMCItG MELT
. 28 SMCEIO OK 38 SMCEIOG MELT 27 SNCEWPr OK I 28 SMCEupro MELT
' 29 SMCEUPJW MELT 30 SMCEuu MELT 31 SMCEF MELT 32 SMCO ME(T 33 SS OK 34 SSG MELT
, 38 SBT OK I '
38 SSEG MELT l
37 SSJW MELT 38 SSE OK 39 SSEG MELT 40 SSE1 OK 41 SSElG MELT 42 SSEIO OK 43 SSEIOG MELT 44 SStu OK I 48 SatuG MELT 48 SSEuN MELT di SSEUK MELT 48 SSEF MELT 48 SSO MELT Figure 5-6. Small LOCA Event Tree.
67 u n ,., ; .n .-r-- -
,;u- ., --. - - - + -
F:
I -
h e
D 5
I i
!~
i i
I f
68 lj I
v - e wv-=- -
- y 7I
- 6. FAULT TREES FOR RESPONSE TO MODE 5 LOCAs Simplified fault trees were used to define the equipment failures and operator errors associated with each event in the event trees described in
~
Section 5. The fault trees for each event are included in Appendix A.
Assumptions and simplifications made in the construction of these fault trees are listed below:
e Manual valves which need not change position during LOCA response are not modeled.
e Support systems (e.g., electric power CCW and ERCW) are not included in the fault trees. Support system failure probabilities are esti-mated separately and these results are weighed in the computation of event sequence probabilities (see 'Section 7).
e The normal CVCS coolant return path to the RCS is not modeled.
e The RHR and 'HPSI hot leg injection paths are not modeled.
e Normally open valves which must fail closed are not modeled.
e Check valve failure to open 'is modeled any time flow has been interrupted in a pipe.
e Random pipe or tank rupture is not considered in the fault trees.
e Effects of ventilation system failure on long-term operation of
. safety-related equipment is not considered.
+
69 2
q
--g. a - , _ , . . _ _ - _ , , - , . . , - - , , , , ,.--e
4 9
70 4 ,**M y, e , ,, 40 4*%i- h "$
- 7. QUANTIFICATION OF EVENT TREES 7.1 DATA FOR QUANTIFICATION The data used in quantifying the fault and event trees is listed in Tabl e 7-1. The component, offsite power and operator error data were ab-stracted from WASH-1400 (Ref. 2). The safe shutdown earthquake (SSE) data was derived from a BNL-NUREG informal report (Ref. 7). Cooling water and electri-cal system reliability was estimateti using WASH-1400 component data and a simplified model of these Sequoyah support systems.
7.2 ASSUMPTIONS REGARDING QUANTIFICATION The following assumptions were made in the quantification of the fault and event trees: -
e Plant conditions in cold shutdown do not significantly affect equipment failure probabilities, e In the event of a safe shutdown earthquake (SSE), a pipe break occurs with a probability of 1.0.
e A normally-closed, motor-operated valve (MOV) outside containment can be treated as a manual valve because of the time available for manual actions and the general availability of manual operating features on MOVs, e Operator practice at Sequoyah of manually seating safety-related ,
MOVs following valve closure does not affect valve failure-to-open probability, e Sequoyah CCW and ERCW system failure probability can be estimated based on the simplified system model described in Section 4.2.
71
.- # c,.. ..,,.,,.,4 .- . , ~ . . - . . . . . - . - .,,,y. - - _. ._ , . . - , - . . ..~ e-_ - - . , - . ,-m. .-,,e - y
These estimated failure probabilities appear in Table 7-1.
e The failure probability for the onsite Class 1E electric power system is dominated by the battery, diesel generator, and diesel ERCW failure probabilities. Failure probability of a Class 1E electrical division can be estimated based on a simplifed model which includes only the above elements. Estimated failure proba-bilities for a Class 1E AC electrical division are included in Table 7-1.
e Operator errors are ranked into two categories (ommission, commis-sion) using engineering judgement as to their likelihood.
e Fault tree truncation is done during the quantification process to eliminate events which do not apply to the sequence being evaluated.
e Event sequences are weig'hted by power availability probabilities and initiating event probabilities.
e Success probabilities were not included in the event sequence quantification.
7.3 DESCRIPTION
OF CALCULATIONS The calculations associated with determining the probability of core me,1t in cold shutdown began by developing the Boolean algebraic expressions for each of the fault trees shown in Appendix A. These expressions contain all the events that could be involved in the analysis. In order to reflect changes in sequence assumptions, these expressions must be truncated to eliminate those events which are no longer involved due to specific sequence assumptions. For example, the fault tree shown in Figure A-2 for event A in t
the event trees deals with the securing of the RHR pumps following a large LOCA. This tree is either left intact or is truncated based on the initial assumption regarding the number of operating RHR pumps (e.g., two or one).
The external events in the tree are simply flags to indicate this truncation 72 m - ,
process. Truncation also occurs due to assumptions dealing with power availability and the occurance of loss of offsite power.
Support system availability was used as a partitioning function in the solution process. The support system was assumed to be at " full support" (both electrical load groups and cooling water loops available), " half sup-port" (one of two load groups and cooling water. loops available), or "no support" (both electrical load groups and cooling water loops unavailable).
The fault tree expressions were truncated and solved under each of these assumptions. Fault trees containing pumps were also solved for various lengths of time of pump operation. In this way a table of fault tree quanti-tative solutions was developed using the data in Table 7-1 for varying assump-tions dealing with support system availability and event tree sequences (See Table 7-2).
Using the tabulated fault tree results, each event tree sequence was appropriately solved by combining the system event failure probabilities for each failed system event in the sequence. This was done under the assumptions of full and half support system availability. Once all sequences in an event tree were quantized, the values associated with sequences leading to melt were summed for each support system availability assumption. These values were then weighted by the appropriate probability of being in that particular support system availability situation. These weighted probabilities were
! summed along with the probability of no support systems available to give the probability of a melt given the initiating event. The probability of the initiating event is then multiplied by the probability of a melt, given the initiating event, to find the resulting probability of core melt.
This process of quantification was done for a number of cases deal-ing with initiating events, LOCA size, maintenance considerations, and length of loss of offsite power. The computations involved were done by hand and thus were done as simply as possible. Details of the quantification are presented in Section 7.4.
7.4 QUANTIFICATION OF CORE MELT PROBABILITY Using the data in Table 7-1, the fault trees in Appendix A were quantified under varying assumptions dealing with their usage in the event trees. Table 7-2 gives the quantification results for the individual fault 73 ,
/
/
. . . . _ . _ . , . _ . _ - - _ , - - _ . ...-_.,,_.-.c . _ ,. ,__ _ m- . . . _ . _ - . , ~ . , . - . . - - -
_ . ~ . . . . . _ , . . , . , . . ~ _ .
' trees. Note that th se values do not include support system failure probabil-i ti e s. Thus failures of electric power or cooling water are not covered probabilistically in the numerical results in Table 7-2. Event A has been evaluated assuming a common mode failure of the operator rather than independ-ent failures of the operator to secure multiple RHR pumps.
The event trees for large and medium LOCAs were solved for various initiating event assumptions using the intermediate results from Table 7-2.
Tables 7-3 to 7-7 present tabular summar.ies of the individual sequence proba-bilities for the following cases:
e Large LOCA L2 initiated by an SSE (Table 7-3),
e Large LOCA Li initiated by an SSE (Table 7-4),
e Medium LOCA initiated by an SSE (Table 7-5),
e large LOCA L2 initiated by operator error with no loss of offsite power (Table 7-6),
e large LOCA L2 initiated by operator error with a one-hour loss of offsite power (Table 7-7).
The small LOCA event tree was not quantified because there. may be conditions when this LOCA category does not exist (e.g., when decay heat levels are low, see Section 3).
The tables 'give conservative event sequence probabilities since they only include multiples of. event failure probabilities and do not include event success probabilities. Also, note that each sequence that is quantified is labeled with a "P" or an "M". The P stands for "possible melt", depending on the time between reactor shutdown and LOCA initiation. The M stands for
" melt" regardless of time of LOCA initiation. Two values are given for each
- sequence; a " full support" probability and a " half support" probability.
These values indicate the likelihood of an event sequence given either full electric power and component cooling support or only a single load group supplying electric power and component cooling functions. These probabilities do not include support system failure probability but only consider random failures of the system components or operator error included in the fault trees in Appendix A.
The summary values at the bottom of Tables 7-3 to 7-7 indicate the sum of the appropriate individual sequences in the P or M category. The M 74
summary result is simply the sum of the M labeled sequences. The P summary result is the sum of the P labeled sequences along with all M labeled sequenc-es which do not have event H as part of the sequence from a success or failure standpoint.
Table 7-8 shows the results of the quantification process for 20 different cases when support system availability likelihood and initiating event probabilities are included. These values are found by the following method:
4
- initiating event) x Pj(melt) = Pi(fPj(full support) x Pi(meltlfull support) +
Pj(half support) x Pj(meltlhalf support) +
- Pj(nosupport)f where
- Pi is a probability function related to the i th case, and is in units of per reactor-year.
Pj(initiating event) is found in Table 7-1 (e.g., an SSE during cold shutdown or an operator error of commission).
Pj(full support), P1(half support), and Pj(no support) are found in Table 7-9.
Pj(meltlfull support) and Pj(meltlhalf support) is found in one of Table 7-3 to 7-7 depending on the case.
To illustrate how the core melt probabilities in Table 7-8 were computed, consider the following example for Case 1, no maintenance. From Table 7-1 we get an initiating event probability for an SSE of 2.0 x 10-4 From Table 7-3 (L2 LOCA initiated by an SSE) for the "P" case of no time constraint, we get values of 1.32 x 10-2 and 2.27 x 10-2 for full and half support core melt probabilities, respectively. From Table 7-9 we use the "LOSP one hour-no maintenance" case to get values of 0.91926, 0.07714, and 0.00360 for full, half, and no support system availabilty likelihoods. Sub-stituting these values into the previatis equation yields: ,
75 e, ,--- -, ,-. . ----- -- m --- - -
P1 (melt) = (2.0 x 10-4) x f(.91926) x (1.32 x 10-2) +
(.07714)x(2.27x10-2)+(.00360)f
=(2.0x10-4)xf(.01213)+(.00175)+(.00360)f
= (2.0 x 10-4) x (.01748)
= 3.50 x 10-6 Other cases are computed in a similar fashion. Maintenance case computations used the data in the lower half of Table 7-9 and only used the half support values in Tables 7-3 to 7-7. These cases assume an entire support train is unavailable due to maintenance (e.g., a diesel generator or a component cooling water loop is unavailable because of maintenance). Note that the unavailability of front-line systems due to maintenance is not modeled.
A total of 40 core melt probabilities are listed in Table 7-8 (20 cases, each with and without maintenance on support systems). No attempt has been made to probabilistically combine these cases to get an overall probability of core melt.
Some points of interest dealing with this analysis are the following:
e The probability of overpressurization of the RCS or RHR in a medium LOCA is of' low probability (3 x 9 in sequences 10 and 11 from Table 7-5). These sequences do not treat a potential common mode operator error which might increase these sequence probabilities to 3 x 10-4 If a common mode operator error were assumed, four sepa-rate operator actions would be grouped into one event. Such an assumption would increase the medium LOCA probabilities listed in Table 7-5. _
e The SSE induced LOCA initiating event probability is only the proba-bility of an SSE and does not include the likelihood of a resulting pipe rupture.
e The operator error induced LOCA initiating event (e.g., due to system misalignment) assumes there is no operator recovery which restors system alignment.
76 v ,,,, - - - , - , - , , ,
7.5 ANALYSIS OF RESULTS From a review of Table 7-8, the following general observations are made:
e A large LOCA initiated while two RHR pumps are running is the " worst case" LOCA. Lower probabilities of core melt are associated with large LOCAs initiated while one RHR pump is running, and with medium LOCAs.
e Unavailability of one support system train due to maintenance increases core melt probability by as much as an order of magnitude.
e Ability of the CVCS to provide adequate makeup from the normal makeup water source (primary water storage tank plus boration as necessary) is a function of time after shutdown. Core melt proba-bilities are reduced as much as 50 percent when this normal makeup water source is adequate (e.g., greater than 24 to 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> afr shutdown).
e Long-term loss of offsite power has a significant impact on core melt probab'ility. If offsite power is assumed to be unavailable for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, core melt probability is estimated to be more than an order of magnitude higher than the case of a one hour loss of offsite power 7.6 TREATME T OF COM40N MODE FAILURE Common mode failures of equipment or operator actions were addressed in this analysis. However, the amount of effort spent on this area was minimal. Those obvious areas of common mode failure were handled appropriate-ly as they arose. For example, the electric power system and other component support systems were treated as potential sources of common mode failure of entire trains of systems. In addition, the operator was assumed to have not secured both RHR pumps in appropriate event sequences if he failed to secure one pump. These were the only common mode failures reflected in the results.
77
. - . - . _ - . . , _ _ _ _ - . . _ - . - . . - - _ .,-.,._-r..
Diesel generator common mode failure (unable to start either diesel) was not modeled but this would only impact the results dealing with short losses of offsite power. Sequences dealing with lengthy loss of offsite power are dominated by diesel running failure.
The cases dealing with medium LOCAs and concern about overpressuriz-ing the RCS do not treat the operator as making a common mode failure when attempting to control RCS pressure through one of three different mechanisms.
Currently these sequences are of the order 3 x 10-9 in probability. A common mode treatment of operator error in these sequences would increase these sequence probabilities to approximately 3 x 10-4 which would impact the medium LOCA result.
Common mode failure of equipment in the same location or environment was not addressed due to the limitations of the scope of this study.
7.7 UNCERTAINTY IN ANALYSIS No sensitivity analyses or uncertainty analyses were done in this study. Point estimates of component, event, and operator failure probabilities were used with no error band consideration. Conservatism was built into some of the data to reflect uncertainty, however, most of the data used were taken directly from the WASH-1400 data base.
Although no uncertainty analysis was done, the study did suggest those inputs to which the results were most sensitive or which were felt to be very conservative. One area of conservatism arises from the assumption that failure to secure an operating RHR pump during a cold shutdown large LOCA would lead to total failure of the pump with probability 1.0. This is not accurate but an accurate estimate of the probabiliti of failure of an airbound pump was not available. This assumption does impact the results and the probability of core melt would likely decrease if more accurate data were available.
Another conservatism was the assumption that the SSE initiating event leads to a LOCA with probability 1.0. Again, no data dealing with pipe failure rate under cold shutdown SSE conditions were readily available. This probably adds a significant conservatism to the calculated probability of core melt. With accurate data, it is possible that the calculated core melt proba-bility could be reduced by orders of magnitude, depending on the actual like- J lihood of pipe rupture.
78 I
The operator error initiating event assumed that no likelihood I exists for terminating a LOCA created by' system misalignment. If the operator has the capability to diagnose, identify, and correct the cause of this type of LOCA, the probability of core melt in the related sequences would be
- reduced. s Operator error in general is handled crudely and may be conservative or not. It is riot readily known which way the error in this model would drive the results.
Another area of significance in the results is the assumed failure rate for an operating diesel generator. This rate dominates the SSE sequences for 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> loss of offsite power cases. It cven begins to dominate for the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> LOSP cases. Any changes in this rate would impact results for those cases mentioned.
4 9
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a 79 i
- - w9fr' -e+i +- -+y-e wew-a.r e-ey -wew y mewww-,e+m,e--e%e.s wyy---r -wm+ -g+wyte-- -,--o--e,as--tew-P-m. we t-w **N-saa'------m4-wwe-'ewvev---r==w--y*--ve-_wTm-e'y,e- -ny-
r Table 7-1. Failure Rate Data for Components.
Failure Failure Component / System Description
- Type
- Probability Check valve Failure to open D 1.0x10~4 Valve outside Closed for long periods D 2.0x10'4 containment Closing or closed for short time and opening D 1.0x10,4 Valve inside MOV opening 0 1.1x10'3 MOV closing D 1.0x10'3 NV opening D 4.0x10'4 NV closing D 3.0x10'4 Relief valve Fail to open 0 3.0x10-2 Fail to close D 1.0x10-2 Pump Fail to start 0 1.0x10~3 Fail to run 0 3.0x10-5/hr Essential raw 100 hr. operation, pump cooling water & check valve in parallel.
(ERCW) loop serving single ERCW loop - 1.68x10,5 Component cooling 100 hr. operation, pump water (CCW) loop *& check valve serving single CCW loop - 4.10x10 3 Onsite Class 1E Diesel starting & running AC power division for specified time, plus battery & ERCW with 2 check valves, 1 MOV 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> - 3.63x10-2 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> - 6.20x10-2 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> - 2.84x10'I Offsite power 0 2.0x10-5/hr Operator error Comission - 3.0x10-3 Omission - 1.0x10-2 Instrumentation Automatic valve closure instrument channel D 1.0x10-2 SSE during cold 20% time per year in cold shutdown shutdown - 2.0x10,4
- Abbreviations used in this table include the following:
MOV = motor-operated valve NV = pneumatic or hydraulic valve SSE = safe shutdown earthquake D = demand failure 0 = operating failure 80
~ Table 7-2. Fault Tree Quantifications Failure Probability Results.
Hours to Full- Half Event Special Assumptions Failure
- Support Support
- A 1 pump at start of event D 1.10 x 10-2 2 pumps at start - both not secured 0 1.00 x 10-27 2 pumps at start - only one secured D 2.00 x 10 3 8 No LOSP D 5.00 x 10'4 1.05 x 10'3 LOSP D 5.02 x 10-4 1.60 x 10'3 C D 1.01 x 10-2 D No LOSP 0 4.76 x 10 6.93 x 10'3 LOSP D 5.62 x 10 -5 7.50 x 10-3 E D 3.10 x 10'3 F D 1.07 x 10-2 1.18 x 10'#
9 1.07 x 10-2 1.21 x 10-2 G RHR used to inject in sequence D 3.10 x 10'3 3.10 x 10~3 RHR did not inject in sequence 0 3.00 x 10~3 4.00 x 10'3 H CVCS used to inject in sequence 89 5.45 x 10~4 90-91 5.46 x 10'4 99 5.48 x 10'4 CVCS did not inject in sequence 89 2.45 x 10-4
~4 90-91 2.46 x 10 99 2.48 x 10'4 I D 1.00 x 10~4
~J- 0 3.00 x 10'3 K D 1.01 x 10'4 1.10 x 10'3 N D 1.00 x 10-2 1.04 x 10-2 0 0 1.00 x 10'# l
- D indicates a failure to start or demand failure t Blanks in this column indicate no difference between half & full support Y Connon mode failure of operator 81
Table 7-3. Quantification of Large LOCA L2 Event Tree (SSE Initiating Event).
5 EVENT STATUS (2) TIME REQUIRE"ENTS SEQUENCE g FOR SYSTEM OPERATI0d3) PROBASILITY w
S
- E'.
= -
E y - M i E v
w E
3 3 0 -"
E 0 M v 2 2 = R 5 w" U U 0 *
= = = 8 O O E U U Q = N 5 FULL HALF A B C D E F G H O I 3 E 8 $UPPORT SUPPORT 1 + + + + + 11 89 2 + + + + - + 100 89 P 3.00 x 10'3 4.00 x 10'3 3 + + + + 100 1.64 x 10-6 -6 11 M 2.18 x 10 4 + + + - + 100 0 89 P 7.05 x 10-5 8.42 x 10'3
+ + + -6 5 - - 100 D J.1 M 3.84 x 10-8 4.59 x 10 6 + + - + + + 100 7 + + - + + - + 99 1 99 P 3.13 x 10-5 -5 3.13 x 10 8 + + - + + - - 99 1 M
.?.1 9 + + - + - + + 91 9 10 + + + + + 9.39 x 10-8
- - - 91 9 91 P 1.25 x 10' 11 + + - + - + - - 91 9 E .M 2.31 x 10'II 3.08 x 10"II 12 + + - + - - + 100 D 100 P 3.35 x 10'I 3.69 x 10'I 13 + + - +' - - - 100 0 M 190 14 + + - - + 100 0 100 P
-5 7.12 x 10~7 8.50 x 10 15 + + - - - 100 D 100 0 M 16 + - + + + 0 100
+ + + -6 17 - - 0 1 M 1.56 x 10-6 4.96 x 10 18 + - + - + + D 91 9 19 + - + - + - 0 9 M 4.67 x 10 1.98 x 10-8 20 + - + - - D ,9,, M 1.67 x 10-8 6.00 x 10~
21 + - - 0 D M 3.54 x 10-8 1.35 x 10
- I
+ + -2 22 - + 100 89 P 1.00 x 10~2 1.00 x 10 23 - + + - 100 g M 5.45 x 10 6 5.45 x 10
24 - + - + + 100 9 91 P 1.01 x 10*# 1.01 x 10
-8 25 - + - + - 100 9 91, M 2.48 x 10-8 2.48 x 10 26 - + - - + 100 D 100 P 1.08 x 10~8 1.19 x 10-6 27 - + - - - 100 D MJ IQQ
-5 28 - - 0 M 5.04 x 10-6 1.60 x 10 I M 4.72 x 10' 1.65 a 10
P 1,32 x 10 2 2.27 x 10' l
82
Table 7-3. Quantification of Large LOCA L2 Event Tree (SSE Initiating Event) (Continued).
Notes: (1) Event sequence nusbers correspond to the sequence numbers in the event tree.
(2) Event status is indicated as "+" (system success), " " (systen failure),
or blank (system does not appear in the particular sequence).
(3) Listed time requirements, in hours; were used in computing equipment failure probability over a 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> interval following LOCA. Underlined times indicate system failure within the specified interval. The code "0" indicates a demand failure of the system in question.
(4) Consequences are listed as "M" (core melt), or as "P" (core melt, subject to specified time constraints).
(5) Probabilities are per year. " Full Support" probabilities are values assuming no failure occurs in support systems (electric power, component cooling water and essential raw cooling water). " Half Support" probabilities are values which include the probability of failure of one electrical division and/or one cooling water train.
4 I
i 1
l l
83 )
i
Table 7-4. Quantification of Large LOCA Lj Event Tree (SSE Initiating Event)*.
b EVENT STATUS (2) TIME REQUIREMENTS SEQUENCE gg) g FOR SYSTEM OPERATION (3) PROBA8ILITY R
3 g - U
- 0 h "
c h h R
- e e - - . e =
s g E D 0 E E 3 8 0 e E E M m sur a A B C D E F G H O - N 3 b SUPPORT SUPPCRT 1 + + + + + 11 89 2 + + + + - + 100 89 P 3.00 x 10'3 4.00 x 10'
+ + + + -6 3 - - 100 8,,9 M 1.64 x 10'0 2.18 x 10 4 + + + - + 100 0 89 P 5.62 x 10 5 7.50 x 10' S + + + - - 100 0 _8,,9, M 3.06 x 10-8 4.09 x 10'0 6 + + -
+ + + 100 7 + + - + + - + 93 1 99 P 3.13 x 10'I 3.13 x 10-5
+ + + +
8 - - - 99 1 3 M 9 + + - + - + + 91 9 10 + + - + - + - + 91 9 91 P 9.39 x 10 8 1.25 x 10*I 11 + + - + - + - - 91 9 H M 2.31 x 10'II 3.08 x 10'II 12 + + - + - - + 100 0 100 P
,7 ,7 13 + + - + - - - 100 0 0 J,00 MJ 14 + + - - + 100 0 100 P 5.68 x 10*I 7.58 x 10' 15 + + - - - 100 0 J,3 M 16 + - + + + 0 100
+ + + -6 17 -- - 0 1 M 1.56 x 10-6 4.96 x 10 18 + - + - + + 0 91 9 19 + - + - + - 0 9 M 4.67 x 10'I 1.98 x 10' 20 + - + - - 0 _9,, M 1.67 x 10'0 6.00 x 10'
+ -5 21 - - 0 0 M 2.82 x 10 8 1.20 x 10 22 - + + + + 11 89 23 - + + + - + 100 89 P 3.30 x 10-5 4.40 x 10
+ + + -8 24 - - - 100 ,8, 99, M 1.80 x 10 8 2.40 x 10 25 - + + - + 100 0 89 P 8.25 x 10 5 8.25 x 10-5 26 - + + - - 100 0 g M 4.50 x 10-8 4.50 x 10-8 27, - + - + + + 100 28 - + - + + - + 99 1 99 P
,7
,7 29 - + - + + - - 99 1 9,,9 M) 30 - + - + - + + 91 9 I
84 J
[ '
Table 7-4. Quantification of Large LOCA L Event Tree (SSEInitatingEvent)(Continudd).
b EVENT STATUS I TM REW!RDmTS SEW BCE g FOR SYSTEM OPERATION (3) PROBABILITY ($)
R
=
5 - -
E j g 5 - M E x e a g g a g l- g -
- = 0 0 3 $ $ g a e e E I a I e . . ,
$ U R E g FULL HALF s A 8 C D E F G H e w U u SUPPORT SUFPORT 31 - + - + - + - + 91 9 91 P 1.03 x 10 1.38 x 10
32 - + - + - + - - 91 9 E M 2.54 x 10*I3 3.39 x 10'I3 33 - + - + - - + 100 D 100 P 3.69 x 10 4.06 x 10
34 - + - + - - - 100 0 g M 35 - + - - + 100 0 100 P 8.33 x 10*I 8.33 x 10'I 36 - + . - - 100 D 0 1,00, M 37 - - + + + 0 100 38 - - + + - 0 1 M 1.71 x 10'8 5.46 x 10-8 39 - - + - + + D 91 9 40 - - + - + - D 9 M 5.14 x 10*II 2.18 x 10-10
+ 6.60 x 1010 41 - - - - D 1 M 1.83 x 10-10 42 - - - D D M 4.14 x 10-8 1.32 x 10*I M 3.68 x 10'I 1,43 x 10
-2 P 3.21 x 10'3 1.18 x 10
'See Notes, Table 7-3.
I 1
l 85
Table 7 Quantification of Medium LOCA Event Tree (SSE Initiating Event)*-
EVENT STATUS (2) mE EWIREMENTS 3) SEWENCE III FOR SYSTEM OPERATIO PR08 ABILITY
- 5 =
0 . 8 2 ts g .g 5 . v . 7
~
5 g G h .-
"E 8 g :5 E w 3 c R S j G = N O E R E a . .
v a a - -
a = . g
- W g 5 5 -
" 5 5 g g y =. ,
- 5 v v = 8
- 8 8 8 * "
O . g FULL HALF UO 8 C D E F I J K N G H O E G u SUPPORT SUPPCRT 1 + + + + + 11 89 7 + + + + - + 100 89 P 3.00 x 10'3 4.00 x 10'3 3 '+ + + + - - 100 ,Q9 M 1.64 x 10 2.18 x 10-6 4 + + + . + + 11 89 5 + + + - + - + 100 89 P 3.00 x 10*I 4.00 x 10*I 6 + + + . + - - 100 89, M 1.64 x 10*II 2.18 x 10-10 7 + + + - - + + + 11 89' 8 + + + - . + + - + 100 89 P 9.00 x 10-10 1.20 x 10**
9 + + + - - + + - - 100 E M 4.92 x 10*I3 6.54 x 10'I3 10 + + + - - + - 11 M 3.00 x 10'I 3.12 x 10*I 11 + + + - - - 11 M 3.03 x 10*II 3.30 x 10-10 12 + + . + + 100 0 89 P 5.62 x 10 5 7.50 x 10'3 13 + + - + . 100 D E M 3.06 x 10-8 4.09 x 10-6 14 + + . - + + 100 0 89 P 5.62 x 10 7.50 x 10*I 15 + + - - + - 100 0 ,!9 M 3.06 x 10-12 4.09 x 10-10 16 + + - - - + + + 100 D 89 P 1.68 x 10 12 2.25 x 10*I 17 + + - - - + + - 100 0 9 8_9, M 9.18 x 10-15 1.23 x 10 12 18 + + - . - + - 11 D M 1.69 x 10'I3 2.34 x 10*II 19 + + - . . . 11 D M 1.70 x 10-15 2.48 x 10-12 20 * - + + + 100 21 + - + + . + 99 1 M P
{ 3.13 x 10 5 3.13 x 10 5 22 + . + + - - '99 1 3 Ml 23 + . + - + + + 91 9 24 + - + - + + - + 91 9 91 P 9.39 x 10-8 1.25 x 10*I 25 + - + . + + - - 91 9 E M 2.31 x 10*II 3.08 x 10*II 26 + . + . + - + + 91 9 27 + - + - + - + - + 91 9 91 P 9.39 x 10 12 1.25 x 10*II 28 + . + . + . + . - 91 9 91, M 2.31 x 10*II 3.08 x 10 15 29 + - + - + . - + + + 91 9 30 + - + - + - - + + - + 91 9 91 P 2.82 x 10*I4 3.75 x 10'I" 86
/
/
Table 7-5. Quantification of Medium LOCA Event Tree (SSE Initiating Event) (Continued).
Event status ( ) TIME Rrou!REMENTs (31 stoutset FOR $YSTEM OPERATION PROBASILITY($)
- 5 =
- . % 2 g C 5 > J 7 g
u j l o a y
E 8
=
5
=
E 5
= l R
~
e 5
- O C g a 5 a a 5 0 8
=5 5 5 a 3 3 % I 3 % $ 0 g W pgtt gggy
$O 8 C D E F I J K N G H O 3 h $UPPORT $UPPORT 31 + - + - + - - + + . . 91 9 1 M 6.93 x 10-18 9.24 x 10-18 32 + . + - + - - + . 9 M 9.39 x 10*I4 9.77 x 10*I4 33 + - + - + - . - 9 M 9.49 x 10-16 1.03 x 10*I4 34 + - + - - + 100 D 100 P 3.35 x 10*I 3.69 x 10*I 35 + - + - - - 100 0 jo0 0 M 36 + . . + 100 0 100 P 5.68 x 10*I 7.58 x 10-5 37 + - - . 100 0 M M 38 - + + + 0 100 39 - + + - D 1 M 1.56 x 10~0 4.96 x 10 4 40 - + . + + + D 91 9 8
41 - + . + + - D 9 M 4.67 x 10*I 1.98 x 10 42 . + - + - + + 0 91 9 43 . + . + - + . D 9 M 4.67 x 10*I3 1.98 x 10-12 44 - + . + . . + + + D 91 g 45 . + - + - - + + . D 9 M 1.40 x 10-15 5.94 x 10 15 46 - + - + - - + - D 9 M 4.66 x 10-15 1.55 x 10*I4 47 . + - + - - . D 9 M 4.72 x 10*II 1.64 x 10 15 48 . + - - 0 1 M 1.67 x 10*8 6.00 x 10-8 49 . . D D M 2.82 x 10-8 1.20 x 10-5 M 3.55 x 1J*0 1.31 x 10*4 P 3.09 x 10*3 1.16 x 10*2 i
- See Notes. Table 7-3.
i r
m.
87 i.,
Table 7-6. Quantification of Large LOCA L2 Event Tree (0perator Error Initating Event With No LOSP)*.
EVENT STATUS (2) TIME REQUIREMENTS b# - FOR SYSTEM OPERATIO 3; SEQUENCE PR08A81LITY I5)
.W 5 =
x o -
= 5 I 5 g -
= a W
~
= .v .
u E -
5 -
-* 5 a W v
W w g ') -
e g w W a f
g NO LOSP.
putt NO LOSP, gaty
$ A 8 C D E F G H 0; w SUPPORT SUP90RT ~
1 + + + + + H 89 2 + + + + - +
l.
100 89 P J.00 r 10'3 4.00 x 10~3 3 + + + + - - 100 89 -6 M 1.64 x 10'0 2.18 x 10 -
4 + + + - + 100 0 89 P 4.76 x 10-5 6.93 x IC'3
+ + -6 5 + - - 100 0 g N 2.59 x 10-8 3.78 x 10 6 + + - + + + 100 4
7 + +' - + + - + 99 -5 -5 1 99 P 3.13 x 10 3.13 x 10 8 + + - + + - - 99 1 99 M 7.76 x 10*I 7.76 x 10 9 + + - + - + + g 91 10 + + - + - + - + 91 9 91 P 9.39 x 10-8 1.25 x 10'I 11 + + - + - + - - 91 9 91 M 2.31 x 10'" 3.08 x 10""
12 + + - + - - + 100 0 100 P 3.35 x 10'I 3.69 x 10*I
+ +
13 - + - - - 100 D g M 8.31 x 10 " 9.15 x 10'"
14 + + - -' + 100 0 100 P 4.81 x 10*I 7.00 x 10-5 15 + + - - - 100 0 g H 1.19 x 10-10 1.74 x 10-8 16 + - + + +' O 100 17 + - + + - 0 1 M 1.55 x 10 6 -6 3.26 x 10 18 + - + - + + 0 91 9 19 + - + +
0 9 M 4.65 x 10 1.30 x 10-8 20 + - + - - 0 9 H 1.66 x 10-8 3.94 x 10'8 21 + - -
0 0 M 2.38 x 10~8 7.28 x 10
22 - + + + 100 89 Pt 23 - + + - 100 Of Mo 24 - + - + + 100 9 Pt 91
+'
25 - - + - 100 9 E Mo 26 - + - - + 100 0 100 Pt 1.00 x 10-2 1.00 x 10-2 4
27 - + - - - 100 0 100 Mo 2.48 x 10** 2.48 x 10**
28 - -
0 M 5.00 x 10-6 1.05 x 10 5
< 1.07 x 10 5 M 2.96 x 10-5 I P 1.31 x 10 2 2.10 x 10 2
- 5ee Notes. Taele 7 3.
+Coseined due to initiating event isoact.
oCasef ned due to igittating event tapact.
s 88
- ,. . . , , , , . , , . ,.~m - , _ - , _ . - ---.,-,,.,_-,,_,,,,..-.m_ .~,,-_,,,,,m_ _v.- -_- ~ ,.,_- --y -., , ,. - -
Table 7-7. Quantification of Large LOCA L2 Event Tree (Operator Error Initiating Event Plus 1 Hour LOSP).
3 EVENT STATUS (2) TIME REQUIREMENTS y' SEQUENCE g FOR SYSTE's OPERATIO PRCB ASILITY ($)
R N & n W. '
O b v E E , q $
u e e E e z W E W W
H E E "
- Q a E. FULL HALF
- A B C D E F G H O E G E h SUPPORT $UPPORT 1 + + + + + 11 89 2 + + + + - + 100 89 P 3.00 x 10'3 4.00 x 10' 3 + + + + -6
- - 100 89 M 1.64 x 10-6 2.18 x 10 4 + + + + 7.05 x 10 5 100 0 89 P 8.42 x 10'
+ + + -6 5 - - 100 0 89 M 3.84 x 10'8 4.59 x 10 6 + + - + + + 100
+ + + + + 3.13 x 10 5 7 - - 99 1 99 P 3.13 x 10' 8 + + + + 7.76 x 10'9
- - - 99 1 99 M 7.76 x 10
9 + + - + - + + 91 9 10 + + - + - + - + 91 9 91 P 9.39 x 10-8 1.25 x 10'I
+ + + + 2.31 x 10'II 11 - - - - 91 9 9L M 3.08 x 10*II 12 + + + + 3.35 x 10'I
- - - 100 0 100 P 3.69 x 10' 13 + + + 8.31 x 10*II
- - - - 100 0 100 M 9.15 x 10' I 14 + + - + 100 0 7.12 x 10'I 100 P 8.50 x 10 15 + + - 100 1.77 x 10-10
- - 0 100 M 2.11 x 10' 16 + - + + + 0 100 17 + + + 1.56 x 10-6
- - 0 1 M 4.96 x 10' 18 + - + - + + 0 91 9 19 + - * - + - 0 9 M 4.67 x 10 1.98 x 10' 20 + - + - - 0 M 1.67 x 10-8 6.00 x 10-8 9,
21 + - -
0 0 M 3.54 x 10-8 1.35 x 10 22 - + + + 100 89 Pt 23 - + + - 100 89 Mo 24 - + + + + 100 9 91 Pt 25 - + + + - 100 9 E Mo 26 - + - - + 100 0 lu0 Pt 1.00 x 10*2 1.00 x 10-2 27 - + - - - 100 0 100 Mo 2.48 x 10'0 2.44 x 10 23 O M 5.04 x 10-6 1.60 x 10 5 l' M 1.08 x 10-5 4.38 x 10' P 1.31 x 10-2 2.26 x 10'
'See Notes. Table 7 3.
- Comined cue to initiating eveet impact.
oCombined due to initiating event impact.
89 i
t
/ / / / 03 g,0t n (g t g,01 m yt t g 03 : CS*l g 01 m 96 t / / / / 61 g.0L E tl*S g.01 m gg t / / / / at g.0L u tl S g 01 8 59't / / / / lt g.01 m pt S 01 I $9'I / / / / 91 9 03 St*t ,,0t u DZ't / / / / St
/ / / / pt 9 03 St t ,,01 m pg t g 01 8 SE*1 ,,01 E'DZ't / / / / tt 9 0L u 9p g g,01 m 62*l / / / / ZI g.01 E 6Z*l / / / / II 9 01 8 9t*9 It g / / / / Og 9 0L v gy g g,01 g.0t 8 16*S g 01 m ll*t - / / / / 6 o cn
/ / / / g g.01 m 16*S g,0g u gg t g.01 m 90 9 g,01 E 96*t / / / / l g.01 x LS*t / f / / g
,,0g u g0 3 g.0L
- iS*t 9 0L u 30 g f f f / g g.0L
- ll*t ,,01 m go'D / / / / t g.0L E 80*t ,,01 m gp i / / / -
/ t g.01
- 10*t 9 01 2 6t*t / - / / / 2 g 0t a g2*t ,,01 OS't / / / / I (S)ammatvig (*)nueva I 2 nin en "l* u 1 1 ag 00t >g Ot ag i Sea ou aetendo isS g ar m al 3t w aao) Jo (t) 1m J W 8358J10 lo 5501 (Z)3"II '3 m eA-Jol>eag aad angi pappy dlII"**3 lleg3 gut K18tl4 W J Paluul35] V301
- (S apow) umop2ngs pto3 ut paie plul y301 e 6uimotto.4 po. pad anoH 00t e Supna K2tligegoJd llaw aao3 aamuns3 02 suopelnate3 Jo KaeununS '8-4 atqe1 e'
____________.____m_.-_..__ _ _ _ - _ . _ _ _ _ _ _ - _ _ _ . . _ _ _ _ _ _ _ ._. _ .__m.___ - _____-__._m_ -A
Table 7-8. Summary of Calculations to Estimate Core Melt Probability During a 100 Hour Period Following a LOCA Initiated in Cold Shutdown (Mode 5)
(Continued).
Notes: (1) Calculations were performed for LOCAs initiated by a safe shutdown earthquake (SSE) or by operator misalignment of valves capable of diverting water from the reactor coolant system during Mode 5 operations.
(2) Some event sequences were potential successpaths only after a specified time in Mode 5 following shutdown. These paths were nonsuccess paths during the initial period of Mode 5 operation.
When the added time constraint was considered (e.e., time after shutdown greater than limittneti effective makeup water source.
(3) L = 1arge LOCA with two lulR pumps initially operating.
2 Lg = la m m wm one M pop inttlaHy watW.
3 M = nedlun LOCA.
(4) The numerical results listed in this column apply when no equipment is out of service for '
maintenance. Only random failures affect the availability of support systems.
(5) The numerical results listed in this column apply when one support system train is unavailable because of maintenance (e.g.. one diesel generator being overhauled, one CCW heat enchanger being inspected). Restoration of the unavailable support system train following the LOCA is not considered.
Table 7-9. Probability of Support System Availability.
Assumptions Full Support Half Support No Support No LOSP - No Maintenance .99178 .00820 .00002 LOSP 1 hr - No Maintenance .91926 .07714 .00360**
LOSP 10 hr - No Maintenance .87104 .12283 .00613**
LOSP 100 hr - No Maintenance .50844 .40922 .08234 No LOSP - Maintenance * -- .99588 .00412 LOSP 1 hr - Maintenance * --
.95783 .04217 LOSP 10 hr - Maintenance * --
.93246 .06754 LOSP 100 hr - Maintenance * -- .71305 .28695
- Maintenance affecting one support system train is assumed to be in progress (e.g., one diesel generator being overhauled, one CCW heat exchanger being inspected). The affected support system train is assumed to remain unavailable throughout the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> post-LOCA period.
- Conservative values due to combination of probability of 2nd loss of offsite power with the no support probability.
92 l
l
- 8. SumARY, CONCLUSIONS, AND RECOMENDATIONS 8.1 SUMARY AND CONCLUSIONS During cold shutdown, there are few automatic actions associated with response to a LOCA. The operator must, therefore, take a more active role in LOCA detection and mitigation than during power operation. There is considerable time available, however, for the operator to take required protective actions.
The estimated probability of core melt following a postulated LOCA
$ during cold shutdown was evaluated for 20 different cases with varying as- l
- sumptions regarding the LOCA initiating event (safe shutdown earthquake or operator error), time of LOCA initiation following reactor shutdown, LOCA 1
size, availability of offsite power during a 100-hour period following the I LOCA and maintenance status. With no maintenance in progress, the probability -
of core melt for those 20 cases was estimated to be in the range of 3.96 x 10-5 to 1.14 x 10-7 per reactor-year. If maintenance affecting one support
' system train is in progress (e.g., one diesel generator or component cooling water loop is out of' service) the probability of core melt for the 20 cases 4 increases and is estimated to be in the range from 7.53 x 10-5 to 8.46 x 10-6 per reactor-year. In contrast, an overall core melt probability of 6 x 10-5 was reported in WASH-1400 (Ref. 2). The estimates of core melt probability following a postulated LOCA during cold shutdown are believed to be very conservative and should be interpreted as upper limits, however, these values are nontrivial in comparison to the WASH-1400 results. .
The large LOCA was the " worst case" LOCA identified in this study. ;
Major contributors to core melt probability include the following:
e Initiating event probability 1
. e Probability of operator error during response e Probability of failure of an airbound RHR pump e Diesel generator reliability during long-term unavailability of 2
offsite power 93
Of these, only operator error during response may have been treated in a nonconservative manner. The large conservatism in the treatment of the other items, particularly initiating event probabilities, tend to suggest that the overall results are very conservative. A more detailed analysis would most likely indicate that the probability of core melt in Mode 5 is significantly lower, perhaps by an order of magnitude or more.
8.2 REC 0tNENDATIONS Based on this preliminary assessment of core melt probability in cold shutdown following a postulated LOCA, several operating practices and design changes are recommended for consideration as potential means for reduc-ing core melt probability.
8.2.1 Operating Practices e Secure Operating RHR Pumps Operating RHR pumps should be immediately secured following a LOCA in which pressurizer level indication is lost. This action is intended to minimize the probability of RHR pump failure due to cavitation ^ or airbinding. This operator action is not unlike the current practice of manually securing any operating main coolant pumps when required RCS temperature and pressure conditions cannot be satisfied.
e Align Centrifugal Charging Pump Suction to RWST The operating centrifugal charging pump suction should be realigned to the RWST to provide necessary reactor coolant makeup. Under the assumptions made in the analysis, there is a time dependency associated with obtaining adequate makeup from the normal makeup water source (primary water storage tank, with boration as necessary). There is no time dependency associated with the RWST, and it supplies highly borated water, therefore, the RWST should be the primary water source for LOCA response.
94
~
l e Align Centrifugal Charging Pump Discharge to ECCS fnjection Path ;
The normal CVCS return path to the RCS is a single, or at m'ost, a :
dual path. A LOCA in RCS loops 1 or 3 could divert a large portion f of the makeup flow (as much as 100 percent). By aligning the l
~
centrifugal charging pump to the ECCS injection path, four parallel injection paths to the RCS are provided. A LOCA would be expected ;
to affect only one of these paths, thus a maximum of 25 percent of i the makeup water would be diverted from the reactor vessel.
I e Use Only a Single High-Head Pump For Coolant Makeup The large and medium LOCA definitions in Table 3-2 are based on the flowrate of a single high-head makeup pump at 600 psig (RHR system design pressure). Use of more than one high-head pump for makeup is not necessary for adequate core cooling. In addition, the range of break sizes falling into the medium LOCA category will increase if more than one high-head pump is providing makeup. It is likely that the probability of overpressurizing the RCS will increase.
I e Use Only Makeup Water That is Adequately Borated Boron dilution accidents were not considered in this analysis, however, such accidents may be credible if extended use is made of unborated water (e.g., fire water, condensate, etc.).
e Restore Containment Isolation If an equipment hatch is open, it may take a considerable period of time to restore containment isolation. It is therefore important that efforts to restore containment isolation be initiated in a timely manner to limit potential offsite radiologia1. consequences.
A preferred operating practice would be to have the equipment hatch open only when required to move equipment that cannot pass through the more easily isolated personnel airlocks. This operating prac-l tice is implemented by several utilities, and should aid in minimi-zing the probability that the containment could not be reisolated following a LOCA ia Mode 5.
l 95 ;
l
.. .- --- _ - - - . . - . . - - - - - - . - . L l
8.2.2 Design Changes 1
e RHR Pumps Designed For Airbound Operation s In Reference 7, it was noted that RHR pumps are not currently re- l quired to be designed for operation under no-suction (e.g., air-bound) conditions. It was further noted in Reference 7 that it I
would be desirable to design RHR pumps "so that they will not fail because of airbinding, with this ability verified by tests." With L this design capability, it would not be necessary to initiate a manual protective shutdown of the RHR pumps following a LOCA in
. which pressurizer level indication is lost. The large LOCA event trees in Figures 5-1 and 5-4 would be simplified by the deletion of Event A.
s Reactor Vessel Wide-Range Level Indication A wide-range reactor vessel level indication capability would pro-vide useful information to the operator and would aid in making optimum use of available water resources, including the cold leg and UHI accumulators. It is possible that this monitoring capability would allow the LOCA source to be identified, or at least localized ,
to one RCS loop. This capability would be particularly important if
- the RHR sys' tem could not be aligned for RHR sump recirculation, and core cooling had to be provided by the CVCS in a feed-only mode.
This type of level instrument could be isolated during normal reac-tor operation if necessary, and only placed in service at some suitable time during the RCS cooldown sequence.
e Redundant RHR Suction Line Safety Valves The Sequoyah RHR system has a single RHR suction line with a single safety valve to protect the RHR system suction piping against over-pressurization. Redundant suction safety valves could reduce the failure probability of Event J, and the subsequent challenge of the RHR suction isolation valves and the pressurizer PORVs for RHR and RCS overpressure protection.
4 96
- , -+ --e ,-, -. -
e Power-Operated Relief Valve Variable Setpoint Capability When the RHR suction line isolation valves are shut in Mode 5, the RCS is left without an overpressure protection system that is de-signed to limit RCS pressure rise commensurate with the applicable temperature-pressure limits (see Figures 3-6 and 3-7). Main coolant pump seal leakage and any coolant letdown via the CVCS high pressure letdown path may aid in limiting RCS pressure under this condition, however, their effect is uncertain.
One approach for providing additional RCS overpressure protection in Mode 5 is to add a dual-setpoint capability to the control circuitry for the pressurizer power-operated relief valves (PORVs). During power operation, these valves would be set to lift at their normal setpoint of 2350 psig. At some suitable point in the RCS cooldown sequence, the higher-pressure setpoint of the PORV control circuit could be bypassed and the lower-pressure setpoint activated. .The normal higher-pressure setpoint.would be restored during the subse-quent startup sequence. This design feature could reduce the fail-ure probability of Event N in the medium and small LOCA trees by adding an automatic PORV actuation capability in parallel with the existing remote-manual capability.
97
/
/
.m 8
6 98
. .... . 'F'
- 9. REFERENCES
- 1. NUREG-0789, " Technical Specifications, Sequoyah Nuclear Plant, Unit 2, Docket 50-328," U.S. Nuclear Regulatory Comission, September 1981.
- 2. WASH-1400, " Reactor Safety Study," Appendix III, U.S. Nuclear Regulatory Commission, October 1975.
- 3. B-SAR-205, " Babcock & Wilcox Standard Nuclear Steam Supply System,"
Docket STN-50-561, Babcock & Wilcox Company.
- 4. Sequoyah Nuclear Plant FSAR, Dockets 50-327, 50-328.
- 5. San Onofre Nuclear Generating Station, Units 2 and 3 FSAR, Dockets 50-361, 50-362.
- 6. NRC Inspection and Enforcement Report No. 50-327/81-07, " Containment Spray Event of February 11, 1981, TVA Sequoyah Nuclear Plant."
- 7. Baslik, A. J. and Bari, R.A., " Risk Reduction From Safety-Grade Means of Reaching and Ma'intaining Cold Shutdown," BNL-NUREG Informal Report, Brookhaven National Laboratory, January 1981.
1 l
l l
99 g - - r -, wv-
4
{
4 s
100 o - -
e v. - ~e. -r a ,w .,7+.~ raw,.,-..~
e - *-n - , y s -, .-.e-. ,,,, - -
m -, - ~ , - 2 , -- , =- - - ,-p w- ,
APPENDIX A - FAULT TREES This appendix contains the fault trees associated with the large, medium and small LOCA event trees described in Section 5. Basic fault tree symbols are shown in Figure A-1. Fault trees for the following events are included in the remainder of this appendix:
Event A: Failure to secure operating RHR pumps (Figure A-2)
Event B: Failure of 2-of-2 C.VCS pump common (Figure A-3)
Event C: Failure of CVCS alignment to RWST (Figure A-4)
Event D: Failure of 2-of-2 RHR pump common (Figure A-5)
Event E: Failure of RHR alignment to RWST (Figure A-6)
Event F: Failure of 2.of-2 SI pump injection (Figure A-7)
Event G: Failure of RHR alignment to recirculate (Figure A-8)
Event H: Failure of CVCS normal makeup (Figure A-9)
Event I: Failure of operator to control pressure (Figure A-10)
Event J: Failure of RHR relief valve' to lift (Figure A-10)
Event K: Failure of RHR isolation alignment (Figure A-11)
Event N: Failure of operator PORV actuation (Figure A-12)
Ever.c 0: Failure of RHR relief val ~ve to close (Figure A-12) 5 Al
AND n
OR O
BASIC EVENT O
EXTERNAL EVENT (CONDITIONAL)
A TRANSFER OUT A
TRANSFER IN
\
Figure A-1. Basic Fault Tree Symbols A2
. ~ _ . , , . .
7,~.
- ll !
I 9
TS O
NP M
SU EP D
E I O DR RR O
O I
EC P PE u OS R
R E
T$
N I IP M
R MO A E PT R
U t C
E R
H E M n S e O
I R T S
l OS T I I
F O S Y v P E EM E L Ru RSO UP u R r L
IR AH E ILPT AUO MN o FR I
T FPC f A
R E e P
O e I
8 r
T tT lR uA t PT S l R
5 P HT TA u
NP a
UP C M F ER 5U O
SH I - EP R I 0 O I 0R TG H N RR .
EI O RT UA D TE AR 2 LR E RU -
PE OS A
FO R R e E
TA r I
N IP M g u
TU UP i N
R F RH ER W
OO E PT R
U R M C
E H E O
S I R T I I S OA I F
O Y
S T
P EB I
E L RL RAO UP U R L
IAH R g i
LrT IAuOeN FR T FPC A
R E
P I
G A
eT R
sR tA PT S
R etT lA wco
, : , . i;: . , .;. !9 .3 5$ i 1
FAILURE OF 2.(4-2 CVC5 PlHP C0mVN O
T' I I I I FAlluRE Of CVC5 FAltuRE OF 2-0F-2 FAILURE OF VALVE FAILURE OF VALVE lautCTION PATHS CVC5 PUMP 5 ISOLATION SET A 150LAll0m SET S n n
- i i g failure Or nov 63-2510 OPLN FAliuRt OF nov 63-1410 OPEN f AILURE OF MOV 63-39 TO OPLM FAILURE 08 M0v 63-4010 OPEN
- O O O o I I l l
. r, a
FAILURL UF CHLCK FA! Luke OF CHECK f AILURE OF CHECK FAILURE OF CHLCK FAltuRE OF CHECK VALVE 63-581 VALVE 63-586 VALVE 63-587 VALVE 63-588 VALVE 63-569 10 OPLM TO OPEN 10 OPEN TO OPEN TO OPEN O O O O O Figure A-3. Fault Tree for Event B.
FAILURE OF 2-OF-2 CVC5 Pules O
- 1I I I FAILURE OF CVC5 FAILURE OF CVC5 Puff A 10 CPERATE PUMP B 10 CPERATE
'e
~ _
i l Ln FA L R OF HECE FAILURE OF CVCS FAILURE OF CVCS FAILURE OF CVCS 2 52 Ai 5 Puff A 10 Run 810 EN 2' 10 Opgg 10 OPEN
}
l O O O O O O Figure A-3. Fault Tree for Event B (Continued).
i
VN OE O
MP l O F
OO T
E R6 U3 L1 I -
S A2 P F6 NM Gu IP L
LT 5
IAW FR Vh .
OE C
O MP l O F
O0 t 1
E R5 n U3 e L1 IA2 - v F6 E K
C E r H o I
C4 F5 O-E6N R
UEP LVO 0
2 E
0 ,
f e
e r
T IAA0 FV1 L
T S .
SW t CA V l CO F
T u OT N
a EE F AM UN LG II AL T FA T5 OW .
NR 4 5O -
O ET I O DS A
C RV OC T
e r
AN u RG EI PL g
OA i.
E F VS DO l
ML F
O01 E
R3 C
O U3 IL1-E A2 T F6 A
L O
E RL UA LM R
IAO FN E
Vs Do O
Ml l C F
O0 1
E R2 U3 IL1-A2 F6 m
1l;i
FAILURE OF 2-0F-2 RHR PUMP C0f010N O
1
~
3 N I l l l l
F FAILURE OF RHR O! VERS!ON OF RHg FAILURE OF RHR FAILURE OF RHR DIVERSION OF RHR WE,0
, g PUMP A TO OPERATE PUMP A FLOW A Ilu[Cil0A Pupe 5 TO OPERATE PUMP S FLOW pggg3 A A A A A A I
Tr I I T' I t
mura mus mur0R mas mur0R mas m ur0R m as VALVE 72-40 VALVE 14-34 VALVE 12-41 VALVE 14-14 O
O O O Figure A-5. Fault Tree for Event D.
IAltuRE OF RHR FAlttaf 0F RHR PUMP A 10 OPERATE PUPF 810 OPERATE A n A 3 I I I I
TAILUBE 10 STARg FAILURE CF RHR FAILUP! TO START FAlluRE OF RHR PUMP A TO RUN RHR PUMP 5 FLOW PUMP B 10 RUN
. SHR PUMP A FLOW 1
li O O I
TF II I I l l
! CO RNR PUMP 8 SECURED FAILURE 10 SEGIN FLOW RHR PUMP A SECURED FAILURE TO BEGIm FRGM RHR Ptf@ S OR NOT OPERAllNG FLOW FRUM RHR PUMP A OR NOT OPLAAlthG O b TT' I I I I f ALLURE OF CHECK FAILURE OF CHLEK FAILURE OF RHR yALyt 74 515 FAltukE Of kl8 VALVE 74-514 PUMP A IO START PistP B 10 START g
10 QPEN 10 OPLM 4
O O O O Figure A-5. Fault Tree for Event D (Continued).
.A
\
4
5, 5.
37 37 On wOm
!W5 EWE E Efa E!a i=
C e
C O
(J w
u y W~
v Wm u
- ~ - o- C h
wem >
w W Sw5
- a. Nemm.5 W
8 8 2 IE-20 e E 2" W 32- .- 6 m
3*, 3"e H M
ffs.
Nh!.. -
3 3g =
_ _ g m
- B. - 3
, Om m wO=
!=5 !wE -
Eis Eis j Q
L 3
CD w
b
. W W
W
_ -- _ g
.3 8.
o ;, **=
wen 4$ *Wo WWo
... ==-
A9
. _ . . . .s , . . . ,
FAILURE OF Ret ALIGelEmi TO RW5T
,\
'N A j 3 x -
.. 4
'd 1
d SUCil0N PATH FROM RCS OPERATOR DOES NOT IM TO RW51 FAILURE TO ALIGN h5! 10 RNR PUMP 5
- -)
s O __
7 l I I I FAILUPE TO ALIGN 3 FAILURE OF MOV FAILURE OF MDV Rwst ygg gCES FAILURE 10 ALIGN RwSI VIA CONTAlffthi fall:1RE TO ALIGN Rid 5T WIA CaliTAIMENI o 74-1 TO CLO5E 74-2 TO CLOSE MAEEUP PATM 5pRAg PAIN 4 SPRAf PAIN 2
- O O A e ,
Tr I I FAILURE OF MDV FAILuNE OF CHECK 63-1 TO OPEN VALVE 63-505 TO OPEN O
Figure A-6. Fault Tree for Event E.
i O
l
I FAILURE TO ALIGN HW5T VIA COhTAINNT SPRAV PATH I A
i
't l l
-J J
. . 3 FAILURE OF CHECK FAILURE OF MV ' FAILURE OF NDV 72-22 TO OPEN 72-2310 OPEN IL O
- O O O b FAILURE TO ALIGN RWST VIA Cob 1AIMNT
' "4 li SPRAY PATH 2 n
e%
l I FAILURE OF MV FAILURE OF CHECK FAILURE OF MV 72-20 TO OPEN 72-21 TO OPEN
{0 0 0 0 Figurek-f3. Fault Tree for Event E (Continued).
T K C C E E
!J H A O
$N C7 6
I FI OO l
F5 O -
T 3 E E6N RS 0 E U 9EP ILP M ILVOL T Au AAO 2C FP FVT FE 0 -
l 2I0 I F1 1 O
5 EP N
UW LP IAI FS T K C C E
Su I
H l
Fl O0 E
RA U
LP T A l C5 F5 O-E6N R
UEP 5
3 E
O F IAu p ILVOL t P
M FP AAO FVT n u
P5 _
e tI M v SA E l P F
ON O
r EI RT f, o
uC tE lJ 2N AN e
- O FI
' K C
e FI 0T E H
r T
O
- C C3 2E I 5 FN OI F5 O-3 t LP E6N l RM S LEP E
u Uu LP ILVOL a IA1 F5 T
AAO FVT F
OT N5 5R W
l E
0O UT R1 O1 T
AN RG 0 7 A
e EI PL r OA K u V
C E
g O H o.
i MN O
C1 I
FP E I F5 5 F OO O -
3 EO E6N RT R E U UEP IL5 - ILVOL A3 AAO F6 FVT Ns IGr Lu e
l l
LT IS AW FR K C
o E
H C01 I
F5 O - 3 E6N R E -
UEP LVG _
L IAA0 FV1 ywN _
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I .! i6-
. P :' _
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I a
t 9
FAILURE OF 51
- , . PUMP A 10 IEJECT
<3 I I FAILURE OF St FAILURE OF MOV FAILURE OF 51 Pur A TO OPERAlt 63-47 TO OPEN PUMP A INJECTION PATH A O "
T TF 1 l l l l l B
w FAILURE TO RERim FAILURE OF 51 FAILURF OF 5g FAILURE OF CHECK II A FAILURE OF MDV FAILURE OF MOV PtetP A TO START PUMP A TO RLs VALVE 63-524 63-2210 OPEN 63-15210 OPEN 10 OPER O O O O O O f
i l
Figure A-7. Fault Tree for Event F (Continued).
i i
4 f
I
?
I
'f FAILURE OF 51 1 PUMP S TO INJECT em i I
. FAILURE OF 51 FAILURE OF M0W FAILUNE OF 51 i PUMP S 10 OPERAlt 63-48 TO OPEN PtMP S INJECTION PAIN
. 1 A "
? O T' II I I I I I I t FAILWE TO KRIM FAILURE OF SI FAILURE OF CNEEK FAILUNE OF SI FAILUNE OF MDW . FAILURE OF M0W Si M S PLMP S TO 5 TART PUMP 810 kWN 63-22 TO OPEN 63-153 TO OPEN
,j f0 O O O O O O 4
Figure A-7. Fault Tree for Event F (Continued).
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APPENDIX B - ESTIMATION OF LOCA LEAK RATES This appendix contains the code listing and the detailed results of the calculated leak rates from various size pipe breaks postulated to occur in Mode ' 5. The basic geometry of the problem can be visualized as a large diameter-RCS loop pipe (e.g., at least 27.5 inches in diameter) with a smaller
. diameter connecting pipe (e.g., 0.5 to 14 inches in diameter) that has a guillotine break either one foot or 100 feet from the interface with the RCS loop piping. Initially, the RCS is assumed to be at 600 psig (RHR system design pressure) and 2000F (upper limit for Mode 5). Additional assumptions include the following:
o All pipes are hydraulically smooth o The geometry of the RCS loop pipe and the connecting smaller diam-eter pipe can be considered as an ordinary entrance.
o Choke flow conditions do not exist The code listing appears in Table B-1. This calculational routine is based en the mechanical energy balance equation as presented in Bennett and Meyers(1) and Perry (2). The results of the leak rate calculations for eleven different pipe sizes (0.5, 0.75,1.0,1.5, 2, 3, 4, 6, 8,10, and 14 inch actual pipe inside diameter) are found in Table B-2. The calculated leak rates are sumarized graphically in Section 3. .
. (1) Bennet, C. O. and Meyers, J. E., Momentum, Heat and Mass Transfer, McGraw Hill, 1962. :
(2) Perry, R. H., Chilton, C. H., and Kirkpatrick, S. D. (editors), Chemical l Engineering Handbook 5th Edition, McGraw Hill,1973.
I B1
/
= ,- __-
Table B-1 Code Listing for Estimating Mode 5 LOCA Leak Rates 00610 REAL*4 LWORE.EIN*.TC.LTERM 00920 DIMENSION A(4).C(6) 0: 0".3 DATA GC.E3TGP/00.17.00/
30040 DATA C/0.0306.-e.01034.-O.3309.O.4033.0.22E-04 d6030 1.-258.2/
00040 DATA A/-0.1367.0.0607.-0.04206.0.001028/
96970 FnICT(X1.".2)=C( 1 ) +C(0)*X1 +C(3) 'C &C(4) X1*X2 09080 1+C(5)*X1*XI+C(6)*X2*X2 00090 AbbX(X)=A(1)+X*(A(2)+X*(A(3)+X*A(4)))
C0100 OPEN(UNIT =32. DIALOG ='DSED PIPES.OUT')
00110 IOU =32 00120 C 00130 C 'I1IIS CODE CALCULATES 'I1IE FLUID VELOCITY THROUGH 00140 C A PIPE GIVEN SOME INITIAL PRESSURE AND 2ERO 00150 C VELOCITY. THE EXIT OF THE PIPE IS AT ZERO PSIG.
00160 C AN ORDINARY ENTRANCE IS ASSUMED. PIPE WALL ROUGHNESS.
00170 C E/D. CAN RANCE FROM 0. TO 0.0023 00100 C TIIE HECLANICAL EDEEGY BALANCE EQUATION IS USED AS 90190 C DESCRIBED IN: MOMElrTUM. HEAT AND MASS TRANSFER 00200 C BY C.O. liENNETT S J .E. MYERS. (1962).
00210 C PAGES 164 TIIn0UCII 160. POLYNOTII AL CTnESSIONS 00220 C FOR THE FRICTION FACTOR (PACE 166) AS A FU3CTION 00230 C OF REYNOLDS NUNBER AND PIPE ROUGHNESS. AND 03240 C EQUIVALENT LENGTH AS A FUNCTION OF INSIDE PIPE 00250 C DIAMETER WERE FITTED TO SOME DATA AND WRITTEN 00260 C AS FUNCTIONS AT THE BEGINNING OF TIIE CODE.
00270 C 00280 C JUNE 10 1982 00290 C 00300 TYPE 10 00310 10 FORMAT (tX.* ENTER MAIN PIPE PRESSURE. PSIC')
00320 ACCEPT 20 PSIG 00330 20 FORMAT (F10.9) 00340 TYPE 30 00350 30 FORMAT (1X.' ENTER PIPE LENGTH. INCHES')
00360 ACCEPT 20 STUDIN 09370 TYPE 40 00380 40 FORMAT (1X ' ENTER PIPE DIAMETER. INCHES')
00390 ACCEPT 20 STUDD 00400 TYPE 50 00410 30 FORMAT (1X.' ENTER PIPE WALL E/D*)
00420 ACCEPT 20 ED 00430 TYPE 60 00440 60 FORMAT (IX.' ENTER WATER VISCOSITY. CP')
- 00450 ACCEPT 20. CP 00460 TYPE 70 00470 70 FODMAT(IX.' ENTER WATER DENSITY. GM/CC')
00480 ACCEPT 20 DEN 00490 C 00300 C ECHO '!IIE INPUT 00310 C 00320 80 TYPE 90 PSIG 005CO 90 FORMAT (/,1X.'t. MAIN PRESSURE = ' 1PEIO.3.' PSIC')
00540 TYPE 100 STUBIN 00550 100 FORMAT (1X.'2. PIPE LENGTH = '.F7.2.' INCHES *)
00560 TYPE tie. STUBD 00570 lie FORMAT (tX.'3. PIPE DIAMETER =
- F6.2.' INCHES')
00500 TYPE 120. ED 00590 120 FORMAT (IX.'4. PIPE WALL E/D = '.F8.5) 00600 TYPE 130. CP 00610 130 FORMAT (IX.*5. WATER VISCOSITY = '.F6.2.' CP')
00620 TYPE 140. DEN 00630 140 FORMAT (IX.'6. WATER DENSITY = '.F6.2.' GM/CC')
00640 TYPE 150 00630 150 FORMAT (/.1X.*WANT TO CHANGE ANY?')
00660 ACCEPT 160 ANS B2
. . s , . _ , . . , - -
-, w Table B-1 (Continued) 00670 150 FORMAT (Al) 00600 IF(ANS.CO.*N') CO TO 250
&vusu T~stE $7u
'JU700 170 FonMAT(/.1X. ' ENTER TIIE LINE UUMBER TO DE CIIANGED')
0o710 ACCEPT 180 LINE 00700 100 FORMAT (Ii) 00700 CO TO (190.200.010.020.200.240) LINE-00740 190 TYPE 10 00730 ACCEPT 20 PSIG 00760 GO TO 80 00770 200 TYPE 30 00700 ACCEM 20. STUBIN 00790 GO TO 80 09800 210 TYPE 40 00010' ACCEPT 20 STUDD 00020 GO TO 80 00830 220 TYPE 30 JoPrio ACCEPT 20 ED 09G50 GH TO 80 00860 230 TYPE 60 00370 ACCEPT 20 CP 00000 GO TO DO 09890 240 TYPE 70 09900 ACCEPT 20. DEN on910 GO TO 80 00920 C 09900 C NOW START TIIE CALCUTATION 00940 C 00950 230 VIS=(6.72E-04)*CP 00960 DENSTY=62.4* DEN 00970 DP=144.*PSIG 00980 DIA=STUDD/12.
00990 AREA =0.703*DIA*DIA 01000 GTERM=448.86* AREA 01010 C 01020 C CALCULATE EXTRA LENGTH DUE TO LTTRANCE.
01030 C 01040 XLENTH=ADDX(STUDD) 01030 XLENTII= 10. **XLEIITII O1060 T(yraLX= XLENTII+STUDIN/12.
Ot070 GCD=GC*DIA 01000 C 01090 C GUESS A VELOCITY = 10 IT/SEC 01100 C 01110 V=le.
01120 V2=V*V et130 RETERM=DIA*DENSTY/VIS 0ti40 RE=V*RETERM 0tt50 C 01160 C GET FANNING FRICTION FACTOR. SEE PAGE 166 0tt70 C OF BCINL'IT AND MYERS.
01100 C 01190 REL*ALOG10(RE) 01200 F=FRIC7(REL.ED) 01210 LTERM=2.*TOTALX/GCD 01220 LWORK=F*V2*LTERM 01230 GC2=2.*CC 01240 KINETC=V2/CC2 01250 PCALC=DENSTY*(LWORK+KINETC) 01260 C 01270 C BRACKET THE VELOCITY. >
01200 C 01290 K0UNT=0 01300 IDIR=1 01310 TEST =DP-PCALC 01320 IF(TEST.GT.O.) IDIR=2 ,
B3
.s . . ,.
l l
l l
Table B-1 (Continued) 01330 C 01340 C IDIR=1 NEARS DP-PCALC < 0 DECREASE V.
n 400 4 iisiL=2 i.L Jis ui'-i J/.I.0 ; S. iii;iGF., V. .
.)l 3'a 0 C 0t370 260 VOLD=V 01300 GO TO (070,0G0) IDIR 91300 270 V = 'u 2 .
01400 GO TO 290 01410 280 V=2.*V 01429 290 V2=V*V .
OI430 RE=V*RETERM 01440 REL=ALOC10(RE) 01459 F=FRICT(REL.ED) 01460 LVORK=F*V2*LTERM 01470 KIIIETC=V2/GC2 01480 PC LC=DEFFTY*(LFORK+KINETC) 01490 TEST =DP-PCALC 01500 I F t TEST. LT . O . . AliD . i DIR . EQ .0 ) CO TO 300 0151*) I F (TEST GT. O . . AND. I DIR . I'.Q .1 ) GO TO 300 0iG20 KOUNT=KOUNT+1 01330 IF(KOUIT.GT. ESTOP) CD TO 400 01360 GO TO 260 01550 C 01360 C NOW DO TRIAL AND ERROR TO GET THE VELOCITY.
01370 C 01300 000 KOUNT=0 01390 IF(V.LT.VOLD) CO TO 310 01600 VHIGH=V 01619 VLOW=VOLD 01620 GO TO 320 01639 319 VHIGH=VOLD 01640 VLOWsV 01659 329 V= (VHlGII+VLOW)/2.
01660 V2=V*V OI670 RE=V*RETERN 01600 REL*ALOG10(RE) 01690 F=FRICTtREL.ED)
'O1700 LNORK=F*V2*LTERM 01710. KItiETC=V2/CC2 01720 PCALC=DFliSTY*(LFORK+KINETC) 01730 TEST =DP-PCALC 01740 IF(TEST.GT.O.) CO TO 030 01750 VHIGH=V 01769 CO TO 340 01779 339 VLOWsV 01780 340 - KOUNT=KOUNT+1 .
- 01790 IF(KOUNT.LT. ESTOP) CO TO 320 01800 GPM=GTERM*V TYPE 350. GPM 01810 JlCOO . 330 F03:!ATt,.1X.* FLOW RATE =
- 1PE10.0.* GALL 9"iS/ MINUTE')
01030 C .
01840 C WRITE IT ALL OUT.
01059 C 01860 PCALC=PCALC/144.
OI879 WRITE (10U.360) PSIC.PCALC 01880 369 , FORMAT (/.5X.'SPECIFIED MAIN HEADER PRESSURE = '
01099 1.F8.2.' CALCULATED = '.F8.2.* PSIG')
91960 WRITE (10U.379) CP.VIS 91910 379 FORMAT (GX.' WATER VISCOSITY = '.F5.2.' CE.TIPOISE T 01929 1. OR ' lPE10.3.' LBS(M)/(FT*SEC)')
01938 WRITE (10U 380) STUBIN.STUBD
.F6.1.' INCHES. DI AMETER 01940 . 380 FORMAT (UX.* PIPE LENGTH =
01950 1 = '.F6.2.' INCHES')
01960 WRITE (IOU.390) ED 01979 399 FORMAT (5X.' PIPE WALL ROUGHNESS E/D = '.F8.5) 91989 tiRITE (10U.400) DEN.DENd7Y B4
Table B-1 (Continued) 01990 400 FORMAT (3X.' WATER DEUSITY = '.F6.4.' CM/CC. OR =
02000 1.1PE10.3.' LES(M)/FT**3')
00010 WRITE a10U.410) V.GPM 02020 410 F0HMAT(5X.'
WATER VELOCITY = '
.1PE10.3.' FT/SEC 02030 1. OR = .tPE10.3.' CALLONS/ MINUTE *)
02040 WRITE (10U.420) AREA 02050 420 FORMAT (5X.' PIPE FLOW AREA = '
.lPE10.3.' FT**2')
02060 WHITE (IOU.430) XLENTH.TOTALX 00070 430 FORMAT (5X.* EXTRA LENGTH
- DUE TO ENTRANCE = '.F6.1 02080 1 ' FT. Tt7FAL LENGTH = .1 PE10.3. ' FT' )
02090 WRITE (IOU.440) RE.F 02100 440 FORMAT (5X.'REYNOLDS NUMBER = ' .1PE10.3. ' . FRICTION 02110 1 FACTOR = '.0PF8.5) 02120 WRITE (IOU.450) ElNETC.LWORK 02130 450 FORf!AT(5X.'EINETIC TERM = ' 1PE10.3.'. LOST WORE 02140 1 TERM = '.1PE10.3.' LDS(F)*FT/LDS(M)')
02130 TYPE 460 02160 460 FORMAT (/.1X.'D0 IT AGAIN7')
90!70 ACCEPT 470. ANS 32130 470 FORMAT (Al) 02190 IF(ANS.EQ.'Y') CO TO 80 92100 CO TO 500 02210 480 TYPE 490 02220 490 FORMAT (/.1X. 'NO BRACKET')
00230 500 CONTINUE
' 02240 END b
B5
Table B-2 Output Data Summaries For Mode 5 LOCA Leak Rate Calculations Pipe Diameter 0.5 Inches
""T7!"!"D MAIIT !"'A"21 P"JTS"RT. = r00.90 CALCLAT"U = F9.99 PSIC
' t*MZ: !?C C !TY e 9.20 C":ITIPOI"E. On 2.0 J C-t* i Ld:! (M)/ OT"SE)
Pil'n LE!Trli = 12.0 INCHES. DI ATIIT"R = 0.30 I NC*lFS PIPL WALL ROUGHNESS L'D = n.oe99o WATrn L)CESITY = . 9625 Cil/CC. On = 6.007E+01 LES (M VIT**3 UATER VELOCITY = 2.3370+02 IT/ CEC. OR = 1.46CE+02 GALLO;iS/MII;tTTE F f !C FL*
- AWEA = 1 . a '>:: E -"M FT:".2
".TQ
- LENT i CUE TO ;~,:iTI./G In = 1.1 TT, T T?AL Ls?, .a = C.3 N E-r00 7T ftEY:iuLDS I40*lDFit = 2.46..d+ W., FRICTIOTI FACT 0lt = 0.09
I r++ r a '- " 3.0 ;".6~;. t *yT n;ne Try,;t s 3, ~; s y.311; Ley. " > 4rT/L27(!;)
JP D_ i FI ED ilA ! ri HEAD 3 PRESSt~.E = Wo.00 CALCUL.C..is = 6?O.00 N IG
%suu VI -CMITY a o . 39 e s.t.Ti pO l 5E . on u 0 t ig-6. i.,ns t ,'I) / f iTuSLC )
I' ! I
- 7."'NT!I = 1200.o ITCitES. DI A?dTTJ1 = 0.G9 I NC~ M TG 4 h.6 v37 3 LLL CGUGE;;
m pg7y = ESC DLc;t 9<,35 = ,CC t';t = 6.097E+91 LES(v.)/ W 3 9ATrJ1 VELOCITY = 3.302E+01 IT/SEC. OR s- 3.233E+0! GALLO*fS/ MI:f77TE P I M. Flow A RCA = 1.uo3E-o3 FT*M FETDA LENGTil DUE TO FJITRANCE = 1.1 IT. TOTAL LENGTII = 1.01tE+02 FT ltEYUOLDS !fPMDER = 6.bO7E+03. FBICTION FACTOR = 0.00926 KiliLTIC TERM = 4.402E+0!. LOST WORE TERM = 1.394E+03 LBS(T)*FT/LES(M)
Pipe Diameter 0.75 Inches SPECIFIED ?IAIN IIEADER PEESSTPIE = 600.00 CALCULATED = 600.00 PSIG
'aATER VISPOSITY = 0.3o rEyrf POISE. 911 2.016E-94 Lt'S (M)/ OT4SEC )
P l F t. LENGTH = 12.9 t hCHE:= . DI Ai!LTEli = 0.73 INCilES PIPC WALL COUGRNESS E/D = 0.00000 VATER DCIISITY = . 4626 CM/CC. OR = 6.007E+01 LES(M)/FT**3 WATER VELOCITY = 2.306E+02 FT/SEC. On = 3.4GOE+02 GALLONS / MINUTE PI P E FLOW AREA = 0.066E-03 FT*F2 FErPA LE"CTU DUE TO Et1 TRANCE = ! .3 IT. TtYrAL LE7CTH = 2.309E+00 FT Tit'.?;0LPS in!;iDER = 4.600E+06. FRICTION FACTOR = 0.e0300 K!D" TIC TG 1 = 9. 7r..! E*00 . LOST WOP.K TEnM = 4. r'20E+-)2 LDS(F)"FT/L"S(ft)
SP~.CIFIED IIA N Iir.1DZR PPES31Hin = 600.00. CALCUIAT D = 6.'O.00 P31G WATr.R VI5COSITY = 0.30 CEiirtrolSE. Oil 2.0tez-04 L3S(M)/(FTrSEC)
F!PE LEPCTH = 1000.0 INCHES. DIAMETEIl = 0.73 INC;IES PIPC WALL ROUG1tNESS L'D = 0.00900 WATE!1 DENSITY = . 9626 GM/CC. OR = 6.007E+01 LES(M)/FT**3 HATER VELOCITY = 6.607E+01 FT/SEC. OR = 9.094E+01 GALLONS / MINUTE PIPE FLOW AREA = 3.066E-03 TT**2 EXTD A LENGTil DUE TO EtrTRANCE 1.3 IT. TfYTAL LENGTH = 1.013E+02 FT DEYNOLDS NU:! DER = 1.230E+e6 7RICTION FACTOR = 0.00312 KINETIC TEDM = 6.784E+01. LOST WORK TERM = 1.371E+43 LDS(F)*FT/LBS(M)
B6 4
- ws +- r
l Table B-2 (Continued) i Pipe Diameter 1 Inch SPF.CIFIED MAIN HEADER PRESSURE = 600.00 CALCULATED = 500.00 PSIC WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-o4 LLS(M)/(FT*SEC)
PIPE LENGTH = 12.0 INC11ES. DI A!!LTER = 1.60 I NQitS PIPE WALL ROUGHNESS E/D = 0.00000 WATER DENSITY = .9626 GM/CC. oil = 6 007E+01 LBS (M)/FT**3 WATER VELOCITY = 2.567E+02 FT/SEC. OR = 6.280E+02 GALLONS /ttINUTE PIPE FLOW AREA = 5.451E-03 FT**2 EhTnA LENGTI! DUE TO EIITRANCE = 1.6 IT. TOTAL LENGTH = 2.533E+00 FT RC'l!' OLDS NUIIDER = 6.373E+06. FRICTION TACTOR = 0.90330 KINETIC TERM = 1.024E+03. LOST WORK TERII = 4.143E+02 LBS(F)*FT/LES(M)
SPECITIED MAIN HEADER PRESSURE = 600.00, CALCULAT".D = MO.00 PSIG WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-04 LDS(M)/(FT*SEC)
PIPE LENGTII = 1200.0 INCHES. DI AHLTER = 1.00 INCHES PTPE WALL P.0UGUNE99 E/D = 0.00000 WATER DENSITY = .9626 GM/CC OR = 6.007E+01 LBS(M)/FT**3 WATER VELOCITY = 7.598E+01 FT/SEC. OR = 1.859t+w2 GALLONS / MINUTE PIPE FLOW AREA = 5.451E-03 FT**2 EXTRA LENGTH DUE TO ENTRANCE = 1.6 IT. TCTTAL LENGTH = 1.016E+02 FT REYNOLDS NUilBER = 1.007E+06. F3ICTION FACTOR = 0.00300 KINETIC TERM = 8.973E+01 LOST WORK TERM = 1.349E+03 LBS(F)*fT/LBS(M)
Pipe Diameter 1.5 Inches SPECIFIED MAIN HEADER PRESSURE = 600.00 CALCULATED = 600.00 P91G WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-04 LOS(M)/(f7*SEC)
PIPE LENGTII = 12.0 INCIIES. DIAMETER = 1.30 INCHES '
PIPE WALL ROUGHNESS E/D = 0.00000 WATER DENSITY = .9626 GM/CC. OR = 6.007E+01 LBS(M)/FT**3 WATER VELOCITY = 2.619E+02 FT/SEC. OR = 1.442E+03 GALLONS / MINUTE PIPE FLOW AREA = 1.227E-02 ET**2 EXTRA LENGTH DUE TO ENTRANCE = 2.1 IT. T(yrAL LENGTH = 3.128E+00 FT REY!! OLDS NUllBER = 9.756E+06. FRICTION FACTO!1 = 0.00349 KINETIC TERM = 1.066E+03. LOST WORK TERM = 3.720E+02 LDS(F)*fT/LES(M)
SPECIFIED il4IN HEADER PRESSURE = 600.00, CALCULATED = %0.00 PSIC WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-04 LT2S(M)/(FT*SEC)
PIPE LENGTH = 1200.0 INCHES. DIAMETER = 1.50 INCHES PIPE WALL ROUGHNESS E/D = 0.00000 WATER DENSITY = .9626 GM/CC. OR = 6.007E+01 LBS(M)/FT**3 WATEll VELOCITY = 9.078E+01 FT/SEC. OR = 4.998E+02 GALLONS / MINUTE PIPE FLOW AREA = 1.227E-02 FT**2 EXTRA LENGTH DUE TO EIITRANCE = 2.1 FT. T(FTAL LENGTH = 1.021E+02 FT REYNOLDS NUMBER = 3.381E+06. FRICTION FAC'IOR = 0.00313 KINETIC TERM = 1.281E+02. LOST WORK TERM = 1.310E+03 LBS(F)*FT/LBS(M)
B7
Table B-2 (Continued)
Pipe Diameter 2 inches
.SPECIFIED MAIN HEADER PRESSURE = 609.99. CALCULATED = 600.09 PSIC WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-04 LBS(M)/(FT*SEC)
PIPE LENGTH = 12.0 INCHES. DIAMETER = 2.99 INCHES PIPE WALL ROUCHNESS E/D = 0.99999 WATER DENSITY = .9626 CM/CC. OR = 6.997E+61 LBS(M)/FT**3 WATER VELOCITY = 2. 635E+02 IT/SEC, OR = 2.579E+03 GALLONS / MINUTE
. PIPE FLOW AREA = 2.18tE-92 FT**2 '
EXTRA LENGTH DUE TO ENTRANCE = 2.8 IT. TOTAL LENCTH = 3.800E+00 FT
'REYNOLDS NUMBER = 1.399E+07 FRICTION FACTOR = 0.00365 KINETIC TERM = 1.989E+03, LOST WORK TERM = 3.589E+02 LBS(F)*FT/LBS(M)
SPECIFIED MAIN HEADER PRESSURE = 600.09 CALCULATED = 600.00 PSIC WATER VISCOSITY = 0.39 CENTIPOISE. OR 2.916E-04 LBS(M)/(FT*SEC)
PIPE LENGTH = 1200.0 INCHES. DIAMETER = 2.00 INCIIES PIPE WALL ROUCHRESS E/D = 0.00099 WATER DENSITY = .9626 CM/CC. OR = 6.997E+01 LBS(M)/FT**3 WATER VELOCITY = 1.017E+02 FT/SEC. OR = 9.951E+02 CALLONS/ MINUTE PIPE FLOW AREA = 2.18tE-92 FT**2 EXTRA LENGTH DUE TO ENTRANCE = 2.8 FT. TOTAL LENGTH = 1.028E+02 FT REYNOLDS NUMBER = 5.049E+06. FRICTION FACTOR = 0.00322 KINETIC TERM = 1.697E+92 LOST WORK TERM = 1.278E+93 LBS(F)*fT/LBS(M)
' Pipe Diameter 3 Inches SPECIFIED MAIN HEADER PRESSURE = 600.99. CALCULATED = 600.99 PSIC WATER VISCOSITY = 0.39 CENTIPOISE. Oa 2.016E-94 LBS(M)/(FT*SEC)
PIPE LENGTH = 12.9 INCHES. DIAMETER = 3.99 INCHES PIPE WALL ROUGHNESS E/D = 0.99999 WATER nENSITY = .9626 CM/CC. OR = 6.997E+61 LBS ( M)/FT**3 WATER VELOCITY = 2.632E+92 FT/SEC. OR = 5.797E+03 CALLONS/ MINUTE PIPE FLOW AREA = 4.906E-92 FT**2 EXTRA LENCTH DUE TO ENTRANCE = 4.4 FT. TUTAL LENCTH = 3.367E+00 FT REYNOLDS NUMBER = 1.96tE+07. FRICTION FACTOR = 0.09391 KINETIC TERM = 1.977E+03. LOST WORK TERM = 3.616E+02 LBS(F)*FT/LBS(M)
SPECIFIED MAIN HEADER PRESSURE = 600.09 CALCULATED = 600.00 PSIC WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-04 LBS(M)/(FT*SEC)
PIPE LENGTH = 1200.0 INCHES. DI AMETER = 3.99 INCHES PIPE WALL ROUCHNESS E/D = 0.99999 WATER DENSITY = .9626 CM/CC. OR = 6.997E+61 LBS(M)/FT**3 WATER VELOCITY = 1.173E+02 FT/SEC. OR = 2.582E+03 CALLONS/ MINUTE PIPE FLOW AREA = 4.996E-92 PT**2 EXTRA LENGTH DUE TO ENTRANCE = 4.4 PT TUTAL LENCTH = 1.044E+02 FT 4 REYNOLDS NUMBER = 8.734E+96 FRICTION FACTOR = 0.09343 KINETIC TERM = 2.137E+92. LOFT WORK TERM = 1.225E+03 LBS(F)*FT/LBB(M) 1 i
I i
B8
Table B-2 (Continued)
Pipe Diameter 4 Inches SPECIFIED MAIN HEADER PRESSURE = 609.09. CALCULATED = 600.99 PSIC VATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-94 LBS(M)/(IT*SEC)
PIPE LENC'ITI = 12.0 INCHES. DIAMF1rER = 4.99 INCHES PIPE WALL ROUCHNESS E/D = 9.99999 WATER DENSITY = .9626 CM/CC. OR = 6.997E+01 LBS(M)/IT**3 WATER VELOCITY = 2.619E+92 IT/SEC. OR = 1.925E+94 CALLONS/ MINUTE PIPE FLOW AREA = 8.722E-92 FT**2 EXTRA LENGTH DUE TO ENTRANCE = 6.1 IT. T(TRAL LENGTH = 7.953E+00 FT REYNOLDS NUMBER = 2.601E+07. FRICTION FACTOR = 0.00412 KINETIC TERM = 1.966E+03. LOFF WORK TERM = 3.723E+02 LBS(F)*FT/LLS(M)
SPECIFIED MAIN HEADER PRESSURE = 609.09 CALCULATED = 609.99 PSIC WATER VISCOSITY = 0.30 CENTIPOISE. On 2.016E-04 LDS(M)/(IT*SEC)
P1PE LENGTH = 1299.O INCHES. DIAHETER = 4.09 INCHES PIPE WALL ROUCHNESS E/D = 0.99999 WATER DrUSITY = .9626 CM/CC. OR = 6.007E+01 LBS(M)/IT**3 WATER VELOCITY = 1.283E+02 IT/SEC. Oh = 5.924E+03 CALLONS/ MINUTE PIPE FLOW AREA = 8.722E-92 FT**2 E:CITIA LENGTH DUE TO EffrRANCE = 6.1 IT. T(TRAL LENGTH = 1.961E*02 FT REYNOLDS NUMBER = 1.274E+07. FRICTION FACTOR = f. 09363 KINETIC TERM = 2.559E+02. L0 Err WORK TERM = 1.182E+93 LBS(F)*fT/LDS(M)
Pipe Diamter 6 Inches SPECIFIED MAIN HEADER PRESSURE = 609.09. CALCULATEL,= 609.00 PSIC WATER VISCOSITY = 0.39 CENTIPOISE. OR 2.016E-04 LBS(M)/(IT*SEC)
PIPE LENGTH = 12.9 INCHES. DIAMETER = 6.99 INCHES
^
PIPE WALL ROUCHNESS E/D = 9.99999 WATER DENSITY = .9626 CM/CC. OR = 6.997E+01 LBd(M)/fT**3 WATER VELOCITY = 2.699E+92 IT/SEC. OR = 2.299E+94 CALLONS/MINUTZ PIPE FLOW AREA = 1.962E-91 FT**2 EXTRA LENGTH DUE TO ENTRANCE = 9.9 FT. T(YrAL LENGTH = 1.004E+01 IT REYNOLDS NUMBER = 3.887E+97. FRICTION FACTOR = 0.09447 KINETIC TERM = 1.058E+03. LOST WORK TERM s' 3.892E+02 LBS(F)*fT/LBS(M)
SPECIFIED MAIN HEADER PRESSURE = 600.00 CALCULATED = 600.00 PSIC WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-94 LBS(M)/(FT*SEC)
PIPE LENCTH = 1299.9 INCHES. DIAMETER = 6.99 INCHES PIPE WALL ROUCHNESS E/D = 0.99999 WATER DENSITY = .9626 CM/CC. OR = 6.997E+91 LBS(M)/fT**3 WATER VELOCITY = 1.439E+02 IT/SEC. OR = 1.268E+94 CALLONS/ MINUTE PIPE FLOW AREA = 1.962E-91 FT**2 EXTRA LENGTH DUE TO ElfTRANCE = 9.9 7"I . T(FrAL LENCTH = 1.999E+02 FT REYNOLDS NUMBER = 2.144E+97 FRICTION FACTOR = 0.99398 KINETIC TERM = 3.229E+02. LOErr WORK TERM = 1.116E+03 LBS(T)*f7/LBS(M)
, B9
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Table B-2 (Continued)
I i
l Pipe Diamter 8 Inches i
i SPECIFIED MAIN HEADCI PRESSURE = 600.00. CALCULATED = 600.00 PSIC WATER VISCOSI*IY = 0.30 CE*iTIPOISE. OR 2.016E-04 L3WI)/f FT*SEC)
P I P 5' LENGTTI = 12.0 INCHES. DI AffFTER = 8.00 INCTIES PIPE WALL ROUGHNESS E/D = 0.00000 ,
WATER DENSITY = .9626 CM/CC. OR = 6.007E+01 LBS(M)/f7**3 '
WATER VELOCITY = 2.615E+02 FT/SEC. OR = 4.095E+04 CALLONS/ MINUTE ,
PIPE FLOW AREA = 3.489E-01 FT**2 I t EXITIA LENGTH DUE TO EIITRANCE = 11.4 FT. T(YTAL LENGTH = 1.240E+01 FT l REY!! OLDS IfUMBER = 3.193E+07. FRICTION FACTOR = 0.00475 l
KIHLTIC TERM = 1.063E+03 LOST WOltK TERM = 3.739E+02 LBS(F)*FT/LBS(M)
SPECIFIED MAIN HFADER PRESSURE = 600.00. CALCULATED = 600.00 PSIG MTER VIOCO3ITY = 0.00 CE:.TIPOISE. On 2.016E-04 LO3GI)/(FT*SEC)
PIPE LENGTH = 1200.0 INCHES. DIAIIETER = 0.00 INCHES PIPE WALL ROUGHNESS E/D = 0.00000
'!?.T"R EC:fCITY = .0G20 CI/CC. OR = G.037 +01 LOC CI)/:T=3 WATER VELOCITY = 1.550E+02 FT/SEC. OR = 2.42BE+04 GALLONSeMINUTE PIPE FLOW GEA = 3.489E-01 77**2 -
! CCTRA LENGTH DUE TO EiTRANCE T = 11.4 IT. T(YrAL LENGTH = 1.114E+02 FT REYNOLDS NUMBER = 3.079E+e7. FRICTION FACTOR = 0.00426 KINETIC TERM = 3.736E+02 LOST WORK TERM = 1.06GE+03 LBS(F)*FT/LBS(M)
Pipe Diameter 10 Inches SPECIFIED MAIN HEADER PRESFJRE = 600.00 CALCULATED = 600.00 PSIG WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-04 LOSOI)/(FT*SEC)
PIPE LENGTII =
12.0 INC11ES. DIAMETER = 10.00 INCHES PIPE WALL ROUCHNESS E/D = *f.00e00 WATER DEllSITY = .9626 CM/CC. O!! = 6.007E+01 LBS(M)/IT**3 WATER VELOCITY.= 2.590E+02 FT/SEC. OR = 6.338E+04 CALLONS/ MINUTE PIPE FLOW AREA = 5.45tE-et FT**2 EXTTIA LENGTII DUE TO ENTRANCF = 14.9 IT. T(YTAL LENGTH = 1.587E+01 FT REYNOLDS NUMBER = 6.432E+07. FRICTION FACTOR = 0.00498 KIHETIC TERf1 = 1.043E+03 LD5T WORK TERM = 3.93GE+02 LBS(F)*FT/LBS(M)
SPCCIFIED IIAIN IIEADER PRESSimE = 600.00. CALCULAT.D = 600.00 PSIG WATZH VISCOSITY = 0.30 CEtiTIPOISE. OR 2. 0 t eE-04 LUS GI)/ ( FT*SEC )
PIPE LENGTH = 1200.0 INCHES. DIAMETER = 10.00 INCHES PIPE WALL ROUCHNESS E/D = 0.00000 WATER DENSITY = .9626 CM/CC. OR = 6.007E+0i LBS(M)/f7**3 WATER VELOCITY = 1.629E+02 FT/SEC. OR = 3.987E+04 CALLONS/MINITTE PIPE FLOW AREA = 5.45iE-9i FT**2 EXTRA LENCTH DUE TO ENTRANCE = 14.9 IT. TOTAL LENGTH = 1.149E+02 PT ~
REYNOLDS NUMBER = 4.043E+07 FRICTION FACTOR = 0.00458 KINETIC TERM = 4.126E+02. LOST WORK TERM = 1.026E+03 LBS(F)*fT/LBS(M)
~
B10
Table B-2 (Continued)
Pipe Diameter 14 Inches SPECIFIED MAIN HEADER PRESSURE = 609.99. CALCULATED = 600.00 PSIC WATER VISCOSITY =' O.30 CrJrTIPOISE. OR 2.016E-04 L3S(M)/(IT*SEC)
PIPE LENG*Ill = 12.0 INCHES. DIAMETER = 14.00 INCHES PIPE WALL ROUGHNESS E/D = 0.99999 WATER DENSITY = .9626 CM/CC. OR = 6.007E+61 LBS(M)/IT**3 WATER VELOCITY = 2.991E+02 IT/SEC OR = 1.003E+95 GALLONS /MINITTE PIPE FLOW AREA = 1.068E+99 IT**2 EXTRA LENGTII DUE TO ErrTRANCE = 62.7 FT. T(yTAL LENGTH = 6.372E+91 IT REYNOLDS NUMBER = 7.268E+07. FRICTION FACTOR = 0.09511 KINETIC TERM = 6.794E+02. LOST WORK TERM = 7.599E+02 LBS(F)*fT/LBS(M)
SPECIFIED MAIN HEADER PRESSURE = 600.09. CALCULATED = 600.00 PSIC e
WATER VISCOSITY = 0.30 CENTIPOISE. OR 2.016E-04 LDS(M)/(1T*SEC)
PIPE LENGTH = 1200.0 INCHES. DIAMETER = 14.09 INCHES PIPC WALL ROUCHNESS E/D = 0.u9000 WATER DENSITY = .9626 GM/CC. OR = 6.007E+01 LBS(M)/fT**3 WATER VELOCITY = 1.303E+02 IT/SEC. OR = 7.399E+06 CALLONS/ MINUTE PIPE FLOW AREA = 1.968E+99 FT**2 EXTRA LENGTH DUE TO ENTRANCE = 62.7 FT. TOTAL LENGTH = 1.627E+02 FT l
REYNOLDS NUMBER = 5.596E+97. FRICTION FACTOR = 0.09481 KINETIC TERM = 3.992E+02. LOST WORK TERM = 1.048E+03 LBS(F)*fT/LBS(M) 811 e_
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