ML19329D988
| ML19329D988 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/31/1967 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| References | |
| NUDOCS 8004090502 | |
| Download: ML19329D988 (40) | |
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TABLE OF CONTENTS s
6.
ENGINEERED SAFEGUARDS Section Page 6.1 EMERGENCY INJECTION 6.1-1 6.1.1 DESIGN BASES 6.1-1 6.
1.2 DESCRIPTION
6.1-1 6.1.3 DESIGN EVALUATION 6.1-5 6.1.3.1 Failurc Analysis 6.1-5 5.1.3.2 Emergency Injection Response 6.1-13 6.1.3.3 Special Features 6.1-14 6.1.3.4 Check Valve Leakage - Core Flooding System 6.1-14 6.1.4 TEST AND INSPECTIONS 6.1-15
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6.2 REACTOR BUILDING ATMOSPHERE C00 LING AND WASHING 6.2-1 6.2.1 DESIGN BASES 6.2-1 6.
2.2 DESCRIPTION
6.2-1 6.2.3 DESIGN EVALUATION 6.2-2 6.2.3.1 Failure Analysis 6.2-3 6.2.3.2 Reac. tor Building Cooling Response 6.2-3 6.2.3.3 Special Features 6.2-7
,< w 6.2.4 TESTS AND INFPECTIONS 6.2-7
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i 6.3 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS 6.3-1 6.
3.1 INTRODUCTION
6.3-1 6.3.2
SUMMARY
OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATION 6.3-1 6.3.3 LEAKAGE ASSUMPTIONS 6.3-2 6.3.4 DESIGN BASIS LEAKAGE 6.3-2 6.3.5 LEAKAGE ANALYSIS CONCLUSIONS 6.3-2
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000 00223 6-i Amendment 3
LIST OF TABLES Table Number Title Page 6.1-1 Core Flooding System Performance and Equipment Data 6.1-4 6.1-2 Single Failure Analysis - Emergency Injection 6.1-6 6.1-3 Emergency Injection Equipment Performance Testing 6.1-16 6.2-1 Reactor Building Cooling Unit Performance and Equipment Data 6.2-1 6.2-2 Reactor Building Spray System Performance and Equipment Data 6.2-2 6.2-3 Single Failure Analysis - Reactor Building Atmosphere Cooling and Washing 6.2-4 6.3-1 Leakage Quantities to Auxiliary Building Atmosphere 6.3-3 9)
LIST OF FIGURES Figure Number Title 6.0-1 Engineered Safeguards Systems 6.1-1 Emergency Injection Safeguards 6.1-2 Makeup Pump Characteristics 6.1-3 Decay Heat Removal Pump Characteristics 6.1-4 Decay Heat Removal Cooler Characteristics 6.2-1 Reactor Building Emergency Cooling j'
22// sessmei-6-11 Amendment 3 I
7 6.
ENGINEERED SAFEGUARDS x;
Engineered safeguards for the nuclear unit are provided to fulfill four functions in the unlikely event of a serious loss-of-coolant accident:
a.
Protect the fuel cladding.
b.
Ensure reactor building integrity.
c.
Reduce the driving force for building leakage, d.
Remove fission products from the reactor building atmosphere.
Emergency injection of coolant to the reactor coolant system satisfies the first function above, while building atmosphere cooling and washing satisfy the latter three functions. Each of these operations is performed by two or more systems which, in addition, employ multiple components to ensure operability. All equipment requiring electrical power for operation is supplied by the emergency electrical power system as described in 8.2.3.
Engineered safeguards are separated into two completely independent trains of equal capability.
Either train is capable of handling the entire emer-gency coolant injection and emergency cooling load.
Failure of either train can not affect the other.
Each engineered safeguards train, as (K[7 shown in Figure 6.0-1, envelops core flooding tanks, high pressure and 2
_j low pressure coolant injection, reactor building emergency cooling sys-tem and reactor building spray and iodine removal system with associated heat rejection system.
Applicable codes and standards for design, fabrication, and testing of components used as safeguards are listed in the introduction to Section 9, and seismic requirements are given in Section 2.
The safety analysis presented in Section 14 demonstrates the performance of installed equip-ment in relation to functional objectives with assumed failures.
Some of the engineered safeguards functions noted above are accomplished l2 with the post-accident use of equipment serving normal functions.
The design approach is based on the belief that regular use of equipment pro-vides the best possible means for monitoring equipment availability and conditions.
Because some of the equipment used serves a normal function, the need for periodic testing is minimized.
In cases where the equipment is used for emergencies only, the systems have been designed to permit meaningful periodic tests.
Additional descriptive information and design details on equipment used for normal operation are presented in Section 9.
Section 6 presents design bases for safeguards protection, equipment opera-tional descriptions, design evaluation of equipment, failure analysis, and a preliminary operational testing program for systems used as engineered safeguards.
N 225 Amendment 2 6. 0-1 l
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FIGURE 6.0-1 ENGINEERED SAFEGUARDS SYSTEMS
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SACRAMENTO MUNICIPAL UTILITY DISTRICT Amendment 3
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6.1 EMERGENCY INJECTION
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gj 6.1.1 DESIGN BASES The principal design basis for emergency injection is as follows:
Emergency core injection is provided to prevent clad melting for the entire spectrum of reactor coolant system failures ranging from the smallest leak to the complete severance of the largest reactor coolant pipe.
High pressure injection is provided to prevent uncovering of the core for small coolant piping leaks at high pressure and to delay uncovering of the core for intermediate-sized leaks. The core flooding system and the decay heat removal system (which provides low pressure injection) are provided to recover the core at intermediate-to-low pressures so as to maintain core integrity during leaks ranging from intermediate to the largest size.
This equipment has been conservatively sized to limit the temperature transient to a clad temperature of 2,300 F or less.
6.
1.2 DESCRIPTION
Figure 6.1-1 is the schematic flow diagram for the emergency injection and associated instrumentation.
Emergency injection fluid, pumped to the reactor coolant system during safeguards operations, is supplied in each case from the borated water storage tank.
This tank contains the volume of borated water necessary to fill the fuel transfer canal during refueling operations and is connected 2
to the injection pump suction headers by two lines.
Additional coolant for emergency injection supply is contained in core flooding tanks which inject coolant without fluid pumping as described later in this section.
Emergency injection into the reactor coolant system will be initiated in the event of a) an abnormally low reactor coolant system pressure of 1,800 psig or b) a reactor building pressure of 10 psig during power operation.
Either of these signals will automatically increase high pressure injec-tion flow to the reactor coolant system with the following changes in the 2
operating mode of the makeup and purification system described in Section 9:
(a) the high pressure injection pumps will start and come on the line, (b) the stop valves in each injection supply line to the makeup and decay heat pumps will open, and (c) the injection valve in each of four injection lines will open.
Emergency high pressure injection will continue until reactor coolant system pressure has dropped to the point where core flood-ing tanks begin emergency injection.
The flow characteristic curves for each makeup pump are given in Figure 6.1-2.
7-298 an*
Amendment 2 6.1-1
Emergency Injection The core flooding system is composed of two flooding tanks, each directly connected to a reactor vessel nozzle by a line containing two check valves 1l and one stop valve. The system provides for automatic flooding injection with initiation of flow when the reactor coolant system pressure reaches approximately 600 psig. This injection provision does not require any electrical power, automatic switching, or operator action to ensure supply of emergency coolant to the reactor vessel.
Operator action is required only during reactor cooldown, at which time the stop valves in the core flooding lines are closed to contain the contents of the core flooding tanks. The combined coolant content of the two flooding tanks is suffi-cient to recover the core hot spot assuming no liquid is contained in the reactor vessel, while the gas overpressure and flooding line sizes are sufficient to ensure core reflooding within approximately 25 sec after the largest pipe rupture has occurred.
The decay heat removal system (described in Section 9) is normally main-tained on standby during power operation and provides supplemental core flooding flow through the two core flooding lines after the reactor coolant system pressure reaches 135 psi.
Emergency operation of this system will be initiated by a reactor coolant system pressure of 200 psi or by a reac-tor building pressure of 10 psig during any accident.
The flow character-2 istics of each decay heat pump for injection are shown in Figure 6.1-3; each pump is designed to deli'ver 3,000 gpm flow into the reactor vessel 3l at a vessel pressure of approximately 100 psi.
Low pressure injection, with supply from the borated water storage tank, 15 using the decay heat removal pumps will continue until a low level signal 21 is received from the tank (minimum of,36 minutes at the maximum combined flow of all low and high pressure injection and reactor building spray 2l pumps of 10,000 gpm). At this time, the operator will open the recircula-tion valves controlling suction from the reactor building emergency sump, 1
and flow of coolant from the sump to the reactor vessel and spray headers will begin, closing the check valve from the borated water storage tank.
The decay heat removal pumps are located at an elevation below the reactor building emergency sump with dual' suction lines routed outside the reactor building to separate pump suction headers.
The borated water storage tank and spent fuel pool are also connected to each of these headers.
l The pumps are in individual compartments interconnected by a doorway seven l
feet above the floor. Each cpmpartment is provided with an individual sump I
and pump to handle minor pump seal and packing leakage and equipment fail-l ures. In case of major rupture in either compartment, there will be a high sump level signal. Valves in the lines passing into, out of, or through 2
the compartment can be remotely operated to shut off all sources of possible flooding.
The shut-off valves in the reactor building emergency sump suction lines are jacketed to protect the valves from missile damage and to contain leakage in the unlikely event of valve body failure.
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Men 0WFr-6.1-2 Amendment 3
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Emergency Injec tion
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Preliminary calculations inficate that the available NPSH at the decay heat removal pumps and reactor building spray pump's suction will be met with a water level at the top of the 4 ft 6 in. deep sump.
Specifically, the reactor building total pressure will exceed the sump water vapor pressure by 2.2 psi at the time recirculation starts.
This overpressure is equival-ent to 5.3 feet of water head.
To this is added 15.0 feet static head 2
below the bottom of the sump.
The friction loss through a single line at a full 4,500 gpm flow is calculated to be 4.0 feet giving an NPSH of 16.3 feet at the pumps.
Thus, a minimum NPSH of 16.3 feet is provided compared to the required NPSH of 14.75 feet for the decay heat removal pumps (Figure 6.1-3) and 15 feet for the reactor building spray pumps.
In all calculations, credit has been taken only for the reactor building pressure above the sump water vapor pressure during the time these pumps would be first operating in the recirculation mode.
The heat transfer capability of each decay heat cooler as a function of recirculated water temperature is illustrated in Figure 6.1-4.
The heat transfer capability at the saturation temperature corresponding to reac tor building pressure is in excess of the heat generation of the core.
Design data for core flooding system components are given in Table 6.1-1.
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Amendment 2 6.1-3
3 TABLE 6.1-1 CORE FLOODING SYSTEM PERFORMANCE AND EQUIPMENT DATA Component Data Core Flooding Tanks
- Number 2
Design Pressure, psig 700 Normal Pressure, psig 600 Design Temperature, F 300 Operation Temperature, F 110 3
ft / Tank 1,410 Total Volume, 3
Normal Water Volume, ft / Tank 940 2l Material of Construction Carbon Steel - s.s. Lined Check Valves Number per Flooding Line 2
Size, in.
14 Material SS Design Pressure, psig 2,500 Design Temperature, F 650 Stop Valves Number per Flooding Line 1
Size, in.
14 Material SS Design Pressure, psig 2,500 Design Temperature, F 650 Piping Number of Flooding Lines 2
Size, in.
14 Material SS Design Pressure, psig 2,500 Design Temperature, F 650
- Designed to ASME Section III, Class C.
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6.1-4 Amendment 2 L
Emerg ncy Injection 6.1.3 DESIGN EVALUATION In establishing the required components for the emergency injection, the
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following factors were considered:
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a.
The probability of a major reactor coolant system failure is very low.
b.
The fraction of a given component lifetime for which the component is unavailable because of maintenance is estimated to be a small part of lifetime.
On this basis, it is esti-mated that the probability of a major reactor coolant system accident occurring while a protective component is out for maintenance is two orders of magnitude below the low basic accident probability.
c.
The equipment downtime for maintenance in a well-operated station of ten can be scheduled during reactor shutdown periods. When maintenance of an engineered safeguard com-ponent is required during operation, the periodic test frequency of the remaining equipment can be increased to ensure availability.
d.
Where the systems are designed to operate normally or where meaningful periodic tests can be performed, there is also a low probability that the required emergency action would not be performed when needed. That is, equipment reliability is improved by using it for other than emergency functions.
,m e.
Two high pressure injection pumps and one identical makeup pump are installed.
One make-up pump is operating normally and can be taken out of service for maintenance by temporar-ily using one of the high pressure injection pumps.
Only 2
one high pressure injection (or makeup) pump is required for engineered safeguards.
6.1.3.1 Failure Analysis The single failure analysis presented in Table 6.1-2 is based on the assumption that a major loss-of-coolant accident had occurred.
It was then assu'med that an additional malfunction or failure occurred either in the process of actuating the emergency injection systems or as a secondary accident effect. All credible failures were analyzed.
For example, the analysis includes malfunctions or failures such as electrical circuit or motor failures, stuck check valves, etc.
It was considered incredible that valves would change to the opposite position by accident if they were in the required position when the accident occurred.
In general, failures of the type assumed in this analysis should be unlikely because a program of periodic testing and service rotation of standby egoipment will be incorporated in the Station operating procedures.
( n-232 Amendment 2 6.1-5
TABLE 6.1-2 SINGLE FAILURE ANALYSIS - EMERGENCY INJECTION Component Malfunction Comments and Consequences A.
liigh Pressure Injection 1.
Electric motor tsive at Valve remains open.
When the tank is empty, tank pres-makeup tank outlet.
sure would be less than the high-pressure injection pump suction pressure (with borated water stor-age tank on the line), thus pre-venting the release of hydrogen from the tank to the pump suction line.
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2.
Electric motor-operated Fails to open.
Similar valve in other high 7
suction valve for makeup pressure injection pump line li. P. injection pumps from will deliver required flow.
borated water storage tank.
3.
II. P.
injection pump.
Out for maintenance.
A high pressure and make-up pump will still be available.
Only one pump is required for engineered safeguards.
4.
H. P.
injection pump Fails (stops).
Other high pressure injection 2
pump will deliver the required flow.
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5.
it. P.
Injection pump isolation Left inadvertently closed.
See Item A-3 above.
Valves will E
valve.
normally be left open since the S
check valve in each pump dis-S.
charge will prevent backflow.
5 Operating procedures will call Y
5 for pump isolation valves to be S
i closed only for maintenance.
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$aa Component Malfunction Comments and Consequences 6.
II. P. Injection pump discharge Sticks closed.
This is considered incredible l2 check valve.
since the pump discharge pressure I
of 2,700 psig at no flow would tend to open even a very tightly stuck check disc.
7.
Pressurizer level control Fails to close.
No consequences.
l2 valve.
8.
Seal injection control valve.
Fails to close.
Injection flow through this line would be small compared to the y,
flow through the injection l2
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1, lines due to the high flow re-sistance of the reactor coolant pump seals.
9.
Electric motor-operated valve Fails to open.
Flow from one pump will go through 2
py in high-pressure injection the alternate line. Other pump
(,a
- line, will operate normally.
4 10.
Check valve in injection line Sticks closed.
See comment on Item A-6 above.
(inside reactor building).
B I11.
Injection line inside reactor Rupture Flow rate indicators in the four l2 building.
injection lines would indicate the g
gross di f ference in flow rates, q
Check valve in the injection line would prevent additional loss of
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coolant from the reactor. The m
line is protected from missiles n
by reactor coolant system shielding, o
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TABLE 6.1-2 Continued Component Ital function Comments and Consequences B.
Core Flooding System 1.
Flooding line check valve.
Sticks closed.
This is considered incredible based on the valve size and opening pressure applied.
C.
Decay lleat Removal System 1.
Check valve at reactor Sticks closed.
This is considered incredible since vessel.
these valves will be used periodically during decay heat removal, and the opening force will be approximately 5,000 pounds, cs
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2.
Electric notor-operated Fails to open.
Second injection line will deliver ao injection valve.
required flow.
g 3.
Safety valve.
Stuck open.
Loss of injection flow is small since valve is small, fj 4.
Decay heat cooler.
Isolation valve left Other heat exchanger will take j{
closed.
required injection flow and remove g
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q,a required heat.
Valves will be n
tyT closed only for maintenance of g
heat exchanger.
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Component Ma l func tion Comments and Consequences n
5.
Decay heat cooley.
Massive rupture.
Not credibic.
During normal decay heat removal operation, heat exchanger will be exposed to higher pressure and approximately the same temperature as the postaccident temperature and pressure.
6.
Decay heat cooler.
Out for maintenance.
Remaining heat exchanger will take required injection flow.
Ps) 7.
Decay heat pump isolation Left closed.
Remaining pump will deliver required l2 (s4 valve.
injection flow.
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Decay heat pump discharge Sticks closed.
See comment on Item C-1 above, j3 check valve.
9.
Decay heat pump.
Fails to start.
Remaining pump will deliver required l2 injection flow.
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10.
Electric motor-operated stop Sticks closed.
Alternate line will permit required l2 5 valve or check valve at borat-flow.
ed water storage tank outlet.
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11.
Electric motor valve per-Fails to open.
Two valves are provided; one will B"
2 mitting suction from reac-provide required flow.
These valves tor building sump.
need not be actuated until 36 minutes after start of accident which pro-vides time for manual operation.
A 1
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TABLE 6.1-2 Continued Component Malfunction Comments and Consequences 12.
Reactor building sump out-Becomes clogged.
Clogging of a single line does not let pipe.
impair function because of the dual sump line arrangement, the size of the lines, and the sump design. The two recirculation lines take suction from the different portions of the sump.
A grating will be provided over the sump, and additional heavy duty screens will be provided, 13.
Reactor building sump Valve opens with The emergency procedures will be well m
recirculation valve.
empty sump, established and rehearsed.
Therefore,
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it is not considered reasonable that the operator would inadvertantly open the valve before it is prudent.
If the valve is opened a f ter about 3-1/2 minutes following the accident, suf-ficient coolant inventory will be present in the reactor building sump to maintain a flooded suction for the 1
decay heat removal pumps. Under these g
u circumstances cooling will be provided N
in the recirculation mode, g
5 In the event that a valve in the sump g
line is opened before 3-1/2 minutes, 4
p g
the decay heat removal and reactor 2s 8,
building spray pumps in that pump train g
g would lose suction.
The pumps in the o
g alternate train would provide required n ;
- cooling, g
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No Component Ma l func tion Comments and Consequences n
H 14.
Reactor building sump Opened before borated The consequences of this operator recirculation valve, water storage tank is error would depend on the reactor
- empty, building pressure at the time the valve was opened.
It~ the building pressure were below the static pressure at the borated water storage tank, the pumps would continue to take suction from the storage tank and there would be some flow from the storage tank into the os sump.
Ilowever, since this water will 1
become available when recirculation
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begins, there are no resultant con-
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sequences.
If the building pressure is greater than the static pressure of the storage tank at the time of the valve opening, the flow of borated r\\)
water from the storage tank would be led cut off by the closing of the check valve in the borated water suction p
line.
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TABLE 6.1-2 Continued Component 11alfunction Comments and Consequences The high pressure injection pumps would continue drawing from the storage tank. The decay heat removal and reactor building spray pumps would take suction from the sump.
The hot sump water would be cooled in the decay heat removal coolers before the L. P. injection. The hot sump water would however, cause a drop in the reactor building spray cooling efficiency. The reactor l1 building emergency coolers would es L.
continue to operate at 100% capacity L
and would more than adequately l1 compensate for the loss of spray rs) cooling efficiency.
As soon as the tsa building pressure drops below the
'l) borated storage tank static pressure, the pumps would resume taking suction from the storage tank.
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15.
Dual manual valves Inadvertantly left Not credible that both valves will R
connecting suction headers.
open be inadvertantly left open because 2 x of administrative controls.
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Emergency Injection
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The single failure analysis (Table 6.1-2) and the dynamic postaccident
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performance analysis (Section 14) of the engineered safeguards considered capacity reduction as a result of equipment being out for maintenance or as a result of a failure to start or operate properly. This amounts to adding another factor of conservatism to the analyses because good operat-ing practice requires repairing equipment as quickly as possible.
Plant i
maintenance activities will be scheduled so that the required capacity of the engineered safeguards systems will always be available in the event of an accident.
l2 The adequacy of equipment sizes is demonstrated by the postaccident per-formance analysis described in Section 14, which also discusses the con-sequences of achieving less than the maximum injection flows. There is sufficient redundancy in the emergency injection systems to preclude the possibility of any single credible failure leading to core melting.
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6.1.3.2 Emergency Injection Response The emergency high-pressure injection valves are designed to open within 10 sec.
One makeup pump is normally in operation, and the pipe lines 2
are filled with coolant.
The four high-pressure injection lines contain thermal sleeves at their connections into the reactor coolant piping to prevent overstressing of the pipe juncture when 90 F water is injected into these high temperature lines during emergency operation.
The equip-fgg ment normally operating is handling 125 F water, and hence will experience I, f f
no thermal shock when 90 F water is introduced.
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Injection response of the core flooding system is dependent upon the rate of reduction of reactor coolant system pressure.
For a maximum hypothet-ical rupture, the core flooding system is capable of reflooding the core to the hot spot within a safe period af ter a rupture has occurred.
Emergency low pressure injection by the decay heat removal system will be delivered within 25 sec after the reactor coolant system reaches the actu-ating pressure of 200 psig. This anticipated delay time consists of these intervals:
a.
Total instrumentation lag -
= 1 sec b.
Emergency power source start -
< 15 sec c.
Pump motor startup (from the time the pump motor line circuit breaker closes until the pump attains full speed) -
= 10 sec d.
Injection valve opening time -
< 10,sec e.
Borated water storage tank outlet valves -
< 10 's e c Total (only b and c are additive)
= 25 sec
, (,
I Amendment 2 6.1-13
Emergency Injection 6.1.3.3 Special Features The core flooding nozzles (Figure 3.2-61) will be specially designed to ensure that they will safely take the differential temperatures imposed by the accident condition.
Special attention also will be given to the ability of the injection lines to absorb the expansion resulting from the recirculating water temperature.
For most of their routing, the emergency injection lines will be outside the reactor and steam generator shielding, and hence protected from mis-sites originating within these areas.
The portions of the injection lines located between the primary reactor shield and the reactor vessel wall are not subject to missile damage because there are no credible sources of mis-2 siles in that area. To afford further missile protection, a high-pressure injection line connects to each reactor coolant inlet line and the two core flooding nozzles are located on opposite sides of the reactor vessel.
All water used for emergency injection fluid will be maintained at a minimum concentration of 2,270 ppm of boron (13,000 ppm boric acid).
The temperature, pressure, and level of these tanks will be displayed in the control room, and alarms will sound when any condition is outside the normal limits. The water will be periodically sampled and analyzed to ensure proper boron concentration.
6.1.3.4 Check Valve Leakage - Core Flooding System The action that would be taken in the case of check valve leakage would be a function of the magnitude of the leakage.
Limited check valve leakage will have no adverse effect on reactor opera-tion.
The valves will be specified to meet the tightness requirements of MSS-SP-61, " Hydraulic Testing of Steel Valves."
- For these valves, this amounts to a maximum permissible leaLage of 140 cc/hr per valve.
Two valves in series are provided in each core flooding line; hence, leakage should be below this value.
Leakage across these check valves can have three effects:
(a) it can cause a temperature increase in the line and core flooding tank, (b) it can cause a level and resultant pressure increase in the tank, and (c) it can cause dilution of the borated water in the core flooding tank.
Leak-age at the rate mentioned above causes insignificant changes in any of these parameters.
A leakage of 140 cc/hr causes level increase in the tank of less than 1 in./mo.
The associated temperature and pressure increase is correspondingly low.
- MSS - Manufacturers' Standardization Society
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0; 241 6.1-14 Amendment 2
Emergency Injection x
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If it were assumed that the leakage rate is 100 times greater than speci-fied, then there would still be no significant effect on reactor opera-tion since the level change would be approximately 2 in./ day.
A 2-in.
level change will result in a pressure increase of approximately 10 psig.
With redundant temperature, pressure, and level indicators and alarms available to monitor the core flooding tank conditions, the most signifi-cant effect on reactor operations is expected to be a more frequent sampling of tank boric acid concentration.
To ensure that no temperature increase will occur in the tank, even at higher leakage rates, the portion of the line between the two check valves and the line to the tanks will be left uninsulated to promote convective losses to the building atmosphere.
In summary, reactor operation may continue with no adverse effects coin-cident with check valve leakage.
Maximum permissible limits on core flooding tank parameters (level, temperature, and boron concentration) will be established to ensure compliance with the core protection crite-ria and final safety analyses.
6.1.4 TEST AND INSPECTIONS All active components, as listed in Table 6.1-3, of the emergency
,-~s injection systems will be tested periodically to demonstrate system
'(
)
readiness.
In addition, normally operating components will be inspected for leaks from pump seals, valve packing, flanged joints, and safety valves.
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242
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l Amendment 1 6.1-15
Emergtncy Injection TABLE 6.1-3 EMERGENCY IlUECTION EQUIPMENT PERFORMANCE TESTING L
Equipment Test H.P. Injection pump One makeup pump is operating continuously.
3 The two H.P. Injection pumps will be periodi-cally tested.
)
High Pressure Injection The remotely operated stop valve in each Line Valves line will be opened partially one at a time. The flow devices will indicate flow through the lines.
2l H.P. Pump Suction Valves The makeup tank water level will be raised to equalize the pressure exerted by the borated water storage tank. The valves will then be opened individually and closed.
Decay Heat Pumps In addition to use for shutdown cooling, these pumps will be tested singly by opening the borated water storage tank outlet valves and the bypasses in the borated water stor-age tank fill line. This will allow water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.
i Borated Water Storage Tank The operational readiness of these valves Outlet Valves will be established in completing the pump operational test discussed above.
During this test, each of the valves will be tested separately for flow.
Low Pressure Injection With pumps shut down and borated water Valves storage tank outlet valves closed, these valves will be opened and reclosed by operator action.
Valve for Suction from With pumps shut down and borated water Sump storage tank outlet valves closed, these valves will be opened and reclosed by operator action.
(Deleted) 2 Valves in Core Flooding Valves can be operated during each shut-Injection Lines down to determine performance.
Isolation valves will be closed to contain water in core flooding tanks during shutdown.
s' 6.1-16 Amendment 3
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EMERGENC'l INJECTION SAFEGUARDS e
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m-@suun SACRAMENTO MUNICIPAL UTILITY DISTRICT
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100 200 300 400 500 600 Capacity, gpm FIGURE 6.1-2 MAKEUP PUMP CHARACTERISTICS 246
=ammmeswc t< s)SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT Amendment 2
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U 440 17
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3 240 12 3
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0 800 1600 2400 3200 4000 4500 5600 Capacity, gpm FIGURE 6.1-3 DECAY HEAT REMOVAL s
s PUMP CHARACTERISTICS 247 gggy> m SACRAMENTO MUNICIPAL UTILITY DISTRICT I.
1 Ii Emergency Design Conditions injection Water Flow Per Cooler - 3,000 gpm Nuclear Service Water Flow Per Cooler - 3,000 gpm Nuclear Service Water Inlet Temperature - 95 F 320 300 280 260 Cooler inlet Water Temperature
/
eactor Builaing 240 Surnp Ternp. )
m 220 g
200 i
/
/InjectionWater 180
/
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Temperature
"""U 160
/ /
J 140 120
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100 80 0
20 40 60 80 100 120 140 160 180 Heat Transferred. Btu /hr x 10-6 FIGURE 6.1-4 DECAY 11 EAT RDiOVAL COOLER CllARACTERISTICS
$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT AMENDMENT 3
6.2 REACTOR BUILDING ATMOSPHERE COOLING AND WASHING 7m x_/
6.2.1 DESIGN BASES Emergency building atmosphere cooling and washing is provided to reduce the postaccident level of fission products in the building atmosphere.
Spray 2
additives contained in the building sprays are used to reduce postaccident fission product concentrations in the building atmosphere.
6.
2.2 DESCRIPTION
The schematic flow diagram of the reactor building emergency atmosphere cooling and washing and associated instrumentation is given in Figure 6.2-1.
Four emergency cooling units, each of 25% capacity will be provided.
Each unit contains a single coil sized for emergency service, and a direct-driven 2
fan.
For emergency cooling, all units will operate under post-accident conditions with the heat being rejected to the nuclear service cooling water D
system.
Each of these units can remove 60 x 10 Btu per hour under peak reactor building temperature conditions.
The design data for the cooling units are shown in Table 6.2-1.
The reactor building sprays are supplied with water by two pumps which 1
take suction on the borated water storage tank until this source is exhausted.
The spray additive required for the reactor building sprays is supplied from a storage tank conr.ected by electric motor operated stop check valves, inter-p,_
i
(
locked to open only when the reactor building spray pump motors are operating.
l2 V
TABLE 6.2-1 REACTOR BUILDING EMERGENCY COOLING UNIT PERFORMANCE AND EQUIPMENT DATA (capacities are on a per unit basis)
Equipment Data Duty No. of Units Installed 4
No. Required 4
Type Coil Finned Tube Peak Heat Load, Btu /hr 60 x 106 2
Fan Capacity, cfm 40,000 Reactor Building Atmosphere Inlet Conditions Temperature, F 286 Steam Partial Pressure, psia
~ 53 Air Partial Pressure, psia
~ 21 Total Pressure, psig 59 Cooling Water Flow, gpm 1,500 14 Cooling Water Inlet Temperature, F 95 l2
( [
Cooling Water Outlet Temperature, F 175 l4 a
x 240
/
l Amendment 4 6.2-1
l Recctor Building Atmosphere Cooling cnd Washing Two sets of high efficiency and charcoal filters each with 54,500 cfm
~ }
capacity are provided for normal containment air cleanup and can be used for removal of fission products in the containment volume.
2l Sufficient spray additive is injected into the borated water to create an appropriate concentration in the reactor building water inventory.
After the supply from the borated water storage tank is exhausted, the spray pumps take suction from the reactor building sump recirculation lines.
2 This continued spraying serves to equalize the reactor building atmosphere and the reactor building sump temperature.
Design data for the reactor building cooling and spray system components are given in Tables 6.2-1 and 6.2-2.
Design data for components of the de cay heat removal systems used in this phase of engineered safeguards operation are given in Section 9 and supplemented by Figures 6.1-3 and 6.1-4 of this section.
6.2.3 DESIGN EVALUATION The function of cooling the reactor building atmosphere, after safety injection, is fulfilled by either of the two methods described above, and redundancy of equipment within both methods will provide for protection of building integrity. The reactor building sprays, through duplication, 2l basic washing concept, and spray additive will serve to reduce fission product levels in the building atmosphere.
TABLE 6.2-2 REACTOR BUILDING SPRAY SYSTEM PERFORMANCE AND EQUIPMENT DATA (capacities are on a per unit basis)
Component Data Reactor Building Spray Pumps Number 2
~
Flow, gpm 1,500 Developed Head at Rated Flow, ft 430 Motor Horsepower, hp 250 Material SS Design Pressure, psi 300 Design Temperature, F 300 2l Spray Additive Tank Number i
Volume, ft3 1,500 Material SS Design Pressure, psi 50 2l Design Temperature, F 150 Spray Header Number 2
Spray Nozzles per Header 375 J
~
250 l
6.2-2 Amendmtnt 2
1 Reactor Building Atmosphere Cooling and Washing For the first 30 to 40 minutes following the maximum blowdown loss-of-( ]*
coolant accident, i.e., during the time that the reactor building spray pumps take their suction from the borated water storage tank, this system provides more than 100 percent of the heat removal capacity of the reactor building cooling system.
The reactor building spray system design is based on the process of the spray water being raised to the temperature of the reactor building while falling through the steam-air mixture within the building.
Detailed eval-uation of system performance is presented in Section 14.
Each of the fol-lowing equipment arrangements will provide sufficient heat removal capa-bility to maintain the postaccident reactor building temperature to that which may be required for fission product removal.
a.
Reactor building spray system.
b.
All emergency units in the reactor building cooling system.
c.
Two emergency cooling units and the reactor building spray system at one-half capacity.
The reactor building spray system shares the suction lines from the borated water storage tank and the tank itself with the high and low pressure injection systems.
(n) 6.2.3.1 Failure Analysis v
A single failure analysis has been made on all active components of the systems used to show that the failure of any single active component will not prevent the fulfilling of the design functions. This analysis is shown in Table 6.2-3.
Assumptions inherent in this analysis are the same as those presented in 6.1.3 in regard to valve functioning, failure types, etc.
Results of full and partial performance of these safeguards are presented in Section 14 under analysis of postaccident conditions.
6.2.3.2 Reactor Building Cooling Response Air recirculation is established within 35 seconds af ter the building pres-sure reaches 4 psig by the four reactor building emergency cooling units.
Cooling coils in these ventilation units are supplied with cooling water from the nuclear service cooling water system after reactor building pres-l2 sure increases to 4 psig.
This mode of cooling continues until the building pressure reaches near-atmospher h.
The reactor building spray system will also be activated by a single para-meter signal. Two of three signals signifying high reactor building pres-sure will start the reactor building spray pumps, open the reactor building spray inlet valves, and open the suction valves from the borated water stor-age tank and the spray additive valves.
The system components may also l2
( ls
)
be actuated by operator action from the control room for performance testing.
s 7
251 Amendment 2 6.2-3
TABLE 6.2-3 SINGLE FAILURE ANALYSIS - REACTOR BUILDING ATMOSPHERE COOLING AND WASHING Component Malfunction Comments and Consequences 1.
Reactor Building spray nozzles.
Clogged.
Large number of nozzles (375 on each of two headers) renders clogging of significant number of nozzles as incredible.
2.
Reactor Building spray header.
Rupture.
This is considered incredible due to low operating pressure differential.
2 3.
Check valve in spray header line.
Sticks closed.
Second header and emergency cooling units will provide 150 percent design cooling capacity.
This is considered incredible due to large opening force available at
'f pump shutoff head.
v-2 N 4.
Electric motor operated valve Fails to open.
Second header and Reactor Building emer-w in spray header line.
gency cooling units will provide 150 per-N ce..
of design base cooling requirement.
- =
5.
Spray pump isolation valve.
Left closed.
Flow and cooling capacity reduced to 50 9$
percent of design.
In combination with S
emergency coolers, 150 percent of total E8 design requirement is still provided.
" en
$5 "E
6.
Reactor Building spray pump.
Fails to start.
Flow and coaling capacity reduced to 50 2
percent of design.
In combination with fE emergency coolers, 150 percent of total E"
design requirement is still provided.
EN oo s O
o Reactor Building emergency Fails to start.
Emergency cooling by the other operat-E g-2 g
cooling unit.
ing units with supplemental cooling by the sprays, 175 percent of total design m
(
requirement is still provided.
O O
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f x.s u-g TABLE 6.2-3 Continued U
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Component Halfunction Comments and Consequences N
8.
Reactor Building emergency Rupture of cooling The tubes are designed for 200 psi and l2 cooling unit.
coil.
300 F which exceeds maximum operating I conditions. Tubes are protected against credible missiles, llence,
rupture is not considered credibic.
9.
Reactor Building emergency Rupture of casing Consideration will be given during 2
cooling unit.
and/or ducts.
detailed design to the dynamic forces c
resulting from the pressure buildup during a postaccident situation. The units can be inspected and will be protected against credible missiles.
Cooling with these units will be m
supplemented by the sprays.
w vi a
10.
Reactor Building emergency Rupture of system Rupture is not considered credible l2 cooling units.
piping.
since all piping is Schedule 40, per-mitting an allowable working pressure 111.
of at least 500 psig at 650 F for all s iz es.
Piping is inspectable and prc-nE tected from missiles. Maximum actual gg internal pressure will be less than
-n 200 psig at temperatures below 300 F. l2 70
" en
$E Electric motor-operated valve Sticks closed.
Should a valve fail to open, the l2 ){
at inlet penetration to remaining cooling units plus the 1
emergency cooling unit.
reactor building spray system will gg g
supply 1757. design cooling ::apacity.
g, O
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- 12. Electric motor-operated valve Fails to open.
Comments for Item 11 apply.
E at outlet penetration from 2
I emergency cooling unit.
2
+
TABLE 6.2-3 Continued
- . ~. -
Component Malfunction Comments and Consequences 13.
Check valve at spray additive Fails to open.
Alternate check valve will permit 2
storage tank outlet.
full flow required for sprays.
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Racetor Building Atmosphere Cooling and Washing The expected lag time for the foregoing operations is the same as for the g
low pressure injection (6.1.3.2), and the total time to delivery of reactor building sprays is approximately 25 seconds after building pressure reaches 10 psi.
6.2.3.3 Special Features The casing design for the ventilation units will be of a conventional nature unless additional analysis shows the possibility of pressure wave collapse.
In that event, quick, inward-opening hinged louvers, or other protective features, will be incorporated into the design to maintain post-accident operability. The ventilation units are located outside the concrete shield for the reactor vessel, steam generators, and reactor coolant pumps at an elevation above the water level in the bottom of the reactor butiding at post accident conditions.
In this location, the systems in the reactor building are protected from credible missiles and from flooding during post-accident operations.
~
The spray headers of the reactor building spray system are located outside and above the reactor and steam generator concrete shield.
During opera-tion, a shield also provides missile protection for the area immediately above the reactor vessel. The spray headers are therefore protected from missiles originating within the shield. The spray pumps are located outside the reactor building and are thus available for operative checks during station operation.
6.2.4 TESTS AND INSPECTIONS Active components of the emergency cooling units will be periodically 2
tested.
Components of the reactor building spray system will be tested on a regu-lar schedule as follows:
Reactor Building These pumps will be tested singly by closing the spray Spray Pumps header valves and opening the valves in the test line and the borated water storage tank outlet valves.
Each pump in turn will be started by station operator s
action and checked for flow establishment to each of the spray headers. Flow will also be tested through each of the borated water storage tank outlet valves by operating these valves.
Borated Water These valves will be tested in performing the pump Storage Tank test listed above.
Outlet-Valved Reactor Building With the pumps shut down and the borated -water stor-Spray Injection age tank outlet valves closed, these valves will each C
Ma,1ves
.be opened and closed by operator action.
lw.
O
Reactor Building Atmosphere Cooling and Washing Reactor Building Under the conditions specified for the previous test,
T, Spray Nozzles and with the reactor building spray valves alternately open, smoke will be blown through the test connections.
Spray Additive These valves will each be opened and closed by operator Valve Test action from the control room.
During the test the normally-locked open valve will be closed.
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ggg REACTOR BUILDING LJV EMERGENCY COOLING osuuo SACRAMENTO MUNICIPAL UTILITY DISTRICT
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s Amendment 2
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6.3 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS 6.
3.1 INTRODUCTION
The use of normally-operating equipment for engineered safeguards' functions and the location of some of this equipment outside the reactor building requires that consideration be given to direct radiation levels af ter 5
fission products have accumulated in these systems with leakage from these systems. Although the engineered safeguards ' equipment is designed for control room operation following an accident, long-term postaccident operation could necessitate manual operation of certain valves.
The shielding for components of the engineered safeguards is designed to provide protection for personnel while performing all operations necessary for mitigation of the accident without exceeding acceptable dose limits.
6.3.2 St&&1ARY OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATION Following a loss-of-coolant accident and exhaustion of the borated water storage tank, reactor building sump recirculation to the reactor vessel and the reactor building sprays is initiated.
While the reactor auxiliary systems involved in the recirculation complex are closed to the auxiliary building atmosphere, leakage is possible
,. (' ',
through component flanges, seals, instrumentation, and valves.
( kj The Icakage sources considered are:
a.
Valves (1) Disc leakage when valve is on recirculation complex boundary.
(2)
Stem leakage.
(3)
Flanges c.
Pump stuffing boxes While leakage rates have been assumed for these sources, maintenance and periodic testing of these systems will preclude all but a small percentage of the assumed amounts. With the exception of the boundary valve discs, all.of the potential leakage paths may be examined during periodic test or normal operation. The boundary valve disc leakage is retained in the other closed systems and therefore, will not be released to the auxiliary building.
y x:
259 r_,/
6.3-1
Enginnered Safeguards L:ckcga and Radiation Considerations While valve stem leakage has been assumed for all valves, the manual valves
~x in the recirculation complex are backseating.
6.3.3 LEAKAGE ASSUMPTIONS Source Quantities a.
Valves - Process (1)
Disc leakage 10 cc/hr/in. or nominal disc diameter (2) Stem leakage 1 drop / min (3) Bonnet Flange 10 drops / min b.
Valves - Instrumentation Bonnet flange and stem 1 drop / min c.
Flanges 10 drops / min d.
Pump Stuffing Boxes 50 drops / min For the analysis, it was assumed that the water leaving the reactor building was at 281 F.
This assumption is conservative since this peak j
temperature would only exist for a short period during the postaccident condition. Water downstream of the coolers was assumed to be 115 F.
The auxiliary building was assumed to be at 70 F and 30 percent relative humid t:y. Under these conditions, approximately 22 percent of the leakage upstream of the coolers and 4 percent of the leakage downstream of the coolers would flash into vapor.
For the analysis, however, it was assumed that 50 percent of the leakage upstream of the coolers would become vapor because of additional heat transfer from the hot metal.
6.3.4 DESIGN BASIS LEAKAGE The design basis leakage quantities derived from these assumptions for postaccident sump recirculation are tabulated in Table 6.3-1.
6.3.5 LEAKAGE ANALYSIS CONCLUSIONS It may be concluded from this analysis (in conjunction with the discussion and analysis in Section 14) that leakage from engineered safeguards' equip-ment outside the reactor bu'ilding does not pose a public~ safety problem.
I h
260 N
L?
6.3-2
Engineered Safeguards Leakage and Radiation Considerations 3
TABLE 6.3-1 A
)
LEAKAGE QUANTITIES TO AUXILIARY BUILDING ATMOSPHERE v
Liquid Vapor No. of Per Source,
- Total, Phase,
- Phase, Leakage Source Sources drops / min ec/hr cc/hr cc/hr a.
Pump Seals Decay heat pumps 2
50 300 150 150 l2 Spray pumps 2
50 300 150 150 b.
- 114 10 3,320 1,800 1,520 c.
Process Valves 35 1
105 68 37 d.
Instrumentation Valves 25 1
75 72 3
e.
Valve Seats at Boundaries 11 750 580 170 Total 4,850 2,820 2,030 l2
- Assumes process and boundary valves, and process components are flanged.
- Assumes 10 cc/hr/in. of nominal disc diameter.
261 (lp) us i
Amendment 2 6.3-3