ML19322A788
| ML19322A788 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/01/1966 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 7911250006 | |
| Download: ML19322A788 (62) | |
Text
{{#Wiki_filter:_x ,, _ _ , - _ _ _ _ . _ - _ _ _ - - - . . - W ' m l C,1 l l
;;' .l ,
E' . . 4 J , l l I i l O 33 O _. 7911 250 %
. - -- _ .- . . _. - - - . _ . _ ___.______._____.__l
TABLE OF CONTENTS Section g 6 ENGINEERED SAFEGUARDS 6-1 6.1 EMERGENCY INJECTION SYSTEMS 6-2 6.1.1 DESIGN BASES 6-2 6.1.2 SYSTEM DESIGN 6-2 6.1.2.1 Piping and Instrumentation Diagram 6-2 6.1.2.2 Codes and Standards 6-4 6.1.2.3 Material compatibility 6-4 6.1.2.4 Component Design 6-5 1 Coolant Storage 6-5 6.1.2.5 6.1.2.6 Pump Characteristics 6-5 a 6.1.2.7 Heat Exchanger Chacacteristics 6-5 6.1.2.8 Relief Valve Settings 6-6 6.1.2.9 Reliability Considerations 6-6 1 6.1.2.10 Missile Protection 6-7 l l 6.1.2.11 Actuation 6-7 ! l 6.1.2.12 Environmental Considerations 6-12 l 6.1.2.13 Performance Testing 6-12 6.1.3 DESIGN EVALUATION 6-12 6.1.3.1 High Pressure Injection System 6 13 6.1.3.2 Low Pressure Injection System 6 53 6.1.4 TESTS AND INSPECTIONS 6-14 j 6.2 REACTOR BUILDING SPRAY SYSTEM 6-16 6.2.1 DESIGN BASES 6-16 6.2.2 SYSTEM DESIGN 6-16 I 6.2.2.1 Piping and Instrumentation Diagram 6-16 l l 6-1 (Revised 4-1-67) ---.. 34 L l
b CONTENTS (Cont'd) Section g 6.2.2.2 Codes and Standards 6-16 6.2.2.3 Material compatibility 6-16 6.2.2.4 Component Design 6-16 6.2.2.5 Coolant Storage 6-16 6.2.2.6 Pump Characteristics 6-17 6.2.2.7 Heat Exchanger Characteristics 6-17 6.2.2.8 Relief Valve Settings 6-17 6.2.2.9 Reliability Considerations 6-17 6.2.2.10 Missile Protection 6-17 6.2.2.11 Actuation 6-11 6.2.2.12 Environmental Considerations 6-17 O~ 6.2.2.13 Performance Testing 6-19 6.2.3 DESIGN EVALUATION 6-19 6.2.4 TESTS AND INSPECTIONS 6-20 6.3 REACTOR BUILDING EMERGENCY COOLING UNITS 6-20 6.3.1 DESIGN BASES 6-20
- 6.3.2 SYSTEh DESIGN 6-20 6.3.2.1 Piping and Instrumentation Diagram 6-20 6.3.2.2 Codes and Standards 6-20 6.3.2.3 Materials 6-20 J 6.3.2.4 Component Design 6-20 6.3.2.5 Reactor Building Emergency Circulating Fan
! Characteristics 6-21 6.3.2.6 Reactor Building Emergency Cooler Characteristics 6-21 6.3.2.7 Reliability Considerations 6-21 0 -_ 6-11 (Revised 4-1-67) ' ' - - - 1
4 CONTENTS (Cont'd) Section g 6.3.2.8 Missile Protection 6-21 6.3.2.9 Actuation 6-24 6.3.2.10 Environmental Considerations 6-24 6.3.2.11 Performance Testing 6-24 6.3.3 DESIGN EVALUATION 6-24 6.3.4 TESTS AND INSPECTION 6-24 6.4 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS 6-24 6.
4.1 INTRODUCTION
6-24 6.4.2
SUMMARY
OF POSTACCIDENT RECIRCULATION 6-25 6.4.3 BASES OF LEAKAGE ESTIMATE 5-25 () 6.4.4 6.4.5 LEAKAGE ASSUMPTIONS DESIGN BASIS LEAKAGE 6-26 6-26 6.4.6 LEAKAGE ANALYSIS CONCLUSIONS 6-26 6.5 REACIOR BUILDING PENETRATION ROOM VENTILATION SYSTEM 6-27 6.5.1 DESIGN BASES 6-27 6.5.2 SYSTEM DESIGN 6-27 6.5.2.1 Piping and Instrumentation Diagram 6-27 6.5.2.2 Codes and Standards 6-28 6.5.2.3 interial Compatibility 6-28 6.5.2 4 Equipment Accessibility 6-29 6.5.2.5 Reliability Considerations 6-29
/
6.5.3 DESIGN EVALUATION 6-29
6.6 REFERENCES
,' . 6-30 0 _
36 6-iii (Revised 4-1-67) --
O LIST OF TABLES Table g 6-1 Single Failure Analysis Emergency Injection Systems 6-8 6-2 Emergency Injection Systems Performance Testing 6-15 6-3 Single Failure Analysis Reactor Building Spray System 6-18 6-4 Single Failure Analysis for Reactor Building Emergency Cooling Units 6-22 6-5 Leakage Quantities To Auxiliary Building Atmosphere 6-27 6-6 Single Failure Analysis for Reactor Building Penetra-tion Room Ventilation System 6-29 O O - 6-iv (Revised 4-1-67)
LIST OF FICURES Figure 6-1 Simpliff ed Schematic r'2 gram of Engineered Safeguard System for , Core and Building Protection 6-2 Emergency 1 ceion Systems 6-3 High Pressure Injection Pump Characteristics 6-4 Low Pressure Injection Pump Characteristics 6-5 Low Pressure Injection Cooler Capacity - 6-6 Reactor Building Spray System 6-7 Reactor Building Spray Pump Characteristics 6-8 (Deleted by 4-1-67 Amendment) 6-9 Reactor Building Emergency Cooling Units 6-10 Reactor Building Emergency Cooler 6-11 Reactor Building Emergency Cooler Heat Removal Capacity 6-12 Reactor Building Postaccident Steam-Air Mixture Composition 6-13 (Deleted by 4-1-67 Amendment) 6-14 (Deleted by 4-1-67 Amendment) 6-15 Reactor Building Penetration Room Ventilation System I
; 1
$ 1 6-v (Revised 4-1-67) _
) I LIST OF FIGURES (At Rear of Section) Figure No. Title 6-1 Simplified Schematic Diagram of Engineered Safeguard System for Core and Building Protection 6-2 Emergency Injection Systems 6-3 High Pressure Injection Pump Characteristics 6-4 Low Pressure Injection Pump Characteristics 6-5 Low Pressure Injection Cooler Capacity 6-6 Reactor Building Spray System 6-7 Reactor Building Spray Pump Characteristics > 6-8 Reactor Building Spray Coolez Capacity 6-9 Postaccident Operating Mode for Reactor Building and Component Cooling System 6-10 Reactor Building E=ergency Coolers 6-11 Reactor Building Emergency Cooler Heat Removal Capacity 6-12 Reactor Building Postaccident Steam-Air Mixture Composition 6-13 Reactor Building Component Cooling Pump daracteristics 6-14 Reactor Building Component Cooler Heat Re= oval Capacity 6-15 Reactor Building Penetration Room Ventilation System i e V __ 39 O-vi ... -
x 6 ENGINEERED SAFEGUARDS Engineered safeguards systems in the Ocenee Nuclear Station will fulfill three functions in the unlikely event of a serious loss-of-coolant accident:
- a. Protect the fuel cladding.
- b. Insure Reactor Building integrity,
- c. Prevent uncontrolled leakage of reactor building contents to the atmosphere.
Separate equipment is used to perform each of these functions. All of the equipment requiring electrical power vill be supplied by the emergency elec-trical power sources (8.2.3). The design of these systems results in maximum postaccident use of equipment serving normal functions. This is based on the belief that regular use of equipment provides the best possible means for monitoring its availability. The engineered safeguards systems include the high pressure injection system, the low pressure injection system, the core flooding system, the Reactor Building emergency cooling units, the Reactor Building spray system and the penetration room ventilation system. Figures 6-1, 6-2, 6-6, 6-9 and 6-15 show the engineered safeguards systems. The high and low pressure injection systems and the core flooding system 'are designed to protect the reactor fuel cladding from damage. The high pres-sure injection system is arranged so that three pumps are available to each unit. The low pressure injection system also provides three pumps to each unit. The core flooding system for each unit provides two separate tanks containing borated water at ambient conditions. The tanks are pressurized with nitrogen to provide the required passive injection force. Figure 6-2 shows the injection systems. Reactor Building integrity is insured by two full-capacity independent cooling systems, ie, the Reactor Building spray system and the Reactor Building emergency cooling units. Figures 6-6 and 6-9 show these two systems. Postaccident leakage from the Reactor Building penetrations is collected and filtered. Separate, redundant equipment is provided for each unit. Figure j 6-15 shows this arrangement. i Because much of the equipment in these systems serves a normal function, the need for periodic testing is minimized. In those cases where the equipment 1 is used for emergencies only, the systems have been designed to permit mean-ingful periodic tests. The design bases, a description, a schematic drawing and a design evaluation for each of these systems are presented in the fol-lowing sections. A significant part of the design evaluation is the failure analysis which presents the comments and consequences pertinent to each assumed failure.
\
The design of the in;ection and flooding systems has been based upon [O . .-
.- 6-1 (Revised 4-1-67) 40
. . ...~
+t r v "r --
preventing clad damage as a result of loss-of-coolant accidents. The failure analysis has been performed to illustrate the injection flow available under each assumed failure. The analysis in Section 14 shows the adequacy of the installed equipment, including the assumed failures , to prevent clad damage. All of the engineered safeguards systems are designed to moet the seismic requirements in 2.6. Applicable codes and standards for the engineered safe-guards systems are listed in the introduction in Section 9. Additional descriptive information on certain systems and their normal opera-tion is presented in Section 9. 6.1 EMERGENCY INJECTION SYSTEMS 6.1.1 DESIGN BASES The principal design basis for the emergency injection systems is as follows: Emergency core injection systems are provided to prevent clad and fuel damage that would interfere with continued core cooling and to limit clad-water reaction to less than approximately 1 per cent for the entire spectrum of reactor coolant system failures rang!ng' from the smallest leak to the complete severance of the largest reactor coolant pipe. A high pressure injection system is provided and is designed to prevent un-covering the core for small leaks and to delay uncovering the core for inter-mediate size leaks. Low pressure injection and core flooding systems are provided and designed to recover the core and maintain core integrity for any break size up to a double ended 36" reactor outlet line failure. The injection systems are designed with sufficient redundancy to perform their intended function with a single active component out of service either for maintenance or due to a failure at the time of the accident. 6.1.2 SYSTEM DESIGN 6.1.2.1 Piping and Instrumentation Diagram The piping and instrumentation diagram for the emergency injection systems is shown as Figure 6-2. On the diagram, Unit 1 is shown in the emergency injec-tion operating mode. Unit 2 is not shown, but utilizes an identical arrange-ment. In addition to the core flooding tanks, the high and low pressure injection systems have the following total flow capability available to each unit: l 9,000 gpm Low pressure injection 1,500 gpm High pressure injection 10,500 gpm Total
.~,
41 6-2 (Revised 4-1-67) - l
6.1.2.1.1 High Pressure Injection During normal operation, one high pressure injection pump is supplying seal water to the reactor coolant pumps and returning letdown flow to the reactor coolant system. The pump, at this time, is taking suction from the letdown storage tank (Figure 9-2). A low reactor pressure signal automatically actuates the high pressure injection system with resultant actions as follows: (1) the two standby high pressure injection pumps start and come on the line, (2) ene stop valve in each high pressure injection line opens, and (3) the two parallel stop valves in the suction line from the borated water storage tank open. These pumps will continue to add water to the reactor coolant system until the borated water storage tank is empty. 6.1.2.1.2 Low Pressure Injection Each unit has an independent low pressure injection and decay heat removal system composed of three pumps, two heat exchangers and the required valves and piping. Upon actuation, the three pumps will deliver 9000 gpm of borated water to the reactor vessel when the vessel is at 100 psig. Separate flow paths are provided through two lines from the cooler discharge through the Reactor Building to independent nozzles on opposite sides of the reactor vessel. Upon the initiating-low-reactor-coolant-system-pressure signal, all three of the low pressure injection pumps for that unit will start. These pumps initially take suction from the borated water storage tank supplied with each unit. The tank for each unit provides a reservoir of 350,000 gal of about 90 F water borated to 13,000 ppm boric acid (2270 ppm boron). The low pressure injection pumps, in combination with other systems, will pump from this tank for about 25-40 minutes depending upon the number of pumps in opers. tion. At the end of this period the valve controlling suction from the Reactor Building sump is opened by the operator on low borated water storage tank level to permit the recirculation of spilled reactor coolant and injec-tion water from the Reactor Building. The outlet valves from the borated water storage tank are then closed. At this time, the decay heat being generated in the core is a maximum of 1.8 per cent of full power, or 160 x 106 btu /hr. One low pressure injection cooler will satisfy this duty. 6.1.2.1.3 Core Flooding System The core flooding system is composed of two flooding tanks, each connected to a reactor vessel injection nozzle by a line containing two isolation check valves and a normally open valve. The systen 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 insure the , supply of emergency coolant to the reactor vessel. Operator action is re-quired only during normal reactor cool down, at which time the isolation valves in the flood lines are closed to isolate the contents of the flood tanks. The combined coolant content-of the two flood tanks is sufficient to recover the core hot spot assuming no liquid remains in the reactor vessel. The gas overpressure, gas volume and flooding line sizes are sufficient to 6-3 (Revised 6-16-67)
- 42
accomplish this within approximately 25 seconds af ter a double ended failure O of a 36 in, reactor outlet line. Design data on major system components are as follows: Core Flooding Tanks Number 2 Design Pressure, psig 700 Normal Pressure, psig 600 Design Temperature, F 300 Operation Temperature , F 110 Total Volume, Ft 3 1410 Normal Water Volume, Ft3 940 Material of Construction Carbon Steel Lined with SS Check Valves Number Per Flood Line 2 Size, in. 14 Material SS Design Pressure, psig 2500 Design Temperature, F 650 Isolation Valves Number Per Flood Line 1 Size, in. 14 Material SS Design Pressure, psig 2500 Design Temperature, F 650 Piping Number of Core Flooding Lines 2 Size, in. 14 Material SS Design Pressure, psig 2500 Design Temperature, F 650 6.1.2.2 Codes and Standards The emergency injection systems are designed and will be manufactured to the applicable codes and standards listed in tha introduction to Section 9. 6.1.2.3 Material Compatibility The high pressure injection system, which operates continuously with reactor coolant, is constructed entirely of stainless steel. The low pressure injection system is composed of both aluminum and stainless steel components. The borated water storage tank is aluminum and the remaining components of the system including heat exchangers , piping, , valves, etc, are stainless steel. The piping will be welded except at major components where maintenance requirements dictate flanged connec-tions. 43
. 6-4 (Revised 6-16-67)
b 6.1.2.4 Component Design . The high pressure injection lines are designed for 2500 psie and 650 F from the connection to the reactor coolant inlet piping to the remotely operated valves outside the Reactor Building. From the remotely operated valves to the high pressure injection pumps, the lines are designed for 2850 psig (maximum possible pump discharge pressure at zero flow) and 200 F. With the exception of the borated water storage tank and the high pressure injection lines (2500 psig) on the reactor side of the isolation valves, the low pressure injection system will be designed for 300 psig at 300 F. The design pressure is determined by the alternate duty of portions of this system for decay heat removal on shutdown. The design pressure was set by adding the maximum pump discharge head to the required reactor coolant system pressure at the start of decay heat cooling (250 F). The temperature was set by adding approximately 20 F safety factor to the maximum temperature water pumped during the accident. The borated water storage tanks will be designed for the hydrostatic head of the tank and a temperature of 150 F. The high pressure injection lines will be designed for reactor coolant conditions. The physical arrangement of the coolers will be type BFM as described by the standards of the Tubular Exchanger Manufacturers Association, and the mechani-cal construction will be Class "C" as described in these standards. In addi-tion, the tubes will be seal-welded to the tube sheets. / The pumps will be horizontal centrifugal type, employing double shaft seals b to minimize leakage. The pump motors will use Class B, F or H insulation.' Time to attain full speed will be less than 15 seconds. 6.1.2.5 Coolant Storage i The letdown storage tank has a useful coolant volume of approximately 2200 gallons above the low level alarm. This tank provides water to the pumps until the borated water storage tank outlet valves are opened. Two 350,000 gallon borated water storage tanks are provided, one for each reactor. The tanks are provided with a heating system to maintain the temper-ature at about 90 F. Tae contents of the tank are circulated through the purification equipment for the spent fuel storage pool to demineralize and filter the water after it has been in the fuel. transfer canal. Provisions are also made for adding concentrated boron solution or demineralized water. 6.1.2.6 Pump Characteristics Curves of total dynamic head, NPSH and BHP versus flow are shown on Figure 6-3 for the high pressure pumps and on Figure 6-4 for the low pressure pumps. 6.1.2.7 Heat Exchanger Characteristics The design basis for the low pressure injection coolers is 'that each cooler be capable of removing the decay heat being generated at the start of recir-s culation6 f spilled injection water. For each reactor, this generation is T 160 x 10 btu /hr. O _.
\
44 6-5 (Revised 4-1-67) - --
The heat transfer capability of the low pressure injection cooler as a func-tion of recirculated water temperature is illustrated in Figure 6-5. The heat transfer capability at the satencion temperature corresponding to Reactor Building pressure is in excess of the design basis. 6.1.2.8 Relie f Valve Settings Relief valves are provided to protect the low pressure piping and components from overpressure. Thase relief valves will be set at the desi8n pressure for the systeu, ie, 300 psig. 6.1.2.9 Reliability Considerations The only portion of the high pressure injection system located inside the Reactor Building is the two separate injection lines. Each line is sized for full capacity, and each line contains a remotely operated stop valve and flow rate indicator located outside the Reactor Building. These give the control room operator the means to determine whether flow through one of the lines is not reaching the reactor vessel, and the means to stop the flow through that line. All of the high pressure injection pumps are available for normal operation to supply makeup and seal water to the reactor coolant pumps; one pump will normally be in operation and two pumps will be on standby. Duplicate full capacity valves in parallel are installed in the pump suction line from the borated water storage tank. It is anticipated that only one of the three pumps available to a unit will be secured at a given time. For the icw pressure injection system, 9000 gpm of injection capacity is pro-vided for either reactor to insure the required delivery to the reactor core. The adequacy of equipment sizes is demonstrated by the postaccident perform-ance analysis described in Section 14 which also discusses the consequences of achieving less than the maximum injection flow. There is sufficient redun-dancy in the emergency injection systems so that no credible single failure can lead to significant core damage. This is demonstrated clearly by the single failure analysis presented in Table 6-1. This analysis was based on the assumption that a major loss-of-coolant accident had occurred. It was then assumed that either in the process of actuating the emergency injection systems, or as a secondary accident effect, an additional malfunction or fail-ure occurred. For example, the analysis included 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 strict program of periedic testing and service rotation of standby equipment will be incorporated in the station operating procedures. The single failure analysis (Table 6-1) and the dynamic postaccident perform-ance 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 operating practice 6-6 (Revised 4-1-67) .- .
fh
's requires repairing equipment as quickly as possible. Station 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.
6.1.2.10 Missile Protection The portion of the high pressure injection system within the Reactor Building consists of two 4 in., Schedule 120, stainless steel injection lines. These lines are connected to the reactor coolant inlet piping on opposite sides of the reactor vessel. The portion of the low pressure injection system within the Reactor Building consists of two 14 in., Schedule 160, stainless steel injection lines and two 18 in. , Schedule 40 suction lines. The injection lines are connected to the reactor injection nozzles which are located on opposite sides of the reactor vessel. The nozzles are 2 ft-0 in, above the elevation of the reactor inlet and outlet lines. All four injection lines will penetrate the Reactor Building adjacent to the Auxiliary Building. For most of the routing, these lines will be outaide the reactor and steam generator shielding, and hence protected from missiles originating within these areas. The portions of the injection lines located between the pri-mary reactor shield and the reactor vessel wall are not subject to missile O damage because there are no credible sources of missiles in this area. 6.1.2.11 Actuation The high pressure injection system is actuated automatically by a low reactor coolant system pressure of 1800 psig. All of the pumps and valves can also be operated remotely from the control room. The low pressure injection aystem is automatically actuated whenever the reactor coolant system pressure is less than 200 psig. The active components of the system, ie, the pumps and automatic valves, will also be capable of being started by operator action for performance testing. 4 v . 46 6-7 (Revised 4-1-67)
i l Table 6-1 Single Failure Analysis Emergency Injection Systems i Component Ib1 function Comments and Consequences A. High Pressure Injection System
- 1. Pneumatic valve at letdown Operator fails to close When the tank is empty, tank pressure storage tank outlet valve would be less than the high pressure injection pump suction pressure (with borated water storage tank on the line),
thus preventing the release of hydrogen from the tank to the pump suction line.
- 2. Pneumatic suction valve Fails to open No consequences since therc are two for high pressure injec- full capacity valves arranged in
& tion pumps from borated parallel.
,s water storage tank E '
$, 3. High pressure injection Out for maintenance Two backup pumps are available.
y pump o. p 4. High pressure injection Fails (stops). Two backup pumps are available. 7 pump (operating pump). O
5. High pressure injection Left inadvertently See previous comment. Valves will pump manual isolation closed normally be left open as the check valves valve in each pump discharge will pre-vent backflow. Operating procedures will call for pump isolation valves to be closed only for maintenance.
- 6. High pressure injection Sticks closed This is considered incredible since pump discharge check valve the punp discharge pressure of 2700 psi at no flow would tend to open eve- a
. very tightly stuck check disc.
N I O O ' O
O D Table 6-1 (Cont'd) Component Halfunction Consnents and Consequences
- 7. Pressurizer level con- Fail to close No consequences. It is closed only trol valves to give the operator the best flow comparison between the two injection lines.
- 8. Reactor coolant pump Fail to close Injection flow through this line would seal pressure control be small compared to the flow through valves the two injection lines due to the high flow resistance of the reactor coolant pump seals.
- 9. Pneumatic valve in high Fails to open Injection will route through the i pressure injection line alternate line. See discussion of consequences in Section 14.
9o 1 10. Check valve in injection Sticks closed See conunent on Item 6 above.
$ line (inside Reactor Building)
?
7 11. Injection line inside Rupture Flow rate indicators and alarm in the j Reactor Building two injection lines would indicate the gross difference in flow rates. Check
- valve in the injection line would pre-vent backflow from the reactor.
,' 'B. Low Pressure Injection System
- 1. Check valve at reactor Sticks closed This is considered incredible since
.h vessel those valves will be used periodically CD during decay heat removal and the open-ing force will be approximately 5000 pounds.
I
Table 6-1 (Cont'd) Component thlfun :t ion Comments and Consecuences
- 2. Air-operated injection valve Fails to open Second injection line will deliver total injection flow.
- 3. Safety valve Stuck open Loss of injection flow is small since valve is small.
- 4. Ileat exchanger Isolation valve left closed Other heat exchanger will take required injection flow and remove required heat.
Valves will be closed only for mainten-ance of heat exchanger. IIcat exchanger not required except as a flow path for first 25 minutes. on h' 5. Low pressure injection cooler Massive rupture Not credible. During normal decay heat gg remnval operation, heat exchanger will ej be exposed to higher pressure and approx-g inntely the same temperature as the g postaccident temperature and pressure. p l, 6. Low pressure injection cooler Out for maintenance Remaining heat exchanger will take 4 required injection flow. U
- 7. Low pressure injection pump Left closed Two remaining pumps will deliver more isolation valve than required injection flow.
- 8. Low pressure injection pump Sticks closed See conment on Item 7 above.
, check valve . 9. Low pressure injection pump Fails to start See comment on Item 7 above.
- 10. Air-operated valve at Fails to open Alternate valve vill permit full flow.
42= borated water storage tank
'f) outlet
,e O ,e
i
) [\
\m ,/
(' Tabic 6-1 (Cont'd) Component Malfunction Comments and Consequences
- 11. Check valve at borated Sticks closed Alternate check valve will permit full water storage tank outlet flow.
- 12. Air-operated valve per- Fails to open Redundant full capacity line and valve mitting suction from sump provided. This valve need not be actuated until 25-40 minutes after start of accident.
- 13. Reactor Building sump Becomes-clogged This is not considered credible because inlet pipe of the dual sump line arrangement, the size of the lines (18 in.), and the sump design. The two 18 in, recirculation
? lines take suction from the center por-
[ tion of the sump. A grating will be
-s provided over the sump, and additional E heavy duty strainers will be provided
$. for suction lines.
E C. Core Flooding System T 7 1. Flooding line check Sticks closed This is considered incredible based on O valves the valve size and opening pressure applied.
,~.
' 2. Flooding line isolation Inadvertently closed Administrative controls will prevent valves inadvertent closure; in addition, engineered safeguards signals will pro-vide for valve opening. In the event these control measures were violated, (j$ injection from one tank would be de- '
C;s layed by an amount of time equal to valve opening time. I
6.1.2.12 Environmental Considerations All of the emergency injection systems operating equipment is located outside the Reactor Building, and hence is not required to operate in the steam-air environment produced by the accident. The necessary transmitters f r the actuating signals are located in the Reactor Building, and are designed for the accident environment. 6.1.2.13 Performance Testing Except for cycling the valves, no special provisions are required to facilitate testing of the high pressure injection system because a pump is in normal use, and the remaining pumps will be rotated into service periodically. On the low pressure injection system, bypass lines are provided around the main valves in the borated water fill line to allow all of the injection line exterior to the Reactor Building to be tested. 6.1.3 DESIGN EVALl!ATION In establishing the required flows for the emergency injection systems, the following factors were considered:
- 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 due to maintenance is estimated to be 1 per cent. On this basis it is estimated that the probability that a major reactor coolant system accident would occur while a protective component was out for maintenance is two orders of magnitude below the low basic accident prob-ability.
- c. The equipment downtime for maintenance in a well operated station can often be scheduled during reactor shutdown periods. Where maintenance of an engineered safeguard component is required during operation, perio-die test frequency can be increased to further insure 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 emer-gency functions.
The core flooding system has the capability to recover the hot spot in approxi-mately 25 secords. The single failure analysis for the emergency injection pumping systems indicates that minimum available flow from the combined sys-tems is approximately 7500 gpm. This results when one low pressure pump fails to start leaving 6000 gpm in low pressure pumping capacity and 1500 gpm in high pressure pumping canacity. When the contents of the borated water stor-age tank have been added to the Reactor Building and the high pressure injec-tion pumps are secured, the 3000 gpm recirculated from the building sump to the reactor is more than adequate to cool the core at that time. 6-12 (Revised 4-1-67) .-- -
6.1.3.1 High Pressure Injection System If no puraps are out of service, the three pumps can deliver approximately 1500 gpm at the reduced discharge pressure. The safety analyses in Section 14 have shown that two high pressure injection pumps are sufficient to pre-vent core damage for those smaller leaks which do not allow the reactor coolant system pressure to rapidly decrease to the point where the low pres-sure injection system is initiated. The single failure analysis (6.1.2.9) demonstrates the effects upon the system of all credible failures. The emergency injection system valves are designed to open within 10 seconds. One of the three pumps is already in operation, and the pipe lines are filled with coolant. The two high pressure injection lines contain thermal sleeves at their connections into the reactor coolant piping to prevent overntressing the pipe juncture due to the 90 F water being injected into these high temp-erature lines. The equipment normally operating is handling 125 F water, and hence will experience no thermal shock when 90 F water is introduced. To provide the maximum number of high pressure injection pumps in the event of an accident, no more than one of the t.hree pumps, available to a unit, will be secured at any one time for scheduled maintenance. At least one of the two injection line valves, and one of the two suction valves from the borated storage tank, will be operational. x Operation of this system does not depend on any portion of another engineered safeguards system with the exception of the suction line from the borated water storage tank, which is common to both high pressure and low pressure injection systems, and the Reactor Building spray system. 6.1.3.2 Low Pressure Injection System Three pumps will deliver a total of 9000 gpm to the reactor vessel through two r parate injection lines. If one pump is out of service, approximately 3000 gpm will be delivered to the reactor vessel through each injection line. The low pressure injection system also provides for decay heat removal during routine or planned shutdowns. One of the low pressure injection lines is also used for this purpose. This equipment is not, however, used for decay heat removal until after the reactor has been cooled to 250 F and approri- , mately 150 psig pressure. In the unlikely event of an accident, the system will deliver the required quantity of water within 16 seconds af ter the reactor coolant system reaches the actuating pressure of 200 psig. The total anticipated delay time is composed of the following intervals:
- a. Total instrumentation lag ::::: 1 second
- b. Pump motor start-up (from time pump motor line circuit breaker closes until pump attains full speed) < l5-seconds p c. Injection valve opening time < 10 seconds O "
52 6-13 (Revised 4-1-67) l
- d. Borated water storage tank outlet valves < 10 seconds Total (b, e and d - not additive) :::: 16 seconds Special attention will be given to the design of core flooding nozzles (Figure 3-48) to insure that they will safely take the differential tempera-ture imposed by the accident ;ondition. Special attention will also be given to the ability of the injection lines to absorb the expansion resulting from the recirculating water temperature.
The water in the borated water storage tanks will be maintained at a minimum concentration of 2270 ppm of boron and at a temperature of about 90 F. Both the temperature and level in these tanks will be displayed in the control room, and alarms will sound when either condition is lower than the minimum set points. The water will be periodically sampled and analyzed to insure proper boron concentration. The low pressure injection system is connected with other safeguards systems in three respects, ie, (1) both the high pressure and low pressure injection systems and the Reactor Building spray system take their suction from the borated water storage tank, (2) the low pressure injection pumps and the Reactor Building spray pumps share a common suction line from the Reactor Building sump during recirculation of spilled water and (3) the low pressure injection system and the core flooding system utilize common injection noz-zles on the reactor vessel. 6.1.4 TESTS AND INSPECTIONS All active components, as listed in Table 6-2, of the emergency injection systems will be tested periodically to demonstrate system readiness. The high pressure injection system which operates continuously will be inspected for leaks from pump seals, valve packing and flanged joints. During operational testing of the low pressure injection pumps, the portion of the system subjected to pump pressure will be inspected for leaks. Items for inspection will be pump seals, valve packing, flange gaskets, heat exchanger leaks to ambient and safety valve leaks. I i l l l l l 53 6-14 (Revised 4-1-67)
Table 6-2 Emergency Injection Systems Performance Testing, High Pressure Injection Pumps One pump is operating continuously. The other two pumps will be periodically tested. High Pressure Injection Line The remotely operated stop valve in each Valves line will be opened partially one at a time. The flow devices will indicate flow through the lines. High Pressure Injection Pump The letdown storage tank water level will Suction Valves be raised to equalize the pressure exerted by the storage tank and the borated water storage tank. The valves will then be opened individually anl closed. Low Pressure Injection Pumps In addition to use for shutdown cooling, these pumps will be tested singly by opening the borated watet storage tank outlet valves and the bypasses in the borated water storage 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. Th= 1~- pecccure inj:ctiea p re ea*ving both reector m.its will Se tested fcr fle. ;c cach unit. Borated Water Storage Tank The operational readiness of these valves Outlet Valves wil* 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 Valves With pumps shut down and borated water storage tank outlet valves close<., these valves will be opened and reclosed by operator action. Valve for Suction from Sump With pumps shut down and borated water storage tank outlet valves closed, these , valves will be opened and reclosed by ; operator action. I
-s i e
t [ 6-15 (Revised 4-1-67)
)
i t I
s 6.2 Ef4CT M BUILDING SPRAY SYSTEM 6.2.1 DESIGN BASES The Reactor Building spray system is designed to spray 3000 gpm of borated water into the Reactor Building wnenever the Reactor Building pressure reaches 10 psig. The design basis is to provide sufficient heat removal capability to maintain the postaccident Reactor Building pressure below the design value. This requires an initial heat removal capability of 240 x 10 6 btu /hr. 6.2.2 SYSTEM DESIGN 6.2.2.1 Piping and Instrumentation Diagram The piping and instrumentation diagram of the system is shown in Figure 6-6. The diagram shows Unit 1 in the accident-operating mode. Unit 2 is not shown but utilizes similar equipment and arrangement. The system serves no function for either unit during normal operation. The system for each unit consists of two pumps and the necessary piping, valves and spray nozzles. The system initially takes suction from the borated water storage tank. When low level is reached in the borated water storage tank, suction is transferred to the Reactor Building sump. The system then recirculatee me spilled injection and spray water accumulated in the reactor sump using any subcooling in the spilled water as a heat sink. When the spilled water becomes saturated, cooling is provided by diverting part of the low pressure cooler outlet flow through the reactor spray pumps. 6.2.2.2 Codes and Standards The equipment will be designed to the codes and standards listed in the introduction to Section 9. 6.2.2.3 Material Compatibility This system, like the low pressure injection system, is composed of both aluminum and stainless steel components. The comments of 6.1.2.3 relative to the methods of joining these components also apply to this system. 6.2.2.4 Component Design The entire system will be designed for 300 psig at 300 F. The system design conditions were selected to be compatible with the design conditions for the low pressure injection system since both of these systems share the same suction line. 6.2.2.5 Coolant Storage This system shares borated storage tank capacity with the low pressure injection system and the high pressure injection system. c 6-16 (Revised 4-1-67) -
-- . J5
6.2.2.6 Pump Characteristics The Reactor Building spray pump characteristic curves of total dynamic head and net positive suction head as a function of flow are shown on Figure 6-7. 6c2.2.7 Heat Exchanger Characteristics No separate heat exchanger is provided, but a crossover line is provided to the spray pumps from the discharge of the low pressure injection coolers. 6.2.2.8 Relief Valve Settings Since the suction of this system is shared with the low pressure injection system, the suction side relief valves described in 5.1.2.8 are also appli-cable to this system. 6.2.2.9 Reliability Considerations A failure analysis has been made on all active components of the system to i show that the failure of any single active component will not prevent ful-filling the design function. This analysis is shown in Table 6-3, 6.2.2.10 Missile Protection The spray headers are located outside and above the reactor and steam gener-ator concrete shield. During operation a movable 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. 6.2.2.11 Actuation The Reactor Building spray system will be activated by a single parameter signal. Two of three signals signifying high ReactorIhdiding pressure will start both of the Reactor Building spray pucps, open the Reactor Building sprayinletvalves,openthesuctionvalvesfromtheboratedwaterj 3 g g_rgdgg g the low pressure service water outlet valves cnner on the _
- pr:y -aalere. The system components may also be actuated by operator action from the control room for performance testing.
6.2.2.12 Environmental Considerations All of the active components of the Reactor Building spray system are located outside the Reactor Building, and hence are not required to operate in the steam-air environment produced by the accident. The required instrumentation transmitters located inside the Reactor Building are designed for the postaccident environment.
,.J 6-17 (Revised 4-1-67) ) qg
Table 6-3 Single Failure Analysis Reactor Building Spray System Component Malfunction Comments and Consequences
- 1. Reactor Building spray Clogged Large number of nozzles (120 on each of nozzles two headers) renders clogging of signi-ficant number of nozzles as incredible.
- 2. Reactor Building spray Rupture Air-operated valve on ruptured header header uill be closed. Second spray header will deliver 50 per cent flow, and sys-tem is supplemented by Reactor Building emergency cooling units.
p 3. Check valve in spray Sticks closed No flow through this header. Second g header line header will deliver 50 per cent flow. f 4. Air-operated valve in Fails to open Second header delivers 50 per cent flow.
$ spray header line o- 5. Reactor Building spray Isolation valve left Flow and cooling capacity reduced to 50 7 pump closed per cent of design. In combination with y emergency coolers, 150 per cent of total design requirement is still provided.
- 6. Reactor Building spray Fails to start Flow and cooling capacity reduced to 50 pump per cent of design. In combination sith emergency coolers, 150 per cent of total design requirement is still provided.
+.7. Valves See Table 6-1 for discussion ot valve malfunctions.
(J1 N
)
6.2.2.13 Performance Testing The active components of the Reactor Building spray system will be tested on a regular schedule as follows: Reactor Building These pumps will be tested singly by opening the valves Spray Pumps in the test line and the borated water storage tank out-let valves. Each pump in turn will be started by operator 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 Storage Tank These valves will be tested in performing the pump test Outlet Valves above. Reactor Building With the pumps shut down and the borated water storage Spray Injection tank outlet valves closed, these valves will each be Valves opened and closed by operator action. Reactor Building Under the conditions specified for the previous test and Spray Nozzles with the Reactor Building sprav valves alternately open, low pressure air will be blown through the test connec-O tions. 6.2.3 DESIGN EVALUATION , l For the first 25-40 minutes following the maximum blowdown loss-of-coolant I accident, ie, during the time that the Reactor Building spray pumps take { their suction from the borated water storage tank, this system provides more ' than 100 per cent of the required heat removal capacity to reduce building ) pressure. ) l The system design is based on the spray sater being raised to the tempera- ' ture of the Reactor Building in falling through the steam-air mixture within the building. Any of the following combinations of equipment will provide sufficient heat removal capability to maintain the postaccident Reactor ! Building pressure below the design value: I
- a. The Reactor Building spray system.
- b. Three emergency cooling units.
- c. Two emergency cooling units and the Reactor Building spray system at one-half capacity.
The Reactor Building spray system shares the equipment on the suction side j of the pumps with the low pressure injection system, and shares the suction line from the borated water storage tank and the tank itself with the high l and low pressure injection systems. The Reactor Building spray system is designed to deliver 3000 gpm through the _ 6-19 (Revised 4-1-67,)
spray nozzles within 15-20 seconds af ter the Reactor Building reaches 10 psig. O The expected lag time is the same as for the low pressure injection system (6.1.3.2). 6.2.4 TESTS AND INSPECTION Perfornnnce testing of all active components will be accomplished as described in 6.2.2.13. During these tests the equipment will be visually inspected for leaks. Leaking seals, packing or flanges will be tightened to eliminate the leak. Valves and pumps will be operated and inspected after any maintenance to insure proper operation. 6.3 REACTOR BUILDING EMERGENCY COOLING UNITS 6.3.1 DESIGN BASES The Reactor Building emergency cooling units are designed to remove suffi-cient heat from the Reactor Building to prevent exceeding the peak design pressure at any time following a loss-of-coolant accident. They will return the Reactor Building pressure to near atmospheric within 24 hours af ter an accident. This requires an initial heat removal capability of 240 x 10 6 btu /hr. 6.3.2 SYSTEM DESIGN 6.3.2.1 Piping and Instrumentation Diagram Figure 6-9 schematically illustrates the emergency cooling units. Each emergency cooling unit consists of a fan and a tube cooler. The Reactor Building atmosphere is circulated through the cooling unit by the fan. Cooling. water for the emergency cooling units is supplied by the low pres-sure service water system. Performance of the emergency cooling units is monitored by flow instrumenta-tion in the service water return line from each cooler and by the Reactor Building temperature and pressure instrumentation. 6.3.2.2 Codes and standards The equipment will be deeigned to the applicable coder and standards listed in the introduction to Section 9. 6.3.2.3 Materials The materials for the Reactor Building emergency coolers will be selected to be compatible with the use of untreated service water to minimize corrosion. A flow of cooling water will be maintained during normal operation to pre-vent static water pitting corrosion. l 6.3.2.4 Component Desien The emergency cooling units will be designed to maintain postaccident oper- _ ability. 59 6-20 (Revised 4-1-67) , , , . . . .
f i) V 6.3.2.5 Reactor Building Emergency Circulating Fan Characteristics Each fan is of the centrifugal type with backward-inclined blades. For post-accident operations, it is rated at 54,000 cfm and 3 in. H 2O static pressure. The power requirement is about 100 hp. 6.3.2.6 Reactor Building Emergency Cooler Characteristics The Reactor Building emergency cooler, shown in Figure 6-10, is located on the discharge of the fan. The air-steam mixture flows across the tube bank, resulting in condensation of a portion of the steam and removal of sensible heat from the air. each unit is 80 x 10gtbeu/hr. design Figure postaccident conditions, 6-11 shows the rated capacity the approximate capabilicyof of each unit at reduced Reactor Building temperature and pressure conditions. The physical characteristics of the unit are as follows: Face dimensions, ft 8x8 Depth, ft 3.6 Type Tubular Tube size, in. 3/4 Surface area, ft 2060 6.3.2.7 Reliability Considerations Inside the Reactor Building, the emergency cooling units are located outside the secondary shield at an elevation above the water level in the bottom of the Reactor Building at postaccident conditions. In this location the units are protected against credible missiles and from being flooded during post-accident operations. The Reactor Building emergency cooling units provide complete redundancy for the Reactor Building spray system or vice versa. The major equipment is pro-vided in multiples of three, with three emergency cooling units and three service water supply lines inside the Reactor Building. A failure analysis of the emergency cooling units is presented in Table 6-4. 6.3.2.8 Missile Protection All equipment, piping, valves and instrumentation in the Reactor Building are located to minimize the possibility of missile damage. The emergency cooling units and associated piping are located outside the secondary concrete shield-ing. p . O ___
<n 6-21 (Revised 4-1-67) .,.,.
WU
Table 6-4 Single Failure Analysis for Reactor Building Emergency Cooling Units
', . Component Malfunction Connents and Consequences
- 1. Unit Circulating Fan Fails to operate Multiple independent units are provided:
The operating units will be supplemented by the Reactor Building spray system. The fans may be tested during nornut reactor operation to provide added assurance of reliable emergency opera-tion.
- 2. Emergency cooler Failure of tubes Tube failure is considered unlikely during emergency operation since AP p across tube is less than during normal y service. If failure does occur, service
, water will be spilled into the cooler g since service water pressure is above
;i Reactor Building accident pressure, y Tube leakage can be detected by reduc-a tion of service water flow at discharge f from the cooler and the failed cooler ';' can be isolated.
v
- 3. Cooler service water Fails to fully open Valve will normally be partially open.
outlet valve If the valve fails to open fully, the unit will operate under reduced heat removal capability. The Reactor Build-ing spray system will supplement the heat removal capability of the other units.
- 4. Cooler service water Inadvertently left Valve status will be apparent from lack
, inlet valve closed of flow,and valve may be opened by operator action. If the valve fails to
& respond, the Reactor Building spray e O ,9
i (
,.y Table 6-4 (Cont'd)
Component Mau1 function Comments and Consequences system will supplement the other two emergency cooling units. ,, 5. Service water piping Pipe failure Pipe rupture not considered credible t inside Reactor Building due to location outside the secondary shield. If failure should occur, the pipe failure can be detected and the line isolated,.
- 6. Service water pump Fails to operate The two remaining pumps will provide full low pressure service water flow to i
all components. T
~
L4 l k t E.
?
- 'T e
m 6 I 4 o 9 9 9 . t ' m N I _ _ _ _ _ . ____ - _ _ . _ _ _ _ _ _ _
6.3.2.9 Actua tion In the event of a loss-of-coolant accident resulting in a Reactor Building pressure increase, the cooling units are placed in operation as follows:
- a. The low pressure service water and return line valves are fully opened to allow design emergency flow of service water. (Supply valves are normally open.)
- b. The cooling unit fans are started.
6.3.2.10 Environmental Considerations In the Reactor Building the motors associated with the fans are designed for unlimited operation in the design postaccident conditions. 6.3.2.11 Performance Testing The emergency cooling units will be tested on a regular period as follows:
- a. The fans will be started and inspected for proper operation.
- b. The return line service water valves will be opened, and the line will be checked for flow.
6.3.3 DESIGN EVALUATION The desiga of this system for postaccident operations centers around the emergency coolers. The heat removal capacity as a function of Reactor Building temperature is shown in Figure 6-11. 6.3.4 TESTS AND INSPECTION The equipment, piping, valves and instrumentacion are arranged so that they can be visually inspected. The emergency cooling units and associated piping are located outside the secondary concrete shield around the reactor coolant system loops. Personnel can enter the Reactor Building during power opera-tions to inspect and maintain this equipment. The service water piping and valves outside the Reactor Building are inspectable at all times. Opera-tional tests and inspections will be performed prior to initial start-up. 6.4 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS 6.
4.1 INTRODUCTION
Use of normally operating equipment for engineered safeguards functions and the location of some of this equipment outside the Reactor Building require that consideration be given to direct radiation levels af ter fission pro-ducts have accumulated in these systems and with leakage from these systems. Although the engineered safeguards equipment is designed for remote opera-tion following an accident, long-term postaccident operation could necessi-tate manual operation of certain valves.
'67J 6-24 (Pevised 4-1-67) - --
(3 \') The shielding for components of the engineered safeguards is designed to meet the following objectives in the event of a MHA:
- a. To provide protection for personnet to perform all operations necessary for mitigation of the accident witsin the limits of 10 CFR 100.
- b. To provide sufficient accessibility in all areas around the station to permit safe continued operation of :he second unit.
6.4.2 SID01ARY OF POSTACCIDENT RECIRCULATION Following a loss-of-coolant accident, flow is initiated in the low pressure injection and decay heat removal system from the borated water storage tank to the reactor vessel. Flow is also initiated by the Reactor Building spray system to building spray headers. When the borated. water storage tank is exhausted, recirculation from the Reactor Building sump is initiated to both the reactor vessel injection and the Reactor Building sprays. The postacci-dent recirculation complex includes all piping and equipment external to the Reactor Building as shown on Figures 6-2 and 6-6 up to the stop and test line valves leading to the borated water storage tank. 6.4.3 BASES OF LEAKAGE ESTIMATE 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. The leakage sources considered are:
- a. Valves (1) Disc leakage when valve is on recirculation complex boundary (2) Stem leakage (3) Bonnet flange leakage
- b. Flanges
- c. Pumping 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 tests or nor-mal operation. The boundary valve disc leakage is retained in the other closed systems, and therefore will not be released to the Auxiliary Building.
While valve stemleakage has been assumed for all valves, the manual valves in the recirculation complex are backseating and do not rely on packing alone to prevent stem leakage. [, 6-25 (Revised 4-1-67) ....e
d.4.4 LEAKAGE ASSUMPT10NS Source Quantities
- a. Valves - Process (1) Disc leakage 10 cc/hr/in. of 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 as this peak 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 per cent relative humidity. Under these conditions, approximately 22 per cent of the leakage upstream of the coolers and 4 per cent of the leakage downstream of the coolers would flash into vapor. For the analysis, however, it was assumed that 50 per cent of the leakage upstream of the coolers would become vapor because of additional heat transfer from the hot metal. 6.4.5 DESIGN BASIS LEAKAGE The design basis leakage quantities are tabulated in Table 6-5. 6.4.6 LEAKAGE ANALYSIS CONCLUSIONS It may be concluded from this analysis (in conjunction with the discussion and analysis in 14.2.2.4.4) that lankage from engineered safeguards systems outside the Reactor Building does not pose a public safety problem. 65 k; 6-26 (Revised 4-1-67) ! I
1 I w
> Table 6-5 Leakage Quantities To Auxiliary Building Atmosphere Estimated Quantities Liquic Vapor 1 No. of Per Source Total Phase Phase l Leakage Source Sources drops / min cc/hr cc/hr cc/hr Low Pressure Injec-tion 6 Decay Heat Removal System
- a. Pump Seals Low pressure injection pump 6 50 900 450 450 Spray pump 4 50 600 300 300
- b. Flanges (*) 114 10 3320 1800 1520
- c. Process Valves 35 1 105 68 37
- d. Ins trumentation '
Valves 25 1 75 72 3
- e. Valve Seats at Boundaries 11 (**) 750 580 170 Total 5750 3270 2480
(*) Assumes, conservatively that all components and valves are flanged. , (**) Assuming 10 cc/hr/in. of nominal disc dismeter. l 6.5 REACTOR BUILDING PENETRATION ROOM VENTILATION SYSTEM 6.5.1 DESIGN HASES i This system is designed to collect and process potential Reactor Building i penetration leakage to minimize environmental activity levels resulting l from penetration leakage. Experience (l) has shown that Reactor Building leakage is more likely at penetrations, such as electrical cables and air purging valves, rather than through the liner plates or weld joints. 1 6.5.2 SYSTEM DESIGN 6.5.2.1 Piping and Instrumentation Diagram The system is shown schematically in Figure 6-15. Penetration rooms are l formed adjacent to the outside surface of each Reactor Building by enclosing ! the area around the majority of the penetrations. The only penetrations for l [} v either Reactor Building which do not pass through this area are: 6-27 (Revised 4-1-67) , ,,,
, 1 I
v
- a. Two main steam lines.
- b. Permanent equipment hatch which contains a double gasketed closure.
- c. Normal personnel access lock,
- d. Emergency personnel access lock.
This system is designed to collect and process the leakage from all penetra-tions. The main steam lines are not considered a source of significant leak-age because they are welded to the liner plate. The access openings can be tested during normal operation and are not considered sources of significant leakage. There are, however, double seals at each of these access openings, and the space between these double seals is connected to the penetration room. The main function of the system is to operate in the postaccident cond 4. tion. It will also operate intermittently during normal conditions as required to maintain satisfactory temperatures in the penetration rooms. When the sys-tem is in operation, a slight negative pressure will be maintained in the penetration room. Following a loss-of-coolant accident, a Reactor Building isolation signal will start one of the two full-size blowers. A power-operated butterfly valve, which opens when the blower starts, is provided at the discharge of each blower to prevent recirculation through an idle unit. Air drawn into each of the penetration rooms is discharged to the station vent through a filter assembly consisting of a prefilter, an absolute filter and a charcoal filter in series. The entire system is designed to operate under negative pressure up to the fan discharge. In all cases, the flow from the penetration room will exceed the total maximum Reactor Building leakage rate. (Reactor Building leakage rate of 0.5 per cent per day amounts to approximately 30 scfm as compared to an anticipated penetration room flow rate of 2000 scfm.) The Reactor Building purge equipment, if running, will be shut down by a Reactor Building isolation signal, and the three valves in each purge line penetration closed. After closing of the external valves, a small normally open valve vents the leakage, if any, from the two outermost valves into the penetration room. 1 The system may be actuated by the operator as required during normal opera-tion for testing and temperature control. Temperature indication and radia-tion monitors are provided for operator information. Lifferential pressure indicators are provided across the filters. 6.5.2.2 Codes and Standards i The equipment in this system will be designed to the applicable codes and standards listed in Section 9. 6.5.2.3 Material compatibility Since this system will not experience high temperature or corrosive fluid __ l , '. l 6-28 (Revised 4-1-67) h[
/~'
V} service, it will use carbon steel and suitable coatings to obtain desired service life. 6.5.2.4 Equipment Accessibility The system equipment is fully accessible during all normal station operation for maintenance and perfornance testing. 6.5.2.5 Reliability Considerations Each penetration room is provided with two blowers and two filter assemblies. All equipment will be controlled from the main control room. Both blowers discharge through a single line to the station vent. During normal operaticn, this system is held on standby with one of the two blowers aligned with one of the two filter assemblies. The engineered safe-guards signal from the Reactor Building will actuate the blowers. Control room instrumentation will monitcr operation. The system can be tested during normal operation. 6.5.3 DESIGN EVALUATION A single failure analysis of the various portions of this system is presented in Table 6-6. ( Table 6-6 Single Failure Analysis for Reactor Building Penetration Room Ventilation System Component Nbifunction Comments and Consequences
- 1. Blower Fails to start Spare blower placed in service.
- 2. Blower Fails during Alarm in control room will indicate service loss of negative pressure, and spare blower can be placed in service.
- 3. Blower valve Fails to open Spare blower is placed in service.
- 4. Filter valve Fails to open Failure not considered credible since one filter will always be lined up to operate when needed.
- 5. Ductwork Leakage Leakage of unfiltered air is inward since ductwork will be maintained at negative pressure.
= -
\
N ..... v 68 6-29 (Revised 4-1-67)
6.6 REFERENCES
O (1) Cottrell, W. B. and Savolainen, A. W., Editors, U. S. Reactor Contain-ment Technology, ORNL-NSIC-5, volume II. O 69 - 5-30 (Revised 4-1-67)
1 i l l 1 1 i I i I 9/ -
1 l I l l l 1 1 1 i 1 1 I 1 I i 1 03
l
/
% O i
+ Table 6 6 Single Failure Analysis for Reactor Building Penetration Room Ventilation System Component Malfunction Comments and Consequences
- 1. Blower Fails to start. Spare blouer placed in service.
- 2. Blower Fails during service. Alarm in control room will indicate loss of i
? negative pressure, and spare blower can be "m placed in service.
- 3. Blower valve Fails to open. Spare blower is placed in service.
- 4. Filter valve Fails to open. Failure not considered credible since one fil-ter will always be lined up to operate when needed.
^
5 Ductwork leakage Icakage of unfiltered air is inward since ductwork will be maintained at negative ! pressure. 8
~2 s
~a m b
I ..
I O b b (S-t k.,
- m
=
2c A O - I esi
... -y.tsr 1 l
=
30
'*~'
=e scany Q Puup3
- c. --. -* I I
ep
- m aeamenc.c o COC*.M uni?S SIMPLIFIED SCHEMATIC DIAGRAM OF ENGINEERED SAFEGUARD SYSTEM FOR CORE AND BUILDING PROTECTION puu,nwu OCONEE NUCLEAR STATION FIGURE 6-1 _
79
~ ~ ~ ~ ~ ~ - - - ~ - - . _ _ _ _ _ _
I r, 1 ll l l
~
l RE ACTOR DOILD NG l
)
- I ~
n a 6 6 4 L 6 4 6 L RE ACOR SciLOiNG SPAAv3 -
'/
u u u u u u u a a u l l ES4 I' - CORE CORE FLOCD% L ' L_ r ocopr Nr, 1ANK 7A44 t
)
Y #6 .M l
]
r i
-3
'I f~-
J. ( RE ACTOR f 4 e sNLET LINE f ,g 7
%EMEL UNIT 1 /
\ .,- 1 /
1- 9
~
LP bw _ y I ,,,,I ty
^#.".'.
k .I ae EmtsceNtv C00LIN(. UNITO _ _ RE MTM BLOG EmtoGENcv SA)M5" NOTC: UNIT C AE' DUPLICATE SwOwN AI FOR LEGE SEE FG.S
!/
, o. 80
l i t i i asi eJ 4 J v BO4 ATE D WA'EA SiCAAGE 7ANW UNii I tis r--! -C ^. l
,- s o 9 '( :Y 4'
nv. O cIe= m l e- -r- rs ,
}El%" 13 - -
'~~
l ' * b) b h es i .. ,P.A.
, ~ _ . ,
t>.- r- i - , v ,.- puo,3 b kh b A-
- imectos POMP, i s
'l i t
-CS-i ts-t -
SIMPLIFIED SCHEMATIC DIAGRAM OF
,?, ENGINEERED SAFEGUARD SYSTEM FOR N CORE AND BUILDING PROTECTION 8,-
REr 4 4-67 LODeGEMEM7 Rf DR A*N - SHARFO E QWuf NT Et w/%ATED 8 m e OE uni? 1 EME RGENCY COOLeJG LNTS CHANGED cyQ suu rowla OCONEE NUCLEAR STATION @ kMAICLA7utti .i FIGURE 6-1 2.81 i
-m m
9'?* *EM h, 1 4 l i
,,...,e._, 1, i% .t - u %t e
M K. t. R9 %
.r. =
TO ME AOTOR / ,g
-~
%ctT t ut j "B
= 70 veuw = cts
[0 o - o 4 %% m o,soa%._tq_. sumv wev f A 4, v 'V COME ' 'W d
,.s^ Coni '~ ""
rtooD,%3 F toot >W mme --ms .c.s. ys
- V V k .) %) p 9 .- x - e,ut u,#
. .u k u,~t - x -.
w+t-H - - x +% M % TUNE m ~.NC L
,-e w_._ _2 i i
NM MMW
'M ns esac,3e=
.m 4 l 4 :<
m _ g .v "O > I F u e5 r' 69 000Lt#1 tr-t es- J
'x' ne
.. ,l f f bh Y $
acuCToa p, _I I BLOG t Mt uc-t ucy M9 l 9 gyg g p pum,
" ?
AB i.aoi tN SI DE. OU T SIDE paa SEE f UhJ5~ DvP S=40 82
9 4 l I i
<iwun I
i l l T3
'1% WP1
- PuHif C A7. %
tiOH ATCD-,.3 y wAitu , & 8 LETOOwu US4E I g l
- 1 i t _ - . . ,
_: -{ rew O f 1 63LA* LL 'O 1 I
* "s5 '
_g ~. .,1
,p%(,
,,. ,O Ga
~ g_Keh 2,e 5T '54* A,E
** Yh%m .
y
-6;
}a s.]-] .c-h
_.<,-- t c,- y ,,..
. 4.
Ly - x a h x, > x.
. :& ^
ts. ,
*D %ETCN x, .- .~,
y} am-
,' ( S-I 49 % - F 5- e f
'X..
l' : j - T3 A BS POMPS EMERGENCY INJECTION SYSTEMS b 98_ , ; 4 >.47 ire. ,. s.m event etwa ec.c o ,w;-c a . , ,c o nEP851 OCONEE NUCLEAR STATION e u.. ~an v , FIGURE 6-2 l ce r o. ~ + t
-u % .
J 83 ,
.f
O
, 36 i j 8
4 Z 18 NPSH
#'j/ J o
g Z 0 l 600 a 500 gL
~
7000 j
/ BHP h 6000 '% l s N W \ i
( 5000
\ ,
l i' E ToH Q 4000 \ W \
~
g 3000 \ N
\ '
I h2000 \ b 1000 0 100 ' 00 2 300 400 '500 j GALLONS PER MINUTE HIGH PRESSURE INJECTION PUMP CHARACTERISTICS
# 'O' OCONEE NUCLEAR STATION v FIGURE 6-3 84 ,
O (J
! i i i . / zo l f NPSH [
19 480 , g 18
+ l -
/
W 440 - - - ---- -
}
17 9 s z 400 i
! i -- - --- - -
/ 16 , ' ,
- t, - - - -
k HEAD 15 360 --! W ' ;
~
g 320 't-- 1 r - -~
--- 7 14 A l l y 280 --
---l'---+--- - - - - - -
N 13 Nd & j- q 240 , 12 p_.
' I 600 .
_ BHP #" 400 g l 4 p .. l. l 200 W L I 3200 4000 4800 5000 O h 0 800 1600 2400 GALLONS PER MINUTE LOW PRESSURE INJECTION PUMP CHARA 3TERISTICS V FIGURE 6-4 85 I
O O 300 Design Conditions > Water Flow Per Cooler - 6000 GPhi 260 L.P. Service Water Flow Per Cooler- I 6000 GPh1 L.P. Service Water Inlet Temperature-75F f cf '20 r j Cooler Inlet m 7 kl 3. Temperature / w 180 g [
/ /
/ Cooler Outlet Temperature 140
/
100 ' 0 40 80 120 160 200 HEAT TRANSFERRED, BTU /HR x 10-6 LOW PRESSURE INJECTION COOLER CAPACITY OCONEE NUCLEAR STATION tj FIGURE 6-5 B6
i v t u st oE OUTSI D E RB RB
% AIR CONN FOR SPRAY TEST e s-i p eat.o) i
' m 2 111111
.RB RE ACTOR BLOG
$ PRAY MEADER a RB f' {H$(j -
i m.2 i-111111 " '
\-e.<.o gn TEST LINE
#h AIR CONu FOR SPR AY TEST v
RB RB [U RSS T- --
' J PUMPS ld ESl FROM LA COOLER OUTLST NOTES:
I- UNIT I iS A QU PLIC ATE AR R ANGEMCNT TO TH AT SHOWN ABOVE FROM LP PUMP F-COR LE GENO NOMtiNCL ATU RE SUCTION HEADER SEE 5:lGURE 9-l ) AEV: 4-i-67 REDRAWN - SHAPED EOUIPMENT ELiviNATED 6 R 0 SFRAY CCOLERS CELETED REACTOR BUILDING SPRAY SYSTDI o OCONEE NUCLEAR STATION ! 87
O 500 b 400 TDH / K w % 4 5 300 a ! b 200 25 I 100 s y -
'5 s
10 Z 0 5 10 15 20 CAPACITY IN 100 GPM REACTOR BUILDING SPRAY PUMP CHARACTERISTICS
?"t!w OCONEE NUCLEAR STATION FIGURE 6-7
( __
..... 88
l 320 Design Conditions , Water Flow Per Cooler - 1500 GPM 7
'80 L.P. Service Water Flow Per Cooler - /
1500 GPM L.P. Service Water Inlet Temperature-75 F W td 240
$ Cooler P Inlet
$ r Tempe rature p 200 h1 [
/
/
160 /
)
/ /
/
Cooler Outlet 120 Temperature 7 a l 80 0 20 40 60 80 100 l HEAT TRANSFERRED, BTU /HR x Ig6 REACTOR BUILDING SPRAY COOLER CAPACITY m g o a OCONEE NUCLEAR STATION () L FIGURE 6-8 89 h
s ENomagmeo e j f,P= g , g :___,__________.______
; ,____.7____ ______,_____7____, , S N EM l no A= m Am A me m
- m o in
_________..a 95 ps ,.b kL ps) ps' < ps Y te/s R8 R8 MSIDS IN SI DE IN SIDE CUTSsDE ES-. et me I i l 8 sy. . S. , C Maesst 6 x '
=
- - m.S-l l
.v ,, .S.
k l k 4 S EMERGENCY COOLING UNITS NOTE t *,
- 6. UNIT & st A OUDUC ATE ARR ANGematwT TO TH AT SMCo#N A80v4 2-FOR LEGEND Noodt4CLATust SE E P'Gu ME 9 *9 att 4067 streaeN.eEv0vED # 8 C0wacNENT C00 LAG AATER Sv5?EU t CHANGE
- a 8 FWE*'ANCY C00 LAG UMTS REACTOR BUILDING EMERGENCY COOLING UNITS I 1
a, s mirowa DCONEE NUCLEAR STATION FIGURE 6-9 90 ' l
i a FAN b - . Jr' T, I I I l f
"= ,
/ N / \
,_ s
%- - .smsm..g
----_e
\ /
l 9 rIow l REv: 4.1 67 RECRAwk.:.A At.5HE RS REPLACED 61 COOLING CCILS i l t REACTOR BUILDING EMERGENCY COOLER esu re OCONEE NUCLEAR STATION O ' FIGURE 610 91 l 1
i O 333
- n. Design Conditions f
(!) 10% Saturated Air w' a; (2) 54, J0J ACFM Air Flow
$ (3) 75 F Low Pressure S rvice Water /
< {4) 1350 gpm Service Water Flow r 1 220
# /
2 [ Y , / s /
'180 5 /
l / s 140 f W r 100 0 1J 20 30 4 ') SJ 60 70 SJ HEAT TR ANSFERRED. BT U/HR x iv*b AE V: 4 -l-67 RECRfWN REFLECTS SYSTEM CHANGES REACTOR BUILDING EMERGENCY CQOLER HEAT REMOVAL CAPACITY o OCONEE NUCLEAR STATION
- 11
. . ... 92
O 80 0
$W
$5 [
> 60 /
og 50
/ - - -
W[d # y 40 --
"H 30 > --
O %
*k /
/
20 I W4 UW
/
10 ' II. ._ __ __ gH W W Q. 0 100 120 140 160 180 200 220 240 260 280 300 REACTOR BUILDING TEMPERATURE, F REACTOR BUILDING POST ACCIDENT i' STEAM-AIR MIXTURE COMPOSITION O MEN'te OCONEE NUCLEAR STATION
' FIGURE 6-12 9
l 93
l O l /
-y 40 g NPSH j , g g - - ~
g 350 ------ -- - - - -
+ .
" - * * ' 20 g ra .L '
l l l . I g W 300 - -- - 4- ; - . 10 E 250 . Head Q l l l I
< 200 i .
- 4 ?
i ! .
=
150 -- - --- - -
! r l
100 * * '- t bt -- o 50 -----1------- - ! b-
- A 150 .
H l l 1 l
~ '
,__t . _ . ,__.-. 125 c5
\ _ -- . _ _ . ._. BHP - - -
- h- 100
- l
.I l 75
[= 50 t :a 1 0 200 400 600 800 1000 1200 1400 1600 1800 2000 GALLONS PER hilNUTE FIGURE 6-13 REACTOR BUILDING COhiPONENT COOLING PUhiP CHARACTERISTICS nime OCONEE NUCLEAR STATION FIGURE 613 l 94
300 , , , , 3 g Design Conditions 280 - (1) 5000 GPM Service Water Flow (2) 2700 GPM Component Cooling Water Flow 260 - (3) One Component Cooler Equivalent to Two Emergency Cooling Units , 240 Component Cooling Water 220 Inlet Temperature Y CE 200 A -- M D F >
< 180 M
W 160 f a S Service Water Outlet Temperature I 140 , l l 120 > Component Cooling 100 g Water Outlet - Temperature j/ Service Water Inlet Temperature 60 0 20 40 60 80 100 120 140 160 180 200 HEAT LOAD, BTU /HR x 10 l l REACTOR BUILDING COMPONENT COOLER l HEAT REMOVAL CAPACITY ' w pt roan OCONEE NU:: LEAR STATION j 'u s FIGURE 6-14 s l . 95
e (
\
( s 3,e c. r o.t b. c Jilk G m m* *d m e nl~ r DG -f. $~/.Is-/g,, 7
^g M
%J l
l l PE NET R ATION ROOM FAN 5 E51---- m m - -- - E 5 1 I I I TO REACTOR BLDG ' l PO , PO PURCE EXHAUST FAN Cj (gg p m 4p
,-y g ? PENET RATION ROOM FILTERS s E31 I a
?&l 4&L' '
',s-- % i I /6Th J-{)- -- E s , g v l
-Es a PE NCT R ATION ACOM UNIT 1
-Q (RB PURGE
- valve l
I Es l I i REACTOR ES1 i BUILDING t PENETRATIONS . PO ESl-- k, g , , _ , , _,, . ;FROM Rd Ra puRct , PuRoe ! UNIT 1 ^'"/ REACTOR BUILDING / NOTE Q c a u, f)isa- ,f)+:* j
/.',.-I,l
. )-l >
l- P) 44, . swa - " v' w aticiv '
,r' v-/*.. __ -l / 1 / ,
,1-. ,1 tvu
~ _ .u F@
( .// 'u-u c < 6 4 , , v -
- f. ,- J'. ./
a a '. --s
... -) / sa gn a r c re , q.- r,
,) r - - -
.a ne n ~
i, ~* l'rs u v,r iir s. l };u 7'), g.pg YMS /!At thdb!!rfen/,} p4/ c,s '?%5 c/w,, i 4
\ .. ..
96 g.
r e,
/!
\
4 l l l l l
.T IO N VENT d i l
6H ; PENETR ATION M ROOM FANS E S - - - m m - - - - E S-2 I I I TO RE ACTOR BLOG Po - PURCE E1HAU57 FAN PE N E7 R ATIOM ' 1 9 2C dPt ROOM FILTER $ g y ?(( j ,
? Y ES 2 9a Af l L
l l g.. A es-2--Q [ I
$ PE N E T R AT I O N ROOH UusT 2 ES 2- ,
RBFURCEl* VALVE LS 2 l l Eye PO I
- Es-2 ACTOR BLDG iUPPLY FANS
*h-~} ~
~~'
RB PURCE s VALVE UNIT 2
\. RE ACTOR BUILDING VE T R ATIO N ROO M REACTOR BUILDING TiL ATiO N sysTE M PENETRATION ROOM $ UNir i SHOWN VENTIIATION SYSTEM ,, @PE R ATING MODE.
) LE0CND REv. 4 -I- 67 Y %ENCLATURE ACDEP E S SIGNALS Q FIGURE 9-l. 08" ron OCONEE NUCLEAR STATION FIGORE 6-15 N ..
.... 97 i e
.}}